ML20150D838
| ML20150D838 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/21/1978 |
| From: | Boyd R Office of Nuclear Reactor Regulation |
| To: | Mittl R Public Service Enterprise Group |
| References | |
| NUDOCS 7812070061 | |
| Download: ML20150D838 (41) | |
Text
I
[p Ef Gybe,d UNITID STi,TES
[h NUCLEAR REGULATORY COMMISSION 3 (Q[/
)f2f j
WASHINGTON, D. C. 20555 Yf*****
10V 2 0 UiG Docket Nos. 50-354 and 50-355 Mr. R. L. Mittl General Manager - Projects Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101
Dear Mr. Mitti:
SUBJECT:
IMPLEMENTATION OF STAFF REVIEU REQUIREMENTS - HOPE CREEK GENERATING STATION, UNITS 1 & 2 During the last several years we have reviewed and approved several new regulatory guides and branch technical positions or other mocifications 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 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 enanges 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 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 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 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 acceptabl,e alternate,.
then staff-approved design revjs. ions.will be required.
wp a
78120700G1 glg 0000Mhf CONTAll$
A POOR QUAlifV PAGES c
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l Mr. R. L. Mitti l 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. is a list of_the Category 2 matters. is a list of the Category 3 matters.
In addition to the RRRC categories, there also exists an NRR Category 4 l
list which are those matters not yet reviewed by the RRRC, but which j
the Director, NRR, has deemed to have sufficient attributes to warrant j
l their being addressed and considered in ongoing reviews. These matters will be treated like Category 2 matters until such time as they are l
reviewed by the RRRC, and a definite implementation program is developed.
J A current list of Category 4 matters is attached (Enclosure 4).
These
)
also should be considered 'in your FSAR.
l 1
In some: instances the items in the enclosures may not be applicable to I
your application. Also, we recognize that your application may, in some l
instances, already conform to the stated staff positions.
In your FSAR l
you should note such compliance.
If you have any questions please let us know.
i Sincerely,
/
QRoger S.
oyd, Directo Division of Project Ma M nt Office of Nuclear Reactor Regulation
Enclosures:
As stated cc:
See next page i
I l
i e
_ pwT
a Mr. R. L. Mitti General Manager - Project's
- 0V 2 01972 Public Service Electric & Gas Company 80 Park Place, Room 816 MP Newark, New Jersey 07101 cc: Fred Broadfoot, Esq.
Public Service Electric &
Gas Company Assistant General Counsel 80 Park Place Newark, New Jersey 07101 Mr. John Boettger, Project Manager Public Service Electric & Gas Company 80 Park Place Newark, New Jersey 07101 Honorable Mark L. First Deputy Attorney General State of New Jersey Nuclear Energy Council 36 West State Street Trenton, New Jersey 07102 Richard Fryling, Jr., Esq.
Public Service Electric &
Gas Company 80 Park Place Newark, New Jersey 07101 William Horner, Esq.
67 Market Street Salem, New Jersey 08079 Mr. David A. Caccia Box 70 - A. R. D. #2 Sewell, New Jersey C8080 Dr. John K. LaMarsh 68 North Chatsworth Avenue Larchmont, New York 10538 Manager, Quality Assurance Public Service Electric & Gas Company 80 Park Place
, Newark, New Jersey 07101 l
l e
Mr. R. L. Mitti p.,.,, - - j cc: Mr. N. C. Vasuki, ' Director Division of Environmental Control Tatnall Building Dover, Delaware 19901 Robert D. Westreich, Esq.
Assistant Deputy Public Advocate Department of the Public Advocate Division of Public Interest Advocacy Post Office Box 141 Trenton, New Jersey 08625 Mrs. Richard Horner Main. Street Hancocks Bridge, New Jersey 08038 Troy B. Conner, Jr.
Conner, Moore & Corber 1747 Pennsylvania Avenue, N. W.*
Washington, D. C.
20006 F. Michael Parkowski, Esq.
Deputy Attorney General Tatnall Building Dover, Delaware 19901 Bechtel Power Corporat'on ATTH:
Mr. J. Han, Project Engineer 50 Beale Street P. O. Box 3965 San Francisco, California 94119 Edward Luton, Esq., Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. Ernest E. Hill Lawrence Livermore Laboratory University of California P. O. Box 808, L-123 Livermore, California 94550 h_
-s a
.J Mra R, L. Mitti NOV 2 0 g
-cc: Dr. Oscar H'. Paris Atomic Safety.and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Jerome E. Sharfman, Esq., Chairman Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Richard'S. Salzman, Esq..
Atomic Safety and Licensing Appeal Board V. S. Nuclear. Regulatory Commission Washington, D. C. -20555.
Dr. W. ' Reed Johnson Atomic Safety and. Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555
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UNITED ST ATES
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filVCt.E AR, REG LATOR COMMissi.
~ W ASHING TON, O. C. ' 20555 SEP.2 $ WS.
Lee V. GossickL
' Executive Director-for. Operations
- REGULATORY' REQUIREf1ENTS REVIEW. COMMITTEE MEETING N0. 31, JULY ll, 1975 i plementation of 1.
The Committee Wiscussed. issues related to the m
Regulatory Guides on existing : plants and~ the concerns expressed in the June 24, 1974 memorandum, A. Giambusso to E. G. Case, subject: REGULATORY GUIDEflMPLEMENTATION, and made the following
- recomendations ~ and observations:
a.
Approval of new Regulatory Guides and approval of revisions of e'xisting. guides should move. forward expeditiously.in ~ order
'that the provisions ~ of these regulatory guides be:available for use. as soon'as possible.in'on-going or future staff reviews of license ~ applications.
The Committee noted that over the-
'r recent.past, the approval' of proposed regulatory guides 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' To avoid these delays.,
requiring significant staff effort.
the Comnittee concluded that, henceforth, approval of proposed regulatory guides should be uncoupled from the> consideration
~
of their backfit applicability.
The_ implementation section of new regulatory guides should
~
b.
address', in general, only theLapplicability of the' guide to-1 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 theiissuance. date-
- of the guide for comment.
Exceptions to'this general approach will be handled on a case'-by-case basis.
c.'
The regulatory position of each approved proposed guide (or proposed guide revision) will be characterized by the Committee as to its backfitting pot'ential, by placing it in one of three L
l categories:
L Category 1 - Clearly forward fit only.
No further staff consideration of possible backfitting is required.
h
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ENCLOSURE 1-
3 w.
Lee 1V.' Gossick
-3'
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3.
The Comittee reviewed:the proposed Regulatory Guide 1.XX:
INSTRUMENT -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 comments', to the. satisfaction of the. Director, Office of' Standards.
Developnent.
This guide was. characterized by the Comittee as Category 1
-no backfit.
~4..
The.Comittee' re' viewed Proposed Regulatory Guide :1.97:
INSTRUMENTATION:FOR LIGHT WATER COOLED ?!UCLEAR PO!IER PLAf;TS.
TO ASSESS PLAT T C0ilDITIONS DURING: AND FOLLOWIl!G AN ACCIDENT
.and deferred further consideration to' a later meeting' in
- order to permit-incorporation of recent connents by the Division'of: Technical Review.
Edson.G.
ase,. Chairman Regulatory Requirements Review.
. Committee L
l t
h ENCLOSURE.'1 (CONT'D)
September 15, 19/8-t CATEGORY 2 MATTERS Document Number Revision Date Title RG 1.2/
2 1/76 Ultimate iteat Sink for Nuclear Power Plants RG !.b2 1
7/76 Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adso ption Units of Light Water Cooled Ncclear 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
//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' Des-ign Limits and Loading Combinations for Class 1 Plate-and Shell-Type Component Supports.-
(Continued)
ENCLOSURE 2
e CATEGORY 2 MATTERS (CONT'D)
Continued Document Number 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)
Guidelines for Fire Protection for BTP ASB Nuclear Power Plants (See Implementation 9.5-1 1
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)
4 s
September 15, 1978 CATEGORY 3 MATTERS i
Document Number Revision Date Title 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 1.121 0
8/76 Bases for Plugging Degraded PWR 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 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 l.56 1
7/78 Maintenance of Water Purity in Boiling Water Reactors
Attachment:
BTP RSB 5-2 (Draft)
ENCLOSURE 3
1 BRAflCH TECHt11 CAL POSIT!Of1 RSB 5-2 OVERPRESSURIZAT10fi PROTECTION OF PRESSURIZED WATER REACT 0PS WHILE OPERATitlG AT LOW TEMPERATURES A.
Background
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 occurrence
- s 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 power unit and include but 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 toughness requirements for reactor pressure vessels under all conditions.
To assure that the Appendix G limits of the reactor coolant pressure boundary are not exceeded during any anticipatr operational occurrences, Technical Specificatic pressure-tempera'ure limits are provided for operating the plant.
The primary concern of this position is that during startup and shutdown cunoitions 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 es+ablished for protection against brittle fracture.
This inadvertent overpressurization could be generated by any one of a variety c' mal-functions or operator errors.
Many inciderits have occurred in operating plants as described in Reference 1.
Addit ional discussion en the background of this position is contained in Reference 1.
ENCL 3 (CONT)
,r.
3j B..
Branch Position 1.
A system should be designed and installed which will prevent.
exceeding the applicable Technical Specifications and Appendix G i
limitsi 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, particular.ly while the' reactor coolant system is in a water-solid condition.
2; The system must be able to perform its function assuming'any single-l active component failure. _ Analyses using appropriate calculational techniques must be'provided which demonstrate that the.systee will-provide the required pressure relief. capacity assuming the'most i
limiting single active: failure.
The cause for initiation of. the.
event, e.g., operator error, componentalfunction, will not be-considered as the single active failure. The analysis should assume c
the most limi, ting allowable ocerating conditions and systems 1
configuration at the time of the. oostulateo cause of the overoressure event.
All potential'overpressur12ation events must be considered l
when establishina the worst case event.
Some events may-be prevented by protective interlocks'or by locking out pnwer.
iht se events should be reviewed on an individual basis.
If~the
.)
lnterlock/ power lockout is acceptable, it. car, be excluded from the analyser,.provided,the= controls in prevent the event are in the plant. Technical' Specifications.
- 3. 'The system must meet the design requirements of IEEE 279 (see Im'pl emen ta t i or )..The system may.be manually: enabled, however, the electrical instrumentation.and control system must provide.
alarms to alert tha operator to; i
a.
properly enab'.e the system at the correct plant condition during^ coo,1down, r
b.
indicate if a pressure transient is occurring.
2 l
.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.
l b.
A test for val've operability must, as a minimum be conducted I
as specified in the ASME Code Section XI.
c c.
Subsequent to s/ stem, valve, or electronics maintenance, a test on -thut portion (s) of the system riust be performed prior to' l
-declaring.the system operational.
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- ENCL 3-(CONT)
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7-
)
NN
-[
t 5.
The system must' meet the regnirementsw T c quiatoryfGuide 1.26, e
"Ouality' Group Classifications'and Standog: for Water, Steam,-
=and Radioactive-Whste-Containing Components'of. twclepr Power Plants" and Section-111 ofithe ASME Code.
- 6. - The-overpressure protection. system mus t be designed.t'o function y
during an Operating' Basis Earthquake, it must not compromise;the i,
design criter.ia of:any' other safety-grade system with which it.
wouid interface..such that the requirements of Regulatory. Guide l'.29,:" Seismic. Design Classificati,on" 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 acto
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- component (s) to mitigate the consequences of an overpres"
~
. event must inclu'de additional analyses considering ina.
,ystem initiation / actuation or' provide justificatior existing analyses bound such'an event.
1 C -.
Implementation.
The Branch TechnicalfPosition, as specif-mis.o used in the review of ali^ Preliminary Design Apprv.,i
., Fina.1 Design Approval (FDA), Manufacturing License (ML), Operating License (0L), 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' Commission'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,.978 but for which the ' licensing' action has not been 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.
of IEEE 279 will be applicable to their approved standard designs'and that such designs should be modified, as necessary, to con'orm to the position.
Staff approval of proposed modifications can be applied for either by apolication 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, 1978i j
i J
~ ENCL 3 (CONT' P
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'_a_;
Liikf tf l'
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1.
Those; applicants issued'an OL.during the period between March.14,
.)
1978 and a date 12 months thercafter may merely commit to meeting the position prior to 0L" issuance but'shall, by license condition.-
be required'to install.all required staff-approved modifications 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, PDA, or ML during the period.between March'14, 1978 and a date 6 months-thereafter may merely commit to neeting the position but shall, by license condition, be required to amend the application, witnin 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.
l 4.
Those applicants issued -a CP; PDA, or ML after September. 14, 1978 shall have staff approval of proposed modifications prior to issuance of the CP, PDA, or ML.
l 0.
Referenr's, 1.
NUktG-0138, Staff Discussion of Fif teen Technical Issues Listed in Attachment.to N'avember 3, 1976 Memoranduo from_ Director. NRR, to NRR Staff.
l 4
ENCL 3 (CJ:T)
lElk[k[ $
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1.-
Those a'p'plic'an.t's Jissued an OL "during the' period between March 14, 1978 and a'date 12'mont'hs therea.fter,may.merely commit,to meeting the ' position prior,'to 0L'is"sdanc~e'but'shall.by license. condition,
~
- be 'requi' red to install ' l:1 ?equirEd st'aff.-approved modifications a
p'rior to plant sthrtup 'following thb' first scheduled refueling outage.
2.
Those applicant's issued anf0L beyond_ March 14, 1979-shall install
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all required' staff-approv'ed modifications prior to initial' plant
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startup.
3.
Those applicants "issush a CP',;PDA, or ML during the, period,between-March 14, 1978 and.a'date 6 months thereafter may merely commit
~
to neeting the position. but s'nall, by. licens~b condition, be.
required to amin'd.thi Application, within 6 months of the date of issuance of.the.CP,,PDA, or ML, to include a description of the pro' posed modi,ficat. ions and the bases for their. design, and a request for staff approval'.
4.
Tho'se applichnts i'ssued a CP, PDA, or ML af ter September 14, 1978
, shall have staff $pproval of. p'roposed modifications prior to-
~
issuance of the CP, PDA, or ML.
D.
References 1.
NUR$G-0138, Staff Discussi6n of Fifteen Technical Issues Listed in Attachment to November 3' 1976 Memorandum from Director, NRR, to NRR Staff.
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l 1.
1 l
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r L
L ENCL'3 (CONT).
... ~..
e CATEGORY-4'bt4TTERS A. ' Regulatory Guides not categorized Issue.
.Date Number--
' Revision Title 4/74 1.12' l
' 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 fcr Nuclear Power.
-Plants 9/75 1.79-1 Preoperational Testing of Emergency Core Cooling Systems.for Pressurized Water Reactors 6/74
'l.80' O
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 Sou'rces 2/76 1.104 0
Overhead Crane Handling Systems for i
Nuclear Power Plants 2
ENCLOSURE ~4
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s
B.
SRP Criteria Impl ementa-Applicable' tion Date Branch SRP Section Title 1.
1.1/24/75 NTEB 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.lA Containment Pressure Model 6.2.18 for PWR ECCS Performance 6.2.1.2 Evaluation l
6.2.1.3 6.2.1.4 6.2.1.5 l
3.
11/24/75 CSB 6.2.5 BTP CSB-6-2, Controlicf Combustible Gas Concentra-tions in Containment Following.
a Loss-of-Coolant Accident 4.
11/24/75 CSB 6.2.3 DTP CSB-6-3, Determination of lyrass Leakage Path in Dual 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 Section3.5.3, par.II.l.C)
'ENCLO!sURE 4 (CONT)
{
l
t Impl ementa-
. Applicable' tion Date-Branch SRP Section Title 9.
11/24/75 SEB 3.7.1
' Development of Design Time
. History for 3 oil-Structure Interaction Analysis (SRP Section 3.7.1, par. II.2) 10.~ 11/24/75.
SEB 3.7.2 Procedures.for Seismic System Analysis (SRP Section 3.7.2 par. II) 11.
11/24/75 SEB 3.7.3 Procedures for Seismic Sub -
system Analysis (SRP Section 3.7.3.-
par._II)-
12.
11/24/75 SEB 3.8.1 Design'and Construction of Concrete Contaimnents) SRP Section.3.8.1, par. II).
- 13. 11/24/75 SEB
'3.8.2 Design and Construction of Steel' Containments (SRP Section
'3.8.2, par. II) 14.
11/24/75
'SEB-3.8.3 StructurL1 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 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)
F.
(
t r
F ENCLO9URE 4 (CONT)
. Implementa-Applicable tion Date Branch SRP Section 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. II.4) and R.G. 1.45 23.
11/24/75 RSB 3.2.2 Main Steam Isolation Valve Leakage Control System (SRP Section 10.3 par. III.3 and BTP RSB-3.2)
C.
Other Positions Impl ement a-Applicable tion Date Branch SRP Section Title 1.
12/1/76 SEB 3.5.3 Ductility of Reinforced Concrete and Steel Structural Elements Suojected 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)
Impl ementa-Applicable-tion'Date Branch SRP Section Titl e 7.
9/01/77 RSB-6.3 Control Room Position Indica-tion of Manual (Handwheel) Valves in the ECCS 8.
4/01/77
'RSB 15.1.5 Long-Term Recovety 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
' Centainment. 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 j
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 Hammer in Steam Generators with Top Feedring Design (BTP ASB-10.2) '
19, 1/01/76 (CSB 3.11-Environment'al Control Systems f;r Safety-Related Equipment l
l
' ENCLOSURE 4 (CONT)
,em 2
- =
k s
1 DESCRIPTION-0F POSITIONS IDENTIFIED AS NRR CATEGORY 4 MATTERS IN ENCLOSURE 4, PARAGRAPH C Numbering scheme corresponds to that used in Item C of Enclosure 4.
L 6
ENCLOSURE 4 (CONT)
4 C.1 DUCTILITY OF REINFORCED CONCRETE AND STEEL STRUCTURAL ELEMENTS SUBJECTED T0 IMPACTIVE OR IMPULSIVE LOADS INTRODUCTION in the evaluation of. overall response.of reinforced-concrete structural elements (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 summarizes specific SEB interim positions for review and acceptance of. ductility ratias for reinforced concrete an.1 steel structural elements subjected to impactive and impulsive loads.
SPECIFIC POSITIONS 1.
REINFORCED CONCRETE MEMBERS 1.1 For beams, slabs, and walls where flexure controls design, the permissible ductility ratio ( u ) under impactive and impulsive loads should be taken as 0.05 for p-p'
_.005
=
u p
p' 10 for p-p' 5.
005
=
u where p and p'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.,
p> 100) l 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.
I
(
ENCLOSURE 4 (CONT)
t'
, 1.3 ' For beam-colunns, walls, and slabs carrying axial compression
-loads and subject to impulsive' or impactive loads producing flexure, the permissible ductility ratio in flexure should be as follows:
(a)
When compression 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 structural 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 concrete only p = 1.0 For shear carried by concrete and stirrups or bent bars
= 1.3 y
For shear carried entirely by stirrups y
= 3.0 2.0 STRUCTUR AL STEEL MEMBERS 2.1 For flexure compression and shear p
= 10.0 2.2 For columns with slenderness ratio (1/r) equal to or less than 20 p
= 1.3 l
ENCLOSURE 4 (CONT)
P(wir9 N
.u
\\ k =tHiT _t s O*loh
\\
L s
N I
e og I'. 9y
- y ko ael 4
E Nt
?
?
T Q, 2 2 v i $[
= on
= L Y
i
( '_.. _
ps
.I f3 E ' oft to b
2 9
o to[j A 8
whichevent I
e c
is the smaller g
g [pg.ky to g,3 socnity p.
a 5
FtG 1.
Po.oPosee tocT uTy RATIO UfoR BGA1311M'if, zb l
'-r-
+v.v-.w-ey
- ' ~ '
~mr t
w vwev,-
~ i
-where11' = ef fective.. length' of the member :.
I r = the least radius of. gyration.
For columns with slenderness ratio greater than 20 p'= 1.0 2.3 For members subjected to tension u =.5 cy 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 -
4 vertical design design response spectrum as 2/3 the horizontalfresponse spectrum over the entire range.of frequencies will be accepted.
i For other locations, the vertical re<.,ponse spectrum will be the same as that given in Regulatory. Guide 1.60.
~
C.3 BWR MARK III CONTAINMENT POOL DYNAMICS 1.
POOL SWELL-a.
Bubble pressure, bulk swell and froth swell loads, drag pressure and other pool swell loads should be treated as abnormal pressure loads, P. Appropriate load combinations a
and load f actors should be applied accordingly.
b.
The pool swell loads and accident pressure may be combined in accordance with their actual time histories of occurrence.
L ENCLOSURE 4..(CONT)
?
... ~.
~l 4--
'I 1
2.
SAFETY RELIEF V ALVE.. (SRV) DISCHARGE-
.'a.
The SRVLloads should-be treated as ' live loads in all load s combinations 'l.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 consider'ed'in combination'with=the Design Basis
. Accident (DBA)..
.c.
Appropriate multiple SRV ' discharge should be considered in
~
comt
, tion ~with the Small Break Accident (SBA) and Inter-mediate Break Accident (IBA)..
d.-
Thermal-loads due to-SRV discharge should be treated as TO
'for normal operation and T, for accident conditions.
- e..Thb 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 strength, buckling and low cycle fatigue.
t 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.
.l. ' 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 U
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 1
where, U is the strength capacity of a sectinn 0 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, 'a.nd also for rein-J forced concrete elements in existing plants, some inelastic response may be permitted with appropriate limits on ductility ratios.
l
{
i ENCLOSURC 4 (CONT)
R
- 4. ' Air' blast' generated ground' shock' and air; blast wind pressure may be ignored. 7 Air blast generated missiles may be important in situations where explosions' are postulated to-occur in vessels
-which may fragment.'
5.
0verturning 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. L 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 their1 ability to carry loads applied directly to exterior panels and slabs. Moreover.in-vented struct'ures, interior structures may require analysis even if
'they do not support. exterior structures.
'7.
The equivalent static pressure should be considered as potentially.
acting both inward and' outward.
C.5 TGRNADO MISSILE PROTECTION' As an interim measure,the minimum concrete wall and roof thickness ifor 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-must 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.
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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 LOCA should be con-f 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 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 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 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 ECCS train and that the leak can be isolated; and (5) Alarms that conform with the criteria specified for the control room 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.
1 C.7 CONTROL ROOM POSITION INDICATION OF MANUAL (HANDWHEEL) VALVES Regulatory Guide 1.47 specifies automatic position indication of each bypass or deliberately induced inoperable condition if the following three conditions are met:
(1) The bypass or inoperable condition affects a system that is designed to perfonn an automatic safety function.
1 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 insperable condition is expected to occur when the system is normally required to operate.
Revision one of the Standard Review Plan in Section 6.3 requires confonnance 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 room.
In the PDA extension-reviews it is important to confirm 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 OVERPRESSUR;ZATION (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 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 (i) no additional fuel failures result from the accident, and (11) the pressures following the initiation of the break will not compromise ~the integrity of the reactor coolant 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 scme types of engineered safety feature pumps would entinue to perform their safety function in the long term following an accident.
In such instances there has been followup, including pump redesign in some cases, to assure that long term performance could be met. The following kinds of infor-mation may be sought on a case-by-case basis where such doubt arises, a.
Describe the tests performed to demons' rate 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-' water cooled seals and compo-nents in.the pumps.
j c.
Provide a description of the composition of the pump shaft i
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 MISS'ILES,-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 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 affcrd 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.
~
L ENCLOSURE 4'(CONT)
[
1
, l l
.C.ll CORE' 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 bowing.
l In addition,a commitment to perform tests to verify the transient analysis methods and codes is required.
l 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 power systems. These additional l
requirements are defined in the following staff position.
1.
N require that a second level of voltage protection for the onsite power system be provided and that this second level of voltage. pro-l tection satisfy. the. following requirements:
a) The selection of voltage and time set points shall be determined from an analysis of the voltage requirements of the safety-related loads at all onsite system distribution l
levels;
'b) The voltage protection shall include coincidence logic to preclude spurious trips of the offsite power source; i
i e
r i
ENCLOSURE 4 (CONT)
)
,~
4 s
10--
'c) The time delay' selected shall be based.'on the~ following.
s conditions:
(1)
The allowable time delay,' including margin, shall not exceed the maximum time delay that is. assumed in 2
the SAR accident analyses; (ii)1 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
.stribution system levels shall not l
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; i
(v)
The~ voltage sensors shall be designed.to satisfy the o
applicable requirements of'IEEE Std. 279-1971 " Criteria l
for Protection Systems for Nuclear Power Generating l
. Stations"; and (vi) The Technical Specifications shall. include-limiting conditions for operation, surveillance requirements, L trip set points'with minimum-and maximum Iimits, and allowable values for'.the second-level' voltage protection' l
sensors and associated time; delay devices.
2.
We require that the system' design automatically prevent load shedding of the emergency buses once the onsite sources are supplying. power to all sequenced loads on the emergency buses.
The design shall also' include the capability of the load' shedding:
l
' 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 i
of onsite power sources 'to their respective buses.-
i b
ENCL 6SURE 4. (CONT)-
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ll y
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- L.
L:
..,_L._____
. I
-4._- The' voltage' levels at the safety-rel5ted bus'es 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 transfnrmers.
We require that1the adequacy of the design in this-regard be, a
verified by actual;measurtnent, and by cor. relation-of measured
- values; with analysis. results..
l 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 igeneric Category ALTask: Action Plan A-2 is designed to develop
. generic resolutions for.this matter. 0ur present schedule cills for completing A-2, for PWR's during the first quarter,1979. Penuing completion 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 commit to address the safety issue. as part of their appli-cation for an operating license.
2.
For PWRs at the OL/FDA stage of review, the. staff requires case-by-case analyses, including implementation of any indicated corrective -
measusr 3 prior to the issuance of an operating' license.
'3.
For BWRs, for which this issue is-expected to be of lesser safety significance, the asymmetric 1oading conditions will be evaluated on a case-specific basis prior to the _ issuance of an operating license.
For those cases which analyses.are required, fwe request the-performance 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 pressur.lation. Show how the results of these analyses are used in the design of structures and component supports.
ENCLOSURE 4 (CONT) c
- -12--
C'.14
- CONTAINMENT LEAK TESTING PROGRAM' I
I lTo ' 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.1 ' For this purpose, the.following information.with regard to the containment;1eak' testing program should be; supplied.
- a. jThose systems' that will remain fluid filled for the Type. A test -
'should be'. identified andLjustification given; b.
Show the design provisions that will permit the personnel ' air.
lock doorLseals and the entire air lock to be tested, c.
For each penetration,1~.e.', fluid system' piping, instrument,.
~
electrical,-and equipment and personnel access penerations, identify 1the Type B and/or, Type C local' leak testing that will be done.
t 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 T0. MAIN S':Aci LINE BREAK AND MSLIV FAILURE In recent CP and OL application reviews,:the results' of:
analyses 'for a.postu'.ated main steam line breck accident (MSLB)'.
for designs' utilizing pressurized water reactors with conventionali 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 I
signifiance of the matter is that.it could-result in a requirement for. req'ualifying 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 presently scheduled completion dates for A-21 and A-24 (Short Term Portion).are first quarter,1979 and fourth quarter,1978, respectively.
.Pending: completion of A-21 and-A-24, some interim guidance'will be used as det' ailed'below.'.
l l
We have developed and are implementing a plan in which all applicants for construction permits and operating licenses and those' already issued. con-struction permits -must' provide information to establish a conservative L
-temperature-time envelepe.
ENCLOSURE 4 (CONT) l
,b A'.
x
4
+
._13..
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; 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 cffsite power is evaluated;'and partial loss
~
'of containment cooling rystems.
(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 i
may be superheated)was used, (5) Discuss 'and' justify the. analytical model including the themodynamic 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 'areat and power levels analyzed; (7) For the case which results in the maximum containment atnesphere 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 shutdown; ENCLOSURE 4 (CONT)
E li 2
-14
' (8)' Forcthe case which results in maximum centainment 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 form.
In order to demonstrate that safety-related equipment has been adequately qualified as described'above, provide the following information regard--
ing its environmental qualification.
(1) Provide a comprehensive list of equipment required to be operational in the. event of a main steamline break (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; i
.(d) Electrical power cables;
'(e) Electrical instrumentation cable; and I
(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 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 blowdown period it is appropriate to use a conservatively evaluated forced convection heat transfer correlation. For example, ENCLOSURE 4 (CONT)
g m
-n
-f,
-15$ SNk I M%{
Na = C(Re)-
b e ;7 p -=
Where Nu = Nusselt No..
q,
.._4 b Nb"
=
Re = Reynolds No.
C
= empirical constants dependent on uso m a m i
geometry and Reynolds. No.
y p g i
Since the Reynolds number is dependent on. velocity, it is R
necessary to evaluate the forced flow currents which will be b m h !
generated by the steam 9eneraor blowdown. The CVTR experiments
@ l[hli provide limited data in this regard. : Convective currents of g
u
, E from 10 ft/sec to 30 ft/sec were measured locally. We recomend
- N cbtain forced flow currents to detemine the convective heat l fL.
E
'3 that the CVTR test results be extrapolated conservatively to
[
transfer.coef ficient during the blowdown period. After the L
l blowdown has ceased or been reduced to a negligibly low value, y
I a natural convection heat transfer correlation is acceptable.
{
(3) For each component where themal analysis is done in conjunction l
' M[
with an environmental test at a temperature lower than the peak 4
calculated temperature following a main steam line break accident compare the test thermal response of the component with the accident I
thermal analysis of the component. Provide the basis by which the J
component thermal response was developed from the environmental l
qualification test program.
For instance, graphically show the themocouple data and discuss the themocouple locations, method j
of attachment, and performance characteristics, or provide a f L 3
detailed discussion of the analytical model used to evaluate the a
l component thermal response during the test. This evaluation should I
i cross-sections and temperature sensitive parts where themal stressing,
$,$.}f; 1
be performed for the potential points of failure such as thin temperature-related degradation, steam or chemical interaction at g;;,MU' elevated temperatures, or other thermal effects could result in the
@y;f failure of the component mechanically or electrically.
If the component thermal response comparison results in the prediction of v
a more severe thermal transient for the accident conditions than for the qualification test, provide justification that the affected y'
component will perf; <m its intended function during a MSLB accident, or provide protectico for the component whch would appropriately c
limit the thermal eff ects.
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4;
. k)jk) vgend{f Hil_E m
rh e r, im 1
M l
r m=m ENCLOSURE 4 (CONT) g mm@ip
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- ..q=
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- .g, p.g.ja,.g j j,
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. C.16 ENVIRONMENTA'. EFFECT OF PIPE FAILURES Identify the " break exclusion" regions of the main steam Compartments that contain break and ftedwater lines.
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 ' break excluded' 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 transierts and accidents or that enough time and information are availahla to per-it corrective action by an operator.
We have established the following criteria for that portion of the component cooling water (CCW) system which interf aces 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 f ailure 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 Single active f ailures of cooling to one or more pumps.
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 A single active f alure shall be considered when pumps occurs.
evaluating the consequences of this accident. Moderate leakage cracks should be determined in accordance with Branch Technical Position ASB 3-1.
In order to meet the criteria established above, an NSSS inter-tace 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 whis supplies cooling water to the reactor coolant pumps and notors may be designed to non-seismic Category I requirements and Quality
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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-tective action.
In addition, safety grade instrumentation including alarms should be provided to detect the loss, of compor 'nt cooling water to the reactor coolant pumps and s
motort, 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 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 Ca tegory I require-ments and Quality Group D; or 2.
The component cooling water supply to the pumps and mocors shall be capable of withstanding a single active f ailure 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.
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The reactor coolant (RC) pumps and motors are within the NSSS scope of design. There fore, in order to demonstrate that an RC pump design can operate with loss of component cooling water for at least 30 minutes without 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 cooling water does not result in a LOCA due to seal failure.
ENCLOSURE 4 (CONT) 9
, 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
.oastdown 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 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 criteria and standards.
e.
Information to verify the applicability of the equations and material properties chosen for the analysis (i.e.,
references should be listed, and if empirical relations c
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 terminated for a specified period of time to verify the analysis.
C.18 WATER HAMMER IN F AM 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 uncove-ing 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.
1:NCLOSU RE 4 (CONT)
..,. For CP/PDA and OL/FDA applications, provide the following for steam generators utilizing top feed:
Grevent or delay water draining from the feedring following a 1.
drop dn steam generator water level by means such as.y-Tubes; Minimize the volume of feedwater piping external to the steam 2.
generator whch could pocket steam using the shortest possible (less than si ten feet) horizontal run of inlet piping to the steam generator feedring; and Perform tests acceptable to the staff to verify that unacceptable feed-3.
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 possible draining of the feedring. Provide the procedures for these tests for staff approval before conducting the tests.
Furthermore, we request that the following be provided:
Describe normal operating occurrences of transients that a.
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 strJcture up to the outer isolation valve and restraint, Describe any analysis on the piping system including any c.
forcing functions that will be performed 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 f acility would not result with your design.
t ENCLOSURE 4 (CONT)
I C.19.INVIRONMENTAL CONTROL SYSTEMS FOR SAFETY RELATED E0tilPMENT Host 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 environmentai qualification of the safety related equipment installed in those areas. It appears that there are no requireme'nts for maintaining these environmental control systems in operation while the plant is shutdown or in ho: 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 exc'ed those for which safety related equipment is qualified shall bc trovided. 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 sources, and (4) provide a continuous record of the environmental parameters during the time the environmental conditions exceed the normal limits.
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ENCLOSURE 4 (CONT)
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