ML19283B711
| ML19283B711 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/02/1979 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Gary R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| References | |
| NUDOCS 7903060657 | |
| Download: ML19283B711 (44) | |
Text
m TERo f
a, UNITED STATES f
y ' *;
NUCLEAR REGULATORY COMMisslON 3
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j WASHINGTON, D. C. 20555 q ~.uk/ i I
FEB 2 1979 Docket Nos.: 50-446 50-446 Mr. R. J. Gary Executive Vice President and General Manager Texas Utilities Generating Company 2001 Bryan Towers Dallas, Texas 75201
Dear Mr. Gary:
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2 Enclosed are requests for additional information which we require to complete our evaluation of your application for an operating license for Comanche Peak. The infomation requested is in addition to that requested in our letters of November 2 and 22, 1978 and covers the review performed by the Analysis Branch. We have also included a request for information regarding regulatory guides and branch tech-nical positions or other modifications to existing staff positions which have been reviewed and approved by our Regulatory Requirements Review Comittee (RRRC) and/or the Director, Office of Nuclear Reactor Regulation for consideration in the licensing review of all nuclear power plants. Please amend your FSAR to include the infomation requested.
Your schedule for responding to the enclosed requests for additional infomation should be submitted by March 1,1979. Based on your schedule for response and our workload, we will detemine any licensing review schedule adjustments and infom you of any significant changes.
Please contact us if you desire any discussion or clarification of the enclosed requests.
Si erely, Q d(
' Steven A. Varga, Chie Light Water Reactors Branch No. 4 Division of Project anagement
Enclosures:
As stated cc: See next page 79030666F7
Texas Utilities Generating Company Ccs:
Nicholas T. Reynolds, Esq.
Debevoise & Liberman 1200 Seventeenth Street Washingtcn, D.C.
20036 Spencer C. Relyea, Esq.
Worsham, Forsythe & Sampels 2001 Bryan Tower Dallas, Texas 75201 Mr. Homer C. Schmidt Project Manager - Nuclear Plants Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Pr. H. R. Rock Gibbs and Hill, Inc.
393 Seventh Avenue New York, New York 10001 Mr. G. L. Hohmann Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 Richard W. Lowerre, Esq.
Office of the Attorney General P. O. Box 12548 Austin, Texas 78711 Ms. Nancy Holdam Jacobson Coordinator, CFUR 1400 Henphill Street Forth Worth, Texas 76104 Mrs. Juanita Ellis, President CASE P. O. Box 4123 Dallas, Texas 75208
a t
ENCLOSURE REQUEST FOR ADDITIONAL INFOPMATICN CC %NCHE PEAX - U"ITS 1f.2 These requests for additional information are numcered such that the three digits to the left of the decimal identify tne tecnnicli review branch and the numoers to the right of the decimal are the sequential request numbers. The numoer in parentnesis indicates the relevant section in the Safety Analysis Recort. The initials RSP indicate the request represents a regulatory staff position.
Branch Technical Positions referenced in :nese recuests can be found in " Standard Review Plan for the Review of Safety Analysis Reports for ::uclear Power Plants," NUREG-75/087 datec September 1975.
~
22T.0 Reactor Analysis Section, Analysis Branch 221.1 The effects of fuel rod bowing must be included in the thermal-hydraulic design. The predicted extent of rod bow (gap closure) versus exposure and the effect of rod bowing on DNBR must be addressed. Use of the staff report " Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors", February 16, 1977 represents an acceptably conservative treatment of rod bowing.
221.2 If an unreviewed themal-hydraulic method such as the "THINC Equivalent" method has been used in any part of the thermal hydraulic designor if such a code will be used to establish the reactor protection system trip setpoints, then the following infomation is required relative to that method:
1.
A complete detailed description of the method including all the equations, boundary conditions and assumptions.
2.
Comparisons between THINC-IV and the method used to generate these data. These comparisons should cover the range of parameters (including axial power distribution) encountered in the trip system.
221.3 Infomation presented in Section 4.4 is not sufficient to demonstrate the required themal hydraulic stability of the core.
Comanche Peak 1 & 2 used the HYDNA code to analyze the flow stability characteristics of the core and references WCAP-7240 (P) and WCAP-7966 to describe the effect of open channel flow on themal hydraulic flow instability. However, the referenced reports do not describe the HYDNA code and its use in the thermal hydraulic design and these reports have been withdrawn by Westinghouse.
The reports provided experimental data intended to show that simu-lated fuel assemblies without an enclosing shroud will provide a larger stability margin than would fuel assemblies with an enclosing shroud. While such data are useful as background information, they are not sufficient to support a conclusion that the HYDNA code conser-vatively preaicts the onset of flow instability in the core. To support such a conclusion, we require that the following infomation be provided in the FSAR or that the FSAR include a coranitment to a topical report containing this infomation.
1.
Provide a complete description of the HYDNA code and its use in the analysis and specific results for the Comanche Peak design or,
. 2.
Provide a discussion, excluding the HYDNA code, wnich supports the contention that the core is themal hydraulically stable.
This discussion must be independent of the HYDNA calculations if approval of the themal hydraulic design is not to depend on the approval of the HYDNA code. State your intent with regard to this matter.
2 21.4 Operating experience on two pressurized water reactors (not of the Westinghouse design) indicate that significant reduction in core flow rate can occur over a relatively short period of time as a result of crud deposition on the fuel rods.
In establishing the Technical Specifications for Comanche Peak 1 & 2, we will require provisions to assare that the minimum design flow rates are not exceeded. Therefore, provide a description of the flow measurements capability for Comanche Peak 1 & 2 as well as a description of the procedures to measure flow and the actions to be taken in the event of an indication of lower than design flow.
221.5 The NRC approval of the THINC-IV code, for use in :ne tnemal hydraulic design, indicates that the pressure g~dient at the core exit must be modeled. Provide a revised THINC-IV calculation at the steady state reactor design conditions including the modeling of the core exit radial pressure gradient. Provide the following specific information from that calculation:
1.
minimum DNB ratio (value and location) 2.
hot channel flow vs axial position 3.
hot channel enthalpy vs axial position 4.
hot channel quality vs axial position 5.
hot channel void fraction vs axial position 6.
the assumed core exit pressure gradient.
221.6 Insufficient infomation has been provided to justify the design power level of 2389 Mwt (70". of full power) during three-loop operation. Temperature differences in the active cold legs of a few degrees could exist during three-loop operation. Therefore a radial power tilt and an increase in enthalpy rise factor could result. As a result, we request that a complete detailed description of the following items be provided:
1.
The method of detemining the temperature distribution among the cold legs and the associated radial power tilt:
. 2.
The method of accounting for differences (if any) in the three-loop thermal hydraulic design; 3.
The instrumentation available and monitoring procedures during three-loop operation; 4.
The ONBR Technical Specifications and how it will be imple-mented for three-loop operation; 5.
The reactor protective system setpoints related to DNBR protection and how they are generated.
6.
The effects of anticipated operational occurrences on the
- old leg temperature distributions and how this effect is included in the design.
7.
A thermal hydraulic design comparison table for three-loop operation.
. 222.0 SYSTEMS ANALYSIS SECTION, ANALYSIS BRANCH 222.1 For calculation of mass and energy releases for containment (6.2.1) sub-compartment analyses, WCAP-8312 is referenced. This report describes use of the SATAN-V code with the Modified Zaloudeck correlation for subcooled critical flow and the Moody correla-tion for saturated critical flow. A 10% margin is added to the resulting mass and energy release data for conservatism. On page 6.2-14 you state that this conservative margin has been removed for your subcompartment analyses. Justify removal of this margin.
222.2 Since pipe restraints are provided for large primary system (6.2.1) lines that cenetrate the subcompartment walls, limited off set type breaks were analyzed. For breaks of this type the break geometry may resemble an crifice in the broken pipe.
Data by a number of investigators has demonstrated that for two-phase flow the mass flow rate per unit area for orifices is higher than fpr pipes. Justify that the SATAN-V methods are conservative for perediction of flow through orifices. Orifice flow data is found in (1) NED0-13418, " Critical Flow of Saturated and Subcooled Water at High Pressure," by Sozzi and Sutherland, July 1975, (2) " Blowdown Flow Rates of Initially Saturated Water",
by v. Simon, Topical Meeting on Water-Reactor Safety, Salt Lake City, Utah, March, 1973, and (3) " Choked Expansion of Subcooled Water and the I.H.E. Flow Model", by R. L. Collins, Journal of Heat Transfer, May 1978.
222.3 Provide a comparison of the initial steam generator water and (6.2.1) steam inventories assumed by Westinghouse in WCAP-8860 with those of the Comanche Peak steam generators.
222.4 Following rupture of a main steam line, flow through the feedwater (6.2.1) pumps will increase as a result of the decrease in discharge pressure. Decreased feedwater pressure will close the check value leading to the intact unit causing all feedwater to flow into the unit with the broken steam pipe.
Provide an analysis of the feedwater flow transient for Comanche Peak following a main steam line break and provide a comparison to the feedwater flow rates assumed by Westinghouse in WCAP-8860.
222.5 Describe any differences between the steam line break analysis methods and rerJlts reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases", with those reported in the Comanche Peak Steam Electric Station FSAR.
~
5 222.6 Describe any differences between the feedline break analysis methods and results reported in WCAP-9220, " Report on the Consequences of a postulated Main Feedline Rupture", with those reported in rpesent FSAR.
222.7 Describe in detail the method useo to analyze the consequences of the steam generator tube rupture accident including all computer programs.
222.8 The transient analysis methods used to evaluate the various transients and accidents are presented in Section 15.
In this regard, we are conducting a review of the '_OFTRAN computer program including its experimental verification. The amount of applicable experimental data available at this time is limited, and additional data is needed to complete this review. To obtain this information, you are requested to provice an outline for a series of carefully palnned transients tests including a discussion of instrumentation requirements. The type of tests to be performed should include the following:
1.
Turbine trip 2.
Loss of Feedwater Flow 3.
Complete loss of primary system flow 4.
Partial loss of primary system flow Test results from similar facilities can be applied when applicable. The transients tests can be performed at less than full power. However, the tests should be sufficiently severe to provide a reasonable test of the analysis methods.
222.9 To facilitate staff audit e.alculations of postulated transients for the Comanche Peak facility, provide the following information:
1.
A detailed descriptton of the steam generator model including the following:
a.
Various heights and diameters inside steam generator, b.
Initial downcomer level (from normal water level to bottom) c.
Tube bundle flow area d.
Riser flow area e.
Tube bundle height f.
Riser height
g.
Drum steam volume h.
Effective separator flow area
- i. Quality above which there is no heat transfer J.
Initial circulation ratio k.
Downcomer, tube bundle, and steam separator shock 1.
Downcomer, tube bundle and steam separator hydraulic diameters m.
Effective height of steam separators 2.
Baron concentration in refueling water tank.
3.
Temperature in safety injection tank.
4.
Volume to be swept out in the safety injection '.ine cefore boron enters the reactor coolant system.
5.
Baron concentraticn in boric acid tank.
6.
Time delay from safety injection actuation signal until trip breakers start to open.
7.
Spacer loss coefficients for fuel bundles.
8.
Hydraulic diameter and heated parameters for fuel channels.
r-Reactor vessel upper plenum volume.
- 10. Reactor vessel lower plenum volume.
- 11. Primary system hot leg volume (each).
- 12. Primary system cold leg volume including crossover (each).
- 13. Steam generator hot plenum volume (each).
- 14. Steam generator cold plenum volume.
- 15. Steam generator U-tube volume (each).
- 16. Table of coolant density versus reactivity.
- 17. Table of radial peaking factors versus core volume fraction.
-7
- 18. Fuel gap thermal conductivity (average).
- 19. Pump flow coastdown curves from full power and hot zero power conditions (asume loss of offsite power).
- 20. Fraction of energy liberated in fuel.
- 21. Fraction of Energy liberated in clad.
- 22. Fraction of energy liberated in moderator.
- 23. Volume cf core coolant channels.
- 24. Steam generator level set points.
- 25. Feedwater ramp down rate after trip.
- 26. Maximum allowable feedwater flow.
- 27. Feedwater control value response characteristic (fraction /second).
- 28. Feedwater enthalpy versus time (pumps tripped).
- 29. Rate of feedwater isolation value closure.
- 30. Maximum area of single feedwater isolation value.
- 31. Auxilary feedwater flow and enthalpy as a function of time.
- 32. Provide the heat capacity and heat transfer area for the primary system metal in contact with the primary coolant.
In addition, a representative metal thickness and the type of metals included in the heat capacity should be proviced for each structure. This information should be proviced for the following regions:
a.
Pressurizer b.
Hot leg piping c.
Steam generator inlet plenum plus upstream half of steam generator tub 6s.
d.
Steam generator outlet plenum plus downstream half of steam generator tubes.
e.
Cold leg piping f.
Reactor vessel inlet plenum g.
Core region h.
Reactor vessel outlet plenum
8-1.
Reactor vessel upper head region
- 33. The diameter and length of the piping from the accumulator to the reactor coolant system.
In addition, what gravity head is assumed for the accumulator system in the analysis.
- 34. The setpoints and flow rates (saturated steam) of the steam generator power operated relief valves and the steam gen-erator safety valves.
9 400.0 Project Management 400.3 The Regulatory Requirements Review Comittee (RRRC), as approved by the Director, NRR, has determined that certain regulatory guides are to be applicable to all nuclear power plants. Therefore, con-formance with these guides or an acceptable alternative is required.
Appendices l AN and 1AB of your FSAR do not address some of these guides as identified in the following table.
Applicable Regulatory Guide Revision Subject R.G. 1.99 Revision 1 Effects of Residual Elements on Predicted (4/77)
Radiation Damage to Reactor Vessel Materials (Paragraphs C.1 and C.2)
R.G. 1.127 Revision 1 Inspection of Water Control Structures (3/78)
Associated with Nuclear Power Plants (all).
R.G. 1.97 Revision 1 Instrumentation for Light Water Cooled (8/77)
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).
R.G. 1.68.2 Revision 1 Initial Startup Test Program to Demonstrate (7/78)
Remote Shutdown Capability for Water-Cooled Nuclear Power Plants. (all).
Revise FSAR Appendices lAN and 1AB to include a discussion of how Comanche Peak Units 1 and 2 conform to the appropriate positions of these regulatory guides either by adherence to the positions in the guide or to an acceptable alternative.
The RRRC applies a categorization nomenclature to each of its actions.
(A copy of the summary of RRRC Meeting No. 31 concerning this catego-rization is attached as enclosure A). Category 1 matters are those to be applied to applications in accordance with the implementation section of the regulatory guide. The table above reflects those guides which the RRRC has classified as Category 3 which is defined as follows:
Category 3: Conformance or an acceptable alternative is required.
If you do not conform, or do not have an acceptable alternate, then staff-approved design revisions will be required.
We have also included as enclosure B a list of those matters which the RRRC has determined to be Category 2 which is defined as follows:
Category 2: A new position whose applicability is to be determined on a case-by-case basis.
. In addition to the RRRC categories, there also exists an flRR Category 4 list which are those matters not yet reviewed by the RRRC, but which the Director, ilRR, has deemed to have sufficient attributes to warrant their being addressed and considered in ongoing reviews. These matters will be treated like Category 2 matters until such time as they are reviewed by the RRRC, and a definite implementation program is developed.
A current list of Category 4 matters is attached (Enclosure C).
It is intended that the Category 2 and 4 matters be addressed during the technical review of the appropriate sections of your FSAR. To assist in our review o' these areas, you should revise your FSAR to include a description of the extent to wnich your design conforms, or of an acceptable alternative, or of why conformance is not necessary.
In some instances the items in the enclosures 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.
Enclosure A UNITED STATES NUCLEAR REGULATORY Cf. MM t ss.
W ASHiNGToN. D.
C.
20$$$
SEP 2 4 375 Lee V. Gossick Executive Director for Operations REGULATORY REQUIREl'ENTS REVIEW COMMITTEE MEETING NO. 31, JULY 11, 1975 1.
The Committee discussed issues related to the imolementation of Regulatory Guides on existing plants and the concerns expressed in the June 24, 1974 memorandum, A. Giamousso to E. G. Case, subject: REGULATORY GUIDE IMPLEME"TATION, anc made the following recommendations and observations:
a.
Approval of new Regulatory Guides and approval of revisions of existing 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 thu 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 requiring signific?nt staff effort.
To avoid these delays, the Comnittee concluded that, henceforth, approval of proposed regulatory guides should be uncoupled from the consideration of their backfit applicability.
b.
The implementation secticn of new regulatory guides should address, in general, only the applicability of the guide to applications in the licensing review process using, in so far as possible, a standard approach of apolying the guide to those applications docketed 8 months after the issuance date of the guide for comment.
Exceptions to this general approach will be handled on a case-by-case basis.
The regulatory position of each approved proposed guide (or c.
proposed guide revision) will be characterized by the Committee as to its backfitting pot'ential, by placing it in one of three categories :
Catecorv 1 - Clearly forward fit only.
No further staff consiceration of possible backfitting is required.
Lee V. Gessick Category 2 - Further staff consideration of the need for back-fitting apoears to be required for certain 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 detemine the need for backfitting.
Categorv 3 - Clearly backfit.
Existing plants should be evaluated to detemine wnether 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 Connittee categorization apcear in the guide itself.
The purpose of the categorization is to indicate thost itens of the rog;:latory position for which the Committee can make a specific backfit recormiandation without additional staff work (Categories 1 and 3), and to indicate those items for which additional staff work is required in order to detemine backfit considerations (Category 2).
e.
The Comittee recommends that for accroved guides in Category 2, staff efforts be initiated in parallel with the process leading to publication of the guide in order that specific backfit requiremer ts for existing plants be detemined within a reasonable period of time after publication of the guide.
f.
The Committee observed that moi e attentien.eeds to be given to the identification of acceptable alternatives to the positions outlined in the guides in order to provide additional options and flexibility to applicants and licensees, with the possible benefits of additional innovation and exploration in the solution of safety issues.
2.
The Committee reviewed the proposed Regulatory Guide 1.XX: THERMAL OVERLOAD PROTECTION FOR MOTORS Ot' t'0 TOR-OPERATED VALVES and recommended approval.
This guide was characterized by the Cormiittee as Category 1 -
-l backfitting, with the stipulation that as an appropriate occrion presented itself in conjunction with the s
review of some particular aspect of existing plants, the t.hemal overload protection provisions be audited.
ENCLOSURE 1 (CONT'D)
Encl osure A~~ ~~ '
Lee V. Gossick 3.
The Committee reviewed the proposed Regulatory Guide 1.XX:
INSTRUMENT SPAfiS AND SETPOINTS and recomended approval subject to the following comment:
Paragraph 5 of Section C (page 4 of the proposed Guide) should be rewarded in light of Committee comments, to the satisfaction of the Director, Office of Standcrds Cev31oonent.
This guide was characterized by the Committee as Category 1 - no backfit.
4.
The Committee reviewed Procosed Pegulatory Guide 1.97:
INSTRU:'ENTATICf! FCC LIGHT !!ATER CCOLED !'L' CLEAR CO lER PLA*;TS TO ASSESS PLAi;T CONDITIONS OURIT:G AND FOLLC'.!!i:G Af; ACCIDENT and deferred further consideration to a later meetira in order to permit incorpcration of recent connents by the Division of Technical Review.
/{
l Edson G. ' Case, Chairman Regulatory Requirements Review Committee s
ENCLCSURE 1 (CONT'D)
Enclosure B Septemoer 15, 1973 CATEOCRY 2 MATTERS Document NumDer Pevisian Date Ti tle RG 1.2/
2 1/76 Ultimate Heat Sink for *;uclear Power Plants RG 1.52 1
7/76 Design, Testing, ano Maintenance Criteria for Engineered-Safety-Feature Atmosahere Cleanup System Air Filtration and Adsorption Units of Lignt Water Cooled Nuclear Power Plants (Revision 2 has been cublisned but the cnanges from Revision i 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 Assemolies in Containment Structures for Lignt Water Cooled Nuclear Power Plants RG 1.91 1
2/78 Evaluation of Exolosions Postulated to Occur on Transportation Routes Near Nuclear Power Plan Sites RG 1.102 1
9/76 Flood Protection for Nuclear Power Plants RG 1.105 l
11/76 Instrument Setpo,tt RG 1.108 1
8/77 Periodic Testing of Diasel Generator Units 'Ised as Onsite
-' Nuclear Electric Power..
Power Plants
'rajectory RG 1.115 1
7/77 Protection Ag( 7*:
4/78 Tornado Design Classification RG 1.124 1
1/78 Service Limits anu Loading Comoinations for Class I s
Linear Type Component Supports RG 1.130 0
7/77 Design Limits and Loading Comoinations for Clas! 1 Plate-and Shell-Type Component Supports (Continued) 52?CL22'.'.~.0 _
Enclosure B CATEGORY 2 MATTERS (CONT'D1 Continued Document Numoer Revision Date Title RG 1.137 0
1/78 Fuel Oil Systems for Stancby Diesel Generators (Piragracn C.2)
RG 8.8 2
3/77 Information Relevant to Ensuring that Occuoational Radiation Excosures at Nuclear Power Stations Will be as Low as is Reasonably Achievable (Nuclear Dower Reactors)
BTP ASB Guidelines for Fire Protection for 9.5-1 1
Nuclear Power Plants (See Implementation Section, Section 0)
BTP MTE3 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
\\
- ENrt,OSUPE 2 (CCNT'Di
Enclosure C CATEGORY 4 MATTERS A.
Regulatory Guides not categori:ed Issue Date Number Revision Title 4/74 1.12 1
Instrumentation for Earthcuakes 12/75 1.13 1
Scent Fuel Storage Facility Design Basis 8/75 1.14 1
Reactor Coolant Pump Flywneel Integrity 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
Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors 6/74 1.80 0
Preoperational Testing of Instrument Air Systems 6/74 1.82 0
Sumps for Emergency Core Cooling and Containment Soray Systems 7/75 1.83 1
Inservice Inscection of Pressurized Water Reactor Steam Generator Tubes 11/74 1.89 0
Qualification of Class lE Equipment for Nuclear Power Plants 12/74 1.93 0
Availability of Electric Power Sources 2/76 1.104 0
Overhead Crane Handling Systems for Nuclear Power Plants ENCLOSURE 4
Enclosure C B.
SRP Criteria Implementa-Applicable tion Date
- Branch, SRP Section Title 1.
11/24/75 MTEB 5.4.2.1 BTP MTEB-5-3,. Monitoring of Secondary Side Water Chemistry in PWR Steam Generators 2.
11/24/75 CSB 6.2.1 BTP CSB-6-1, Minimum 6.2.lA Contairment Pressure Model 6.2.1B for PWR ECCS Performance 6.2.1.2 Evaluation 6.2.1.3 6.2.1.1 6.2.1.5 3.
11/24/75 CSB 6.2.5 BTP CSB-6-2, Control of Combustible Gas Concentra-tions in Containment Following a Loss-of-Coolant Accident 4.
11/24/75 CSB 6.2.3 BTP CSB-6-3, Determination of Bypass Leakage Path in 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, Oversead 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 B.
11/24/75 SEB 3.5.3 Procedures for Composite Section Local Damage Prediction (SRP Section 3.5.3, par. II.l.C) s
_. ENCLOSURE.4 (CCNT)
Enclosure C
. Imolementa-Acolicable tion Date Branch SRP Section Title 9.
11/24/75 SEB 3.7.1 Develoceent of Design Time History for Soil-Structure Interaction Analysis (SRP Section 3.7.1, par. II.2)
- 10. 11/24/75 SE3 3.7.2 Procedures for Seismic System Analysis (SRP Section 3.7.2 par. II)
- 12. 11/24/75 SE3 3.8.1 Design and Construction of Concrete Contaimrents) SRP Section 3.8.1, par. II)
- 13. 11/24/75 SES 3.8.2 Design and Construction of Steel Containments (SRP Section 3.8.2, par. II) 14.
11/24/75 SES 3.8.3 Structural Design Criteria for Category I Structures Insid.e Contai nment (SRP Section 3.3.3, 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)
- 17. 11/24/75 SE3 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) s ENCLOSURE 4(CCNT)
Enclosure C Implementa-Aeolicable tien Date Branch SRP Section Titl e
- 18. 11/24/75 SE3-3.3.2 Tornado Load Effect Combi-nations (SRP Section 3.3.2, par. II.2.d)
- 19. 11/24/75 SES 3.4.2 Dynamic Efects of Wave Action (SRP Section 3.4.2, par. II) 20.
10/01/75 ASB 10.4.7 Water Hammer for Steam Generaters with 3reneaters (SRP Section 10.4.7 par. I.2.b) 21.
11/24/75 AB 4.A Thermal wydraulic Stacility (SRP Section 4.4, par. II.5)
- 22. 11/24/75 RS3 5.2.5 Intersystem Leakage Detection (SRP Section 5.2.5 par. II.4) and R.G.1.45
- 23. 11/24/75 RS3 3.2.2 Main Steam Isolation valve Leakage Centrol System (SRP Section 10.3 par. III.3 and BTP RS3-3.2)
C.
Other Positions Implementa-Applicable tion Date Branch SRP Section Titl e 1.
12/1/76 SE3 3.5.3 Ouctility of Reinforced Concrete and Steel Structural Elements Subjected to Impactive or Impulstve Loads 2.
8/01/76 SE3 3.7.1 Response Spectra in Vertical Di rection 3.
4/01/76 SE3 3.8.1 BWR Mark III Containment Pool 3.8.2 Dynamics 4
9/01/76 SE3 3.8.4 Air Blast Loads 5.
10/01/76 SE3 3.5.3 Tornado Missile Impact 6.
6/01/77 RSS 6.3 Passive Failures During Lon'g-Term Cooling Fo11cwing LOCA EMCLOSURE 4 (CCNT)
Enclosure C 5-Implementa-Acclicable tion Date Brinch SRP Section Titl e 7.
9/01/77 RSB 6.3 Control Recm Positten Indica-tien of Manual (Hancwneel) Valves in the ECCS
~
8.
4/01/77 RSB 15.1.5 Leng-Term Recovery from Steamline Break : Operater Action to Prevent Overpressuri:ation 9.
12/01/77 RS:
5.4.6 Pumo Ocerability Recuirerents 5.4.7 6.3
- 10. 3/28/78 RS3 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 Ccnditions
- 13. 6/01/76 CSB 6.2.1.2 Asymmetric Loads en Ccmponents Located Within Centainment Sub-compartments 14 9/01/77 CSB 6.2.6 Centainment 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 Recuirements for Cec 11ng Water to Reacter Coolant Pumps
- 18. 8/01/76 ASB 10.4.7 Design Guidelines for Water Harmer in Steam Generators with Tcp Feedring Design (BTP ASB-10.2)
- 19. 1/01/76 fCSB 3.11-Environmental Centrol Systems for Safety-Rela ted Equipment s
ENCLOSURE 4 (CONT)
Enclosure C DESCRIPTION OF POSITI0f45 IDE!1TIFIED AS NRR CATEGORY 4 MATTERS IN ENCLOSURE 4,
PARAGRAPH C Numbering scheme corresponds to that used in Item C of Enclosure 4 ENCLOSURE 4 (CONT)
Enclosure C C.1 DUCTILITY OF REINFORCE 0 CONCRETE AND STEEL STRUCTURAL ELEMENTS SUEJECTED TO IMPACTIVE CR IMPULSIVE L0A05 INTRCDUCTION In the evaluation c'f overall response of reinforced concrete structural elements (e.g., missile barriers, columns, slabs, etc.) subjected to imcactive or impulsive loads, such as impacts due to missiles, assumotion 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 ccmoonents su;:ccrted or protacted by the elements are maintained.
The following sunnarizes specific SE3 interim positions for review and acceptance of ductility ratios for reinforced concrete and steel structural elements subjected to imcactive and impulsive loads.
SPECIFIC POSITIONS 1.
REINFORCE 0 CONCRETE MEMBERS 1.1 For beams, slabs, and walls where flexure controls design, the gemissible ductility ratio ( u ) under impactive and impulsive loads should be taken as 0.05 for o-o'
>.t05
.u
=
oo 10 for o-o' 1 005 u
=
where p and o'are the ratics of tensile and compressive reinforcing as defined in ACI-318-71 Code.
1.2 If use of a due:111ty ratio greater than 10 (i.e.,
u> 100) is required to demonstrate design adequacy of structural elements against impactive or imouisive loads, e.g., missile imcact, such a usage should be identified in the plant SAR.'
Information justifying the use of this relatively high ductility value shall be provided for SE3 staff review.
a ENCLOSURE 4 (CONT)
Enclosure C in 7 T!
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~N ~2mi o z-w ENCLOSURE 4 (CONT)
Enclosure C 1.3 For beam-columns, walls, and slabs carrying axial comcression loads and subject to im;ulsive or impactive loads producing flexure, the permissible ductility ratio in flexure should be as follows:
(a)
W1en compressicn centrols the design, as defined by an interaction diagram, the permissible ductility ratio shall be 1.3.
(b) When the ccmcression loads do not exceed 0.l f 'Ag or one-e 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 pernissible dutility ratio shall vary linearly frem 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 ccepressive impulsive or impactive loads only, without flexure, the permissible axial ductility ratio shall be 1.3.
1.5 For shear carried by concrete only u = I.0 For shear carried by concrete and stirrups or bent bars "I3 u
For shear carried entirely by stirrups u
= 3.0 2.0 STP.dCTURAL STEEL MEMBERS 2.1 Aar flexure compression and shear u
= 10. 0
- 2. 2 For colunns with slenderness ratio (1/r) equal to or less than 20 u
= 1.3 ENCLOSURE 4 (CCNT)
Enclosure C
. where 1 = effective length of the memoer r = the least radius of gyration
.For columns with slenderness ratio greater than 20 u = 1.0.
2.3 For memoers subjected to tension u =.5
[y" where cu= uniform ultimate strain of the material cY = strain at. yield of material C.2 RE3PONSE SPECTRA IN THE VERTICAL OIRECT!ON Subsecuent 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 recort 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 1.50, the vertical response spectrum is equal to the horizontal response spectrum between 3.5 ces and 33 cos. For the Western Uniteo States only, consistent with the latest available data in NUREG-0003, the option of taking the vertical design design rescanse spectrum as 2/3 the hori: ental 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 Regulatcry Guide 1.60.
C.3 BWR MARK III CONTAINMENT COOL OYNAMICS 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. Aoprcoriate load comoinations a
and load factors should be applied accordingly, s
b.
The pool swell loads and accident pressure may be comoined in accordance with their actual time histories of occurrence.
ENCLOSURE 4 (CONT)
4-Enclosure C 2.
SAFETY RELIEF VALVE (SRV) DISCHARGE a.
The SRV loads should be treated as live loads in all load ccmoinations 1.5Pa where a load factor of.1.25 should be applied to the appropriate SRV loads.
b, A single active failure causing one SRY discharge must be considered in comoination with the Design Basis Accident (08A).
c.
Aporopriate multiple SRV discharge should be considered in ecmbination with the Small Break Accident (SBA) and Inter-mediate Break Accident (ISA).
d.
Thermal loads due to SRV discharge should be treated as i for normal oceration and T for accident conditions.
a e.
The sucaression cool If ner shculd be designed in acccedance with the ASME Soiler and Pressure Vessel Code, Division 1 Subsection NE to resist the SRV negative pressure, considering strength, buckling and low cycle fatigue.
C.4 AIR BLAST LOADS (Pa, Ta, To as defined in ACI 359-740)
The following interim cosition 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 frem the air blast reflected pressure or the overpressure by multiplying these pressures by a factor of two. Any proposed use of a dynamic load factor less than two should be treated on a case by case basis.
Whether the reflected pressure or the overpressure is to be used for individual structural elements depends on whether an incident blast wave could strike the surface of the element.
2.
No load factor need be specified for the air blast loads, and the load ccmbination should be:
U=0+L+B where, U is the strength capacity of a section D is dead load L is live load B is air blast load.
3.
Elastic analysis for air blast is required for concrete structares of new plants. For steel structural elements, and also for rein-forced concrete elements in existing plants, scme inelastic response may be permitted with appropriate limits on ductility ratios.
ENCLOSURE 4 (CONT)
Enclosure C 5
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 riiding stability should be assessed by multiplying the structure's full projected area by the ecuivalent 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 succorting structures should also be analyzed for the effects of air blast to determine their ability to carry loads acclied directly to exterior panels and slabs. Moreover.fn 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 acting both inward and outward.
C.5 TORNADO MISSILE PROTECTION As an interim measure,the minimum concrete wall and roof thickness for tornado missfie protection will be as follows:
Wain inicxness Roof Thicxness Concrete Strength (psi)
(inches)
(inches) 2000 21 24 Region I 4000 24 21 5000 21 18 2000 24 21 Region II 4000 21 18 5000 19 16 2000 21 la Region III 4000 18 16 5000 16 14 These thicknesses are for protecticn against local effects only. Designers must establish indecendently the thickness requirements for overill structura response. Reinforcing steel should satisfy the orovisions of Accendix C, ACI 349 (that is,.25 minimum, E'4Er). The regions are described in Regulatory Guide 1.76.
ENCLOSURE 4 (CONT)
Enclosure C C.5 PASSIVE ECC5 FAILURES DURING LONG-TERM CCOLING FOLLOWING A LOCA Passive failures in :ne ECCS, having leak rates equal to or less than those frem the sudden failure of a pump seal and which may occur during the long-term cooling period following a postulated LCCA,should be con-sidered. To mitigati 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 ac:f ons may be taken on a timely basis.
The diagnostic and corrective actions would include the identification and isolaticn of the faulted ECC5 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) Cecostration that the leak detection system is sensitive enough to initiate and alam on a timely basis, i.e., with sufficient lead time to allow the operator to identify and f aolate the faulted line before the leak can create undesireaole 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 ECC5 train and that the leak can be isolated; and (5) Alarms that confom with the criteria specified for the control room alarms and a leak detection system that confoms with the require-ments of IEEE-279, except that the single failure criterion need not be imposed.
C.7 CONTROLROOMPOSIT!ONIN0fCATICNOFMANUAL(HAN0WHEEthVALVES 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 byeass or tnocerable condition affects a system that is designed to perform an automatic safety function.
ENCLOSURE 4 (CONb)
Enclosure C
. (2)
The bypass or inocerable concition can reasonably be ex ected to occur more frecuently than once per year.
(3)
The bypass or inoperable condition is expected to occur when the
' system is nomally recuired to operate.
Revision One o'f the Standard Review Plan in Secticn 5.3 requires confomance with Regulatory Guide 1.47 with the intent being that any manual (handwheel) valve which could jeocarcize the operation of the ECCS, if inadvertently left in the wrong cosition, must have position indicaticn in the control reen.
In the PDA extension reviews it is important to ecnfim that standard oesigns incluce this
~ design feature.
Most standard designs cc but this matter was probably not specifically addressec in some of the first FDA reviews.
C.8 LCMG-TERM RECOVERY FROM STEAM LINE BREAX - OPERATOR ACTION TO PREVENT OVEAFRE55uRILAsION (FwR)
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 iacreases and expands, refilling the pressurizer. Without operator action, replenishment of RCS inventory by the ECC3 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 frem the accident, and (ii) the pressures following the initiation of the break will not ccmcrcmise the integrity of the reactor coolant pressure boundary giving due considera-tien 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 ECC3.
C.9 PUMP OPERABILITf REGUIREMENTS In some reviews, the staff has found reasonable doubt that some tyces of engineered safety feature pumps would concinue to perform their safety function in the long tem following an accident.
In such instances there has been followup, including pump redesign in some cases, to assure that long term perfomance 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 cerformed to demonstrate that the puces are capable of operating for extended periods under post-LCCA canditiens, including the effects of debris. Of scuss the damage to pump seals caused by cecris over an extenced period of operation.
E!1 CLOSURE 4 (COtiT)
Enclosure C
-8 b.
provide detailed diagrams of all water cooled seals and conoc-nents in thr: cumps.
Provide a description of the composition of the pump shaft c.
' seals and the shafts. Provide an evaluation of less-of shaft seals.
d.
Discuss hew debris and cost-LOCA environmental conditions were factored into the specifications and design of the pump.
C.10 GRAVITY MISSILES VESSEL SEAL RING MISSILES INSIDE CONTAINMENT Safety related systems should be protected against icss of function due to internal missiles from sources :ucn as those associated wita pressuri:ed components and rotating equipment. Such sources would include but not be limited to retaining bolts, control rod drive assemblies, the vessel seal ring, valve bonnets, and valve stems. A description of the methods used to afford protection against suen 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.
\\
s ENCLOSURE 4 (CCNTl
Enclosure C
.g.
C.Il CORE THERMAL-4YORAULIC AtlALYSES In evaluating the thermal-hydraulic performance of the reactor core the follcwing additional areas should be addressec:
1.
The effect of radial pressure gradients at the exit.cf open lattice cores.'
2.
The effect of radial pressure gradients in the upper plenum.
3.
The effect of fuel red bowing.
In addition,a cc=mitment to perform tests to verify the transient analysis methods and codes is required.
c.12 DEGRADED GRID VOLTAGE CONDITIONS As a result of the Millstone Unit Number 2 low grid voltage occurrence, the staff has developed additional requirements concerning (a) sustained degraded voltage conditions at the offsite power source, and (b) inter-action of the offsite and onsite energency ;cwer systems. These additional requirements are defined in the following staff position.
1.
We require that a second level of voltage protection for the ensite power system be provided and that this second level of voltage pro-tection satisfy the folicwino requirements:
a) The selection of voltage and time set points shall be determined frem 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; ENCLOSURE 4 (CONT)
Enclosure C c) The time delay selected shall be based on the follcwing conditions:
(1)
The allowable time delay, including margin, shall not exceec the maximum time delay that is assumed in the SAR accident analyses; (ii) The time delay shall minimize the effect of short duration disturcances frem reducing the availability of the offsite power source (s); and (iii) The allcwable time duration of a degraded voltage condition at all distribution system levels shall not result in failure of safety systems or ccmpenents; (iv) The voltage senscrs shall automatically initiate the disccnnection Of offsite Ocwer scurces wnenever 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 conditions for operation, surveillance requirements, trip set points with minimum and maximum limits, and allowable values for the second-level voltage protection sensors and associated time delay devices.
2.
We require that the system design autcmatically prevent lead shedding of the emergency buses once the onsite sources are supplying power to all secuenced loads on the emergency b;ses.
The design shall also include the capability of the led shedding feature to be autcmatically reinstated if the ensite sx ce supply breakers are tripped. The autcmatic bypass and reinstate, rent feature shall be verified during the periodic testing identified in Itam 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 (a) simulating loss of offsite pcwer; (b) simulating loss tests:
of offsite power in conjunction with a safety injection actuation signal; and (c) simulating interruption and subsequent reconnection of onsite pcwer sourcer to their respective buses.
N EUCLOSURE 4 (CONT)
Enclosure C
. 4 The voltage levels at the safety-related buses shculd be optimized for the full load and minimum load conditions that are exsected througnout the anticipated range of voltage variations of the offsite pcwer source by acercoriate adjust-
' ment of the voltage tao settings of the intervening transformers.
We recuire th.at the adequacy of the design in this regard be verified by actual measurement, and by correlation.of measured values with analysis results.
C.13 ASYMMETRIC LOADS ON COMPONENTS LOCATED WITHIN CONTAINMENT SUSCCMPARTMENTS In the unlikely event of a pi::e rupture inside a :ajor ccm::enent sub-comDartment, the initial bicwdcwn transient wculd lead to cressure loadi :gs on both the structure anc the enc!csed ccm::cnent(s). The staff's generic Category A Task Action Plan A-Z is cesigned to develop generic resolutions for this matter. Our present schedule calls for completing A-2 for PkR's during the first quar er,1979. Pending completion of A-2, the staff is implementing the following program:
- 1. For PkRs at the CP/PCA stage of review, the staff requires apoli-cants to cocinit to address the safety issue as part of tneir appli-cation for an operating license.
2.
For Pk9s at the CL/FDA stage of review, the staff recuires case-by-case analyses, including implementation of any indicated corrective measusres prior to the issuance of an operating license.
3.
For BkRs, for which this issue is expected to be of lesser safety significance, the asynrnetric loading conditions will be evaluated on a case-specific basis prior to the issuance of an operating license.
For those cases which analyses are required we request the ;:er'emance of a succcmpartment, multi-noce pressure res,ponse analysis of the pressure transient resulting frem costulated hot-leg and cold-leg (pump suction and discharge) reactor coolant system pipe ructures within the reactor cavity, pipe penetrations, and steam generator compart:nents.
Provide similar analyses for the pressurizer surge and spray lines, and other high energy lines located in containment compartments that may be subject to pressurization. Show hcw the results of these analyses are used in the design of structures and ccmponent supports.
s s
ENCLOSURE 4 (CCMT)
-12 Enclosure C C lA CONTAINMENT LCAK TESTING PROGRAM To avoid difficulties excerienced in this area in recent CL reviews, the staff has increased its scoce of inouiry at the CP/pCA stage of revi ew.
For this curpose, the folicwing information witn re;ard to the centainment leak testing program should be sucplied.
Those systems that will remain fluid filled for the Type A test a.
should be identified and justification given.
b.
Shcw the design provisions that will permit the personne' air-lock door seals and the entire air 1cck to be tested.
For each penetration,1.e., fluid system picing, instrument, c.
electrical, and equicment and cersonnel access generations, identify the Type B and/cr Type C local leak testing that will be done, d.
Verify that containment penetrations fitted with expansion bellows will be tested at P.
Identify any penetration fitted with a
expansion bellows that does not have the design capability for Type B testing and provide justification.
C.15 CONTAINFENT 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 (MSLS) for designs utilizing pressuri:ed water reactors with conventional containments show that the peak calculated containment temoerature can exceed for a short time period the envircnmental qualification temDerature-time envelope for safety related instruments and ccmconents. 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 recuirement for requalifying safety-related equipment to higner time-temperature envelopes.
The staff's generic Category A Task Action Plans A-21 and A-24 are designed to develop generic resciutions for these matters. The presently scheduled completion dates for A-21 and A-24 (Short Term Partien) are first quarter,1979 and fourth quarter,1978, respectively.
Pending ccmoletion of A-21 and A-24, some interim guidance will be used as detailed belcw.
We have developed and are imclementing a clan in which all acclicants for construction permits and operating licenses and those already ishued con-struction permits must provide information to establish a conservative temoerature-time envelope.
ENCLOSURE 4 (CONT)
Enclosure C
. Therefore, describe and justify the analytical ecdel used to c:nservatively detemine the maximum c:ntainment temoerature and pressure for a spectrum of postulated main steam line breaks for varicus reacter gewer levels.
Include the folicwing in the discussion.
(1) Provide single active failure analyses which specifically identify those safety grade systems and ccmacnents relied upon to limit the mass and fnergy release ano containment pressure /
temperature rescense. The single f ailure analyses should include, but not necessarily be limited to: main steam and connected systems isolation; feedwater auxiliary feedwater, and connected systems isciation; feedwater, condensate, and auxiliary feecwater pump trip, and auxiliary feedwater run-out control system; the :oss of er availability of offsite ocwer; diesel failure when loss of offsite pcwer is evaluated; and partial loss of containment cooling systems.
(2) Discuss and justify the assumptions made regarding the time at wnich active containment heat removal systems become effective.
(3) Discuss and justify the heat transfer c:rrelation(s) (e.g., Tagami, Uchida) used to calculate the heat transfer frem the containment atmosphere to the passive heat sints, and provide a clot of the heat transfer coefficient versus time for the most severe steam line break accident analyzed.
(4) Scecify and justify the temperature used in the calculation of condensing heat transfer to the passive heat sinks; i.e.,
specify whether the saturation temoerature correscending to the partial pressure of vapor, or the atmosphere temperature (which may be superheated)was used.
(5) Discuss and justify the analytical model including the thermodynamic equations used to acccunt for the removal of the c:ndensed mass frem the containment atmosphere due to condensing heat transfer to the passive heat sinks; (6) Provide a table of the peak values of containment atmosahere temcerature and pressure for the spectrum of break areas and gewer levels analyzed; (7) For the case which results in the maximum containment atmosphere temperature, graphically show the containment atmoschere temperature, the containment liner temperature, and the containment concrete temperature as a functicn cf time. C;mpare the calculated contain-ment atmosphere temcerature rescense to the temoerature profile used in the environmental cualification program for these safety related instruments and mecnanical ccmcenents needed Oc mitigate the consequences of the assumed main steam line break and effect safe reactor shutdown; ENCLOSURE 4 (CCNT)
Enclosure C
-la-(8) For the case which results in maximum contairment atmosonere pressure, grachically show the containment pressure as a function of time; and (9) For the case whicn results in the maximum contairment atmosphere pressure and temperature, provide the mass and energy release data in tabular fem.
In order to demonstrate that safety-related ecui; ment has been adecuately qualified as described above, provide the follcwing infomation regard-ing its environmental qualification.
(1) Provide a ecmcrehensive list of ecuicment recuired to be acerational in the event of a main steamline break (MSL3) accident. The list should include, but not necessarily be limited to, the following safety related ecui;: ment:
(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 fo110 wing a main steam line break accident should be based on the fo11cwing:
(a) A condensing heat transfer coefficient based on the reccmmendations in Branch Technical Position CSB 6-1,
" Minimum Containment Pressure Mcael for P'aR ECCS Perfomance Evaluation,"should be used.
(b) A convective heat transfer coefficient should be used wnen the condensing heat flux is calculated to be less than the s
convective heat flux. During the bicwdewn certad it is acercoriate to use a conservatively evaluated forced convection heat transfer correlation. For examcle, ENCLCSURE 4 (CONT)
Enclosure C-Nu = C(Re)
Where Nu = Nusselt No.
Re
- Reynolds No.
C
= empirical constants dependent on geometry and Reynolds No.
Since the Reynolds namter is dependent on velocity, it is necessary to evaluate the forced ficw currents wnicn will be generated by the steam gener?ce bicwoown. The CVTR ex=eriments provide limited data in this regard. Convective currents of from 10 ft/see to 30 ft/sec were measured locally. We recorm:end that the CVTR test results be extracolated conservatively to obtain forced flew currents to detemine the convective heat transfer coefficient during the bicwdown period. After tne bicwdown has ceased or been reduced to a negligibly Icw value, a natural convection heat transfer correlation is acceptacle.
(3) For each ccmponent where themal analysis is done in conjunction with an environmental test at a temperature icwer than the peak calculated temperature following a main steam line break accident ecmcare the test themal. responsa of the C0mponent witn the accident themal analysis of the c:mponent. Provide the basis by wnich the component thermal response was developed frcm the environmental qualification test program. For instance, graphically show the themoccuole data and discuss the themoccuple locations, method of attachment, and perfomance characteristics, or provide a detailed discussion of the analytical model used to evaluate the comconent themal response during the test. This evaluation should be perfomed for the potential points of failure such as thin cross-sections and temperature sensitive parts where themal stressing, temperature-related degradation, steam or chemical interaction at elevated temoeratures, or other themal effects could result in the f ailure of the ecmcr.nent mechanically or electrically.
If the ccmponent themal response ccmparisen results in the predicticn of a more severe themal transient for the accident c0nditions than for the qualification test, provide jur.tification that the affected ccmponent will perfom its intended function during a MSL3 accident, or provide protection for the ccmoonent whch would appropriately limit the themal effects.
7
.h
Enclosure C C.16 ENVIRCNMENTAL EFFECT CF PIPE FAILURES Identtfy the " break exclusion" regions of the main steam and feedwater lines. Comoartments that contain break exclusion regions of main steam and feedwater lines and any safety related equipment in these compartments should be designed to with-stand the environmental effects (pressure, temperature, humidity and flooding) of a crack with a break area equal to the cross sectional area of the ' break excluded' pipe.
C.17 DESIGN REQUIREPENTS FOR CCCLING '4ATER TO REACTCR CCOLANT PUMPS Demonstrate that the reactor coolant system (RCS) cumo seal injection flow will be automatically maintained for all transients and accidents or that enougn time and information are availahla " " " ~ ' '
ta
-.--4+
corrective action by an operator.
We have established the following criteria for that portion of the component cooling water (CCW) system wnich interf aces with the reactor coolant pumps to suoply cooling water to pumo seals and bearings during normal operation, anticipated transients, and accidents.
- 1. A single active failure in the component cooling water system shall not result in fuel damage or a breach of the reactor coolant pressure boundary (RCFB) caused by an extended loss of cooling to one or more pumps. Single active f ailures include operator error, spurious actuation of motor-operated valves, and loss of CCW pumps.
- 2. A pipe crack or other accident (unanticipated occurrence) shall not result in either a breach of the RCPB or excessive fuel damage when an extended loss of cooling to two or more RC pumps occurs. A single active falure shall be considered when evaluating the consequences of this accident. Moderate leakage cracks should be determined in accordance with Branch Tecnnical Position ASB 3-1.
In order to meet the criteria established above, an NSSS inter-f ace requirement should be imposed on the balance-of-alant CCW system that provides cocling water to the RC cuma seals and motor and puma bearings, so that the system will meet the following con-ditions:
s ENCLOSURE 4 (CCNT)
Enclosure C
-17 1.
T. hat portion of the com:enent c:aling water (CCW) system whicn supplies cooling water to the reactor coolant pu=cs and mtcrs may be designed. *w non-seismic Category I recuirements and Quality Groua 0 if it can be demnstrated nat the reactor c:alant pumos will operate without component ecoling aater for at least 20 minutes without loss of function or the need for cperstar pro-tactive action. In addition, safety grade instrumentation including alans should be provided to detect the loss of component cooling water to the reacter coolant euros and motors, and to notify the operator in the control rocm. The entire instrumentation system, including audible and visual alarms, should meet the recuirements of IEEE Std 279-1971.
If it is not demonstrated that the reactor coolant pum:s and motors will operate at least 30 minutes without loss of functicn or acerator 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 reac;or protection system shall be provided to initiate automatic protection of the plant. For this case, the ccmoonent c:oling water supply to the seals and pump and motor bearings may be designed to non-seismic Ca tegory I require-ments and Quality Group 0; or 2.
The ccmponent 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 Catecory I, Quality Group D and ASI4E Section III, Class 3 requirements.
The reactoscolant (RC) pumps and motor's are within the NSSS scope of design. Therefore, in order to demonstrate that an RC pumo design can operate with loss of component cooling water for at least 20 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 c:moonent ecoling water to the RC pumes and an analysis demon-strating that no consecuences imcortant to safety may result frem this event.
Include a discussion of the effect that the s loss of cooling water to the seal coolers has on the RC pumo seals. Shew that the loss of cooling water does not result in a LOCA due to seal failure.
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Enclosure C 2.
A detailed analysis to shew that icss of cooling water tc the RC ; umps and motors will not cause a loss of the flow coastdcwn characteristics or cause seizure of tne pumos, assuming no administrative acticn is taken. The response shculd include a detailed descriptien of the calculation procedure includi'ng:
a.
The equations used.
b.
The parameters used in the equations, such as the design parameters for the motor bearings, motor, cum and any other ecuipment entering into the calculations, and material procerty values for the of_1 and metal parts.
c.
A discussien 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 acerceriate figures) associated with the RC pumo and motor and their design criteria and standards.
e.
Information to verify the acplicability of the equations and material properties chosen for the analysis (i.e.,
references should be listed, and if emoirical relations are used, provide a comparisen of their range of appli-cation to the range used in the analysis).
Should an analysis be provided to demonstrate that loss of ccmponent cooling water to the RC pumps and motor asseccly 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 STEAM GENERATCRS WITH TCP FEEDRING DESIGN Events such as damage to the feedwater system piping at Indian Point Unit No. 2, November 13, 1973, and at other plants, could originate as a consequence of uncovering of the feedwater sparger in the steam generator or uncovering of the steam generator feedwater inlet noz;:les. Subsequent events may in turn lead to the generation of a cressure wave that is propagated thrcugh the pipes and could result in unacceptable damage.
s ENCLOSURE 4 (CONT)
Enclosure C
-19 For CP/DCA and CL/:DA a:clications, pro"ide the fc11cwing for steam generators utilizing top feed:
1.
Rrevent or delay water draining from the feedring following a drop in steam generator water level by means such as,J-Tubes; 2.
.1e volume of feedwater pioing external to the steam generator whch could pocket steam using the shortest possible (less than seven feet) hori: ental run of inlet pioing to the steam generator feedring; and 3.
Perform tests acceptable to the staff to verify that unacceptable feed-water hammer will not occur using the plant ocerating oracedures for normal and energency restoration of stean generator water level folicwing 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:
a.
Describe normal operating occurrences of transients that could cause the water level in the steam generator to drop below the sparger or noz:les to cause uncovering and a11cw steam to enter the sparser and feedwater piping.
b.
Describe your criteria or show by isometric diagrams, the routing of the feedwater piping frem 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 performed or the results of test programs to verify that.either uncovering of feedwater lines could not occur or tnat, if it dia occur, unacceptacle damage such as the experience at the Indian Point Unit No. 2 facility would not result wf th your design, s
ENC'OsunE 4 (CONT)
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Enclosure C
-2 0 -
C.19.U4VIR0 time! ITAL CCl4 TROL SYSTEMS FOR SAFETY llELATED E0llIPMENT Most piang areas that contain safety related equipment depend on the continuous operation of enviremnental control systems to maintiin the
.anvironment in those are'as within the range of enviromaental qualification of the safety related equipment installed in those ar' ras.
It appears
- hat there are no requireme'nts for maintaining these :!nvirer. ment.11 i:cntrol systems in operation while the plant is shutdmm or in ho: standby
.:endi tions. During periods when these environmentai control sys: ens are shutdown, the safety related equipment ceuid be ex osed to en'/ironmental conditions for which it has not been qualified. Therr3 fore, th.i safety related equipment should be qualified to the extreme environmental conditions that could occur when the control equipment is shutdown or these environmental centrol systems should operate centinuously to naintain 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 cualified shall be provided. This environmental monitoring system shall (1) be of hign quality, (2) be periodically tested and calibrated to verify its continued functioning, (3) be energized frem continuous pcwer sources, and (4) provide a continuous record of the environmental parameters during the time the envir nmental conditions exceed the normal limits.
etCLCSUas 4 (CONT)