ML20154J946
| ML20154J946 | |
| Person / Time | |
|---|---|
| Issue date: | 08/05/1985 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20151L097 | List: |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR REGGD-01.XXX, REGGD-1.XXX, NUDOCS 8603110129 | |
| Download: ML20154J946 (76) | |
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ENCLOSURE 1 Regulatory Guide 1.XXX GUIDE FOR LICENSEE PREPARATION AND NRC STAFF REVitW OF PLANT-SPECIFIC ,
ANALYSES REQUIRED BY THE PRESSURIZED THEPJ8AL SH0CX AULE 10CFR50.61 l
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Table of Contents PAGE I. INTRODUCTION . ....................
A. Background and Purpose of this Guide . . . . . . . . . .
B. Obje'ctives of Plant-Specific PTS Safety Analysis ,
Reports ........................
C. Staff Review of Plant-Specific PTS Safety Analysis Reports and Criteria for Acceptability of Continued Ope ra ti o n. . . . . . . . . . . . . . . . . . . . . . . .
II. STANDARD FORMAT AND CONTENT OF PLANT-SPECIFIC PTS SAFETY ANALYSIS REPORTS. . . . . . . . . . . . . . . . .
A. Use of Standard Format . . . . . . . . . . . . . . . . .
- 1. Style and Composition . . . . . . . . . . . . . . .
- 2. Procedure for Updating or Revising Pages. . . . . . '
l 3. Physical Specifications of Submittals . . . . . . .
4 Number of Copies. . . . . . . . . . . . . . . . . .
l B. Content of Reports . . . . . . . . . . . . . . . . . . .
1.0 Overall Approach, Scope of Aqalysis, and Report ;
Organization ...................
2.0 Plant Data
- j 2.1 Systems Pertinent to PTS . . . . . . . . . . .
2.2 Reactor Vessel. ...............
2.3 Fluence. . . . . . . . . . . . . . . . . . . .
2.4 ISI Results. . . . . . . . . . . . . . . . . .
2.5 Plant Operating Experience . . . . . . . . . .
- 2.6 Operating Procedures . . . . . . . . . . . . .
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Table of Contents (continued)
PAGE 3.0 Detemination of Detailed PTS Sequences for .
Analysis
' 3.1 Approach Used ...............
3.2 Sequence Delineation . . . . . . . . . . . .
3.3 Operator Effects . . . . . . . . . . . . .
3.4 Sequence Quantification. . . . . . . . . . .
3.5 Event Tree Collapse. . . . . . . . . . . . .
4.0 Themal-Hydraulic Analysis. . . . . . . . . . . .
4.1 Themal-Hydraulic Analysis Plan. . . . . . .
4.2 Themal-Hydraulic Model. . . . . . . . . . .
4.3 Simplified Analysis Methods. . . . . . . . .
4.4 Thermal Stratification Effects . . . . . . . :
4.5 Themal Hydraulic Analysis Results . . . . .
5.0 Fracture Mechanics Analysis . . . . . . . . . . .
6.0 Integration of Analyses . . . . . . . . . . . . . l t
7.0 Sensitivity and Uncertainty Analysis of
- Through-Wall Crack Frequency. . . . . . . . . . . I t
7.1 Sensitivity Analysis . . . . . . . . . . . . !
7.2 Uncerta'nty Analysis. ........... l 7.2.1 Parameter Uncertainties .......
7.2.2 Modeline uncertainties (Biases) ...
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Table of Contents (continued) ;
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- 8.0 Effect of Corrective Actions on Vessel Through-Wall ,;
Crack Frequency . . . . . . . . . ... . . . . . . . . I 8.1 Operating Procedures and Training Program Improvements . . . . . . . . . . . . . . . . . .
8.2 Flux Reduction Program . . . . . . . . . . . . .
4 8.3 Inservice Inspection and Nondestructive :
Evaluation Program . . . . . . . . . . . . . . . ,
8.4 Plant Modifications .............. .
l 8.5 In-Situ Annealing ...............
9.0 Prediction of Vessel Failure Mode . . . . . . . . . . I i
10.0 Likelihood of Core Melt and Prediction of Risk ...
j 10.1 Risk Analysis .................
10.2 General Guidance ...............
u 10.3 Core Melt Frequency .............. (
10.4 Containment Failure ..............
e 10.5 Source Term .................. ;
j 10.6 Site Consequence Analysis ........... ;
10.7 Risk ..................... I 10.8 Uncertainties .................
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- s. i 11.0 Results and Conclusions Regarding PTS Risk .....
l 11.1 Summary of Analysis .............. l
- 11.2 Basis for Continued Operation ......... [
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! III. REFERENCES i :
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INTRODUCTION A. Background and Purpose of this Guide The Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61, issued in 1985, establishes a screening criterion based on reactor vessel nil-ductility-transition temperature (RTNDT). The screening criterion was established after extrensive industry and NRC analyses regarding the likelihood of PWR vessel failure due to PTS evants. The analyses were applied generically and contained conservative assumptions so as to make the results bounding for any PWR. Based on the results, the NRC concluded that the risk due to PTS events is acceptable at any plant so long as the RT PTS
- of the reactor pressure '
vessel remains below the screening criterion.
Extensive safety analyses are required by the rule for any plant that wishes to operate with RTPTS values above the screening criterion. The recommended methods to be used in perfntming the analyses are outlined in this Guide.
The purpose of the analyses is to assess the risk due to PTS events for proposed operation of the plant with reactor vessel RT PTS above the screening criterion. Completion of these analyses is required by 10CFR50.61 three years before exceeding the screening criterion. This is to allow adequate time for implementation on the plant of any corrective actions assumed in the analyses before the plant operates above the screening criterion.
- To avoid confusion between several (pre-existing) slightly different e i
definitions of RTNOT, the PTS rule contains its own definition o.f an RTNDT (called RTPTS) to be used when comparing plant-specific vessel material properties with the PTS screening criterion. ,
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2-This Regulatory Guide describes an acceptable standard format and content-for these plant-specific PTS safety analyses, and describes acceptance criteria that the NRC staff will use in evaluating licensee analyses and proposed corrective measures.
The references listed in Section III of this Guide include a set of analyses sponsored by the NRC that, taken together, constitute an example of the analyses described by this Guide. The staff reco. amends that those references be extensively utilized, along with this Guide, by those performing the plant-specific PTS analyses required by the PTS rule, 10CFR 50.61.
References 1, 2, and 3, for example, each represent an analysis by Oak Ridge National Laboratory, predic' ting through wall crack frequency for one PWR.
These references will provide guidance through a majority of the analyses.
Reference 3 (analysis of H. B. Robinson) should be most helpful, as it was the last one performed and includes the experience gained in performing the other two earlier analyses.
B. Objectives of Plant-Specific PTS Safety Analysis Reports The PTS Rule (10CFR50.61(b)(4)) requires that licensees whose plant will exceed the screening criteria before expiration of the operating license "shall submit a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent potential failure of the reactor vessel as a result of postulated PTS events if continued operation beyond the screening criterion is allowed." These analyses must include the effects of all corrective actions the licensee believes necessary to achieve an acceptable PTS-related risk for continued operation of the plant. The final objective of the plant-specific PTS study, therefore, is to justify continued operation of the plant by demonstrating that the risk from such operation is acceptable. This will require estimation, as a function of remaining effective full power years of plant life, of:
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- expected frequency of vessel through-wall crack due to PTS expected frequency of core melt due to PTS
- expected frequency of large release of radioactivity from containment due to PTS, and the resulting person-rem exposure and early and late fatalities that may result from PTS
,- effect of various corrective actions on the above frequency and risk estimates In calculating these results, it will be necessary to:
- Identify the domireant accident sequences.
- Identify operator actions, control actions, and plant features important to PTS.
- Estimate the effectiveness of potential corrective actions in reducing the expected frequency of through wall crack, core melt, and large release of radioactivity.
- Identify the sources and approximate magnitude of the major uncertainties and their effects on the conclusions.
- Present and justify the licensee's proposed program for corrective measures.
Present and justify the licensee's proposed basis for continued operation at embrittlement levels above the screening criterion. This must include comparison of the PTS-related risk at the plant, with corrective actions implemented as necessary, with the Acceptance Criteria given in Section I.C of this Guide.
C. Staff Review of Plant-Specific PTS Safety Analysis Reports and Criteria for Acceptability of Continued Operation I
The PTS rule specifies a screening criterion based on RTNDT (called RTPTS IOP '
use as defined within the rule) of 270*F for axial weld and plate materials and 300 F for circumferential weld materials. As detailed in SECY-82-465 i
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(Ref. 23), these values were selected based on generic studies of the e:pected frequency and character of a wide spectrum of transients and accidents that could cause pressurized overcooling of the reactor vessel (PTS events), and on operating experience data. The " risk" due to PTS events was assessed in terms of probabilistic fracture mechanics calculations of the expected frequency of through-wall crack penetration of the pressure vessel due to the PTS events.
In selecting the screening criterion based on those calculations, the conserva-tive assuhiption was made that any through-wall crack could result in severe core degradation or melt. Core melt itself was viewed as an event to be avoided even though risk to the public due to such an event, in terms of person-rem and early and late fatalities, was not calculated with any certainty. The estimated througn-wall crack frequenc9 developed as a function of RT NDT for axial welds (Fipure 6.3, Ref. 23) is showi in Figure 1.
The RTPTS screening criterion selected by the staff corresponds to a mean (or average, "best estimate") surface RT PTS of 210*F. The staff utilized a "2-sigma" value (spread between "best estimate" and " upper limit") of 60*F,* and so the screening criterion, which is expressed in terms of RT PTS which by definition is this upper limit value, was selected at 210 + 60=270 F. For axial weld and plate materials, Figure 1 gives a through-wall crack frequency of about 5 x 10 -6 per reactor year at an RT PTS of 270*F. For circumferential welds, the same frequency occurs at approximately 300*F (Ref. 23). The Commission concluded that the PTS-related risk at any PWR is acceptable so long as the RT PTS values remain below the specified screening criterion.
It was reali7ed that there are many unknowns and uncertainties inherent *in the probabilistic calculations, and so it was with deliberate intent that conservative assumptiors such as those stated above were made. The expectation was that the true risk at any plant due to PTS events would in all likelihood be considerably
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. 6 below that derived from Figure 1 and therefore would be acceptable. Also con-tributing to the belief that the real PTS risk at any given plant was lower than that resulting from the analysis in Reference 23 was the belief that many of the generic plant assumptions made in Reference 23 (for example, material properties, system performance, crack distribution) would prove to be overconservative for analysis of a specific plant, and that the resulting plant-specific analysis, ,
when performed, would likely result in a reduced prediction of PTS risk.
In determining acceptability of the plant-specific PTS anslyses submitted by licensees in accordance with 10CFR50.61, a through-wall crack penetration frequency criterion (mean frequency less than 5 x 10 -6 per reactor year) will be used in conjunction with calculated risk due to PTS events in terms of core melt,. person-rem, and early and late fatalities. Thus, a through-wall crack mean frequency below 5 x 10 -6 per reactor year will not, by itself, define an acceptable analysis result. Rather, an overall, plant-by-plant judgement will be made based on comparison of the plant-specific through-wall crack mean fre-quency to 5 x 10 -6 per reactor year, plus comparison of the calculated risk results (in terms of person-rem and fatalities) to applicable risk criteria (or safety goals) established by the NRC at the t.ime the analyses are required to be performed. Thus, the risk analyses described in Sections II.B.9 and II.B.
10 of this Guide will be required as part of the plant-specific PTS analysis.
In all of the analyses performed, the licensee must justify that the important input values used are valid for the remaining life of the plant.
It is anticipated that the first 10CFR50.61 - required PTS analyses will not be started until several years after publication of this guide. In view of ,
the present extensive on going efforts by the Commission in the areas of quantitative assessment of risk and definition of a safety goal, it is not possible to publish in this guide the quantitative criteria that will be in effect when the first PTS analyses are reviewed. The staff will use the core L__
e melt, person rem, and early and late fatality quantitative criteria in effect at the time when reviewing those portions of a licensee's analyses which purport to justify continued plant operation on those bases.
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II. STANDARD FORMAT AND CONTENT OF PLANT-SPECIFIC PTS SAFETY ANALYSIS REPORTS A. Use of Standard Format The standard format and content is described in succeeding chapters. Use of
! this format will help ensure the completeness of the information provided, will assi%t the NRC staff in locating the information, and will aid in 1 shortening the time needed for the review process. If the licensee chooses to adopt the standard format and content, the numbering system of Section II.B of this document should be followed at least down to the level of subchapter." Certain subchapters may be omitted if they are clear'y unnecessary to provide for comprehension of the analysis or if they are l repetitive. In such cases, adaptation of the standard format to accommodate the particular circumstances is appropriate.
- 1. Style and Composition 1
The licensee should strive for a clear, concise presentation that portrays:
the plant systems analyzed and data used, the engineering analyses performed, j the results obtained, and the conclusions drawn.
The presentation style should be addressed to a typical reader wno is an engineer in one, but not all, of the technical disciplines involved.
l Accordingly,jargonshouldbeavoidedandalistofacronymsprovided, o
Drawings, diagrams, and tables should be used when information may be q presented more adequately or conveniently by such means. Care should be J taken to ensure that all information presented in drawings is legible, that j symbols are defined, and that drawings are not reduced to the extent that f they cannot be easily read.
- The "!!.B" prefix should be dropped in the licensee's analyses. Therefore, i for example, Chapter 2.0 will be " Plant Data," subchapter 2.1 will be l
- " Systems Pertinent to PTS," etc.
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- , o l l A table of contents should be included.
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- 2. Procedure for Updating or Revising Pages l l The updating or revising of data should be on a replacement page basis, i l The changed or revised portion of each page should be highlighted by a I l
vertical 11ne. The line should be on the margin opposite the binding margin l for each line changed or added. All pages submitted to update, revise, or I add pages to the report are to show the date of change. The transmittal letter should include an index page listing the pages to be inserted and the l
pages to be removed. When changes or additions that affeet the table of contents are made, a revised table of contents should also be provided.
- 3. Physical Specifications of Submittals l
The report should conform to the following phys'ical dimensions of page size, quality of paoer and inks, numbering of pages, etc.:
! (1) Page Size 1
l Text pages: 8-1/2 x 11 inches l Orawings and graphics: 8-1/2 x 11 inches preferred; however, a larger size is acceptable provided the finished copy when folded does not exceed 8-1/2 x 11 inches.
l (2) Paper Stock and Ink j I
Suitable quality in substance, paper color, and ink density for handling j l and for microfilming.
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r o (3) Paper Margins A margin of no less than one inch should be maintained on the top, bottom, and binding side ~of all pages submitted.
(4) Printing Comptsition: text pages should be single spaced.
Type face and style: must be suitable for microfilming.
Reproduction: may be mechanically or photographically reproduced. ,
Pages of the text may be printed on both sides with the images pr14,ted ,
head to head. /
(5) Binding Pages should be punched for looseleaf standard 3-hole ring binding.
(6) Page Numbering ,
Pages should be numbered sequentially.
(7) Format References .
4 In the application, references to this standard format should be by chapter and section numbers. *
- 4. Number of Copies The licensee should submit 3 copies to the Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington,_00
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20555. These copies may be filed at the Commission's officas at 1717 H -
Street, NW, Washington, DC or at 7920 Norfolk Avenue, Bethesda, MD.
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B. Content of Reports 1.0 Overall Approach, Scope of Analysis, and Report Organization The report should describe the overall approach to the analysis and outline the individual tasks in terms of the nature and source of input, the methods used for analysis, and the nature and subsequent use of the output. The interreittionship of the tasks should be described and should be illustrated by a flow chart. Describe how the analysis tasks are integrated to achieve the results and conclusions.
Major emphasis should be placed on analyzing event sequences leading to vessel failure; secondary emphasis may be given to analyzing subsequent core melt and release of radioactivity.
The report should include both probabilistic and deterministic fracture mechanics analyses. .The deterministic analyses should be used to predict material condition limits beyond which a given event would cause vessel througn wall crack penetration for the most limiting initial crack size.
This can be expressed in terms of a " critical RTNDT. " The probabilistic analyses should be used to determine the statistical likelihood of vessel through-wall crack penetration assuming a crack size distribution appro-priately justified for the vessel being analyzed, and appropriate uncertainties and distribution of the significant input parameters, such as material properties, etc. Both methods should be used to compare predicted frequency of vessel through-wall crack before and after implementing th4 proposed corrective measures.
The input to the probabilistic analysis should be best-estimate based on appropriate assumptions. Uncertainties and conservatisms should be
, explicitly presented in the decision rationale for the licensee's proposed corrective measures and basis for continued operation.
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- - The analysis should include effects of operator actions,. control system interactions, and support systems such as electric power, instrument air, and service water cooling.
l The report should be organized by starting with a description in Chapter 1.0 of how the report chapters and supporting appendices are interrelated
'and what material is in the appendices.'
The main report should: describe the objectives and,overall approach used in the study, outline the plant systems analyzed, describe the engineering analyses performed, present the results obtained and conclusions drawn, and present and justify the licensee's proposed program of corrective measures.
Appendices should contain data, detailed models, sample calculatioEs, and detailed results needed to support the various chapters of the report.
Appendices should contain little supporting text. Instead, the nature and relevance of material in the appendices should be described in the pertinent chapters of the main report.
Certain details (some are noted later in this Guide) may be available for NRC inspection in Audit Notebooks and need not be directly submitted with the analyses.
2.0 PLANT DATA This chapter should briefly describe plant systems and operations pertinent to PTS. Chapter 2 of Reference 3 (the H. B. Robinson analysis by Oak Ridge !
National Laboratory) provides a good example. Supporting appendices or l references should present the design and operating data used in the analysis f or needed to understand the' analysis. References to otheridocuments (e.g., '
the FSAR) should indicate specific sections. (Reliability data, however, ;
should go in Section 3.4, " Sequence Quantification," or its supporting appendices and references.)
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2.1 Systems Pertinent to PTS Summarize design and operating features of systems pertinent to PTS.
Illustrate each system with a simplified process and instrumentation diagram ,
or a single line diagram. Identify on each illustration cny interfaces with i
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other systems. For each system, include a table summarizing key design and operating data. Give the maximum, minimum, and nominal values for tnose 4 cases whe*re design data may vary with time (for example, HPI water temperature may vary with season). Such values used in the' analysis should [
be identified and justified. Refer to appendices or other dccuments (e.g.,
specific sections of the FSAR) as necessary for more details. Systems to be ,
i considered should include pertinent portions of:
- Reactor Cooling System
- Condensate and Main Feedwater Systems
- Steam System
- Auxiliary Feedwater System
- Reactor Protective System
- Chemical and Volume Control System
- Instrumentation and Control Systems
- Support Systems
-Electric Power
-Instrument Air .
-Service Cooling Water 2.2 Reactor Vessel Summarize the reactor vessel construction and its material properties. Use tables, drawings, or graphs to show:
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- Vessel design (include weld locations and hot leg and cold leg penetrations)
- Vessel materials and chemical composition in the beltline region (include both base and weld material properties.)
- Vessel fabrication procedures particularly wel' ding and cladding j
toughness data including the upper shelf regime, residual stresses, ,
fTaw density distribution, etc.). Describe and justify methods used l to calculate or otherwise determine properties.
Available information on the vessel properties should be re-examined in detail to fill any gaps in the supporting data for making an estimate of RT NDT and to support resolution of any disagreements about the validity of l values used.
Few data are currently available and validated to support the selection of a value for the initial RT HDT. The confidence that can be placed in estimates of the' initial RT NDT depends not only on material tests, but also on the accurate documentation of welding technique, weld wire used, and weld flux used. The credibility of such estimates could be enhanced by performing more tests on archival material, by discovering previously unreported test results on weld specimens from the particular plant, or by evaluating properties of
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welds considered typical of the plant-specific weld.
2.3 Fluence Present the current and projected fluence on the vessel using benchmarked computer programs and methodology and information from neutron flux surveillance dosimetry. Use the weld locations and fluence values to j identify the critical welds. Show how the fluence varies along the length ,
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'their uncertainty.
l 2.4 Inservice Inspection Results j i
i To the extent pertinent to the probabilistic analysis and proposed corrective actions, summarize:
~* Re'sults - The. number, size, depth and . loc'ation of any flaws found j should be well defined and described. <
- Methods Used - The method used to perform the inspection should be well described with documentation of any validation information.
o Note: Only those inservice inspections which have actually'been performed should be discussed in this section. Improved ISI programs as proposed by the licensee should be described under; corrective measures in Chapter 8 and !
not included in the base analysis effort. I 12.5 Plant Operating Experience Summarize overcooling transients that have occurred at this station or at a sister station. Summarize lessons learned from these and.other transients i and indicate actions taken to prevent recurrence or minimize severity of I overcooling transients.
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2.6 Operating Procedures This section provides procedural data; e.g. what the operator is supposed to
~do and when. This section, for example, should present and describe the
- important operator actions as defined by existing procedures associated with potential over-cooling transients. The conditions under which the operator takes each action, the expected time for performing the action, and how the 3 l
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time was derived should be identified. Some examples of these operator actions are:
- Trip reactor coolant pumps
- Throttle / terminate emergency core coolant
- Throttle / terminate main and emergency feedwater
- Restore main and emergency feedwater
- - Isolate break (primary or secondary)
Supply a sumary of training materials associated with overcool.ing events in general and with respect to principal initiators. In addition, a summary of simulator exercises associated with potential overcooling events should be provided.
(Note: Proposed improvements in procedures, diagnostic instrumentation, display systems, and operator training should be presented in section 8.1 under the licensee's program of corrective measures and should not be presented in this section or used as part of the base analysis.)
3.0 DETERMINATION OF DETAILED PTS SEQUENCES FOR ANALYSES This chapter presents the methods and analyses used to identify those transient sequences which'could contribute significantly to the PTS risk.
A good example is presented in Chapter 3 of Reference 3. The scope includes identifying initiating events, developing event trees, modeling and
. quantifying the reliability of relevant systems and operator actions, and collapsing the event trees to identify specific relevant sequences. Detailed models, data, and sample calculations should be included in appendices or referenced. However, the logic of the analysis, criteria used, results, and insights gained should be described in the main report.
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3.1 Accroach Used Describe how the material presented in this chapter fits into the overall PTS study. Provide a general description of the process used to identify PTS sequences. It should be made clear how the approach used will result in completeness of identification of all classes of events that could contribute significantly to PTS risk, how specific events are selected for more detailed analysis 'to represent each class, and finally how the events so analysed are used to determine total PTS risk at the plant.
3.2 Secuence Delineation Identify potential overcooling transients in a well defined manner and document them in such a way that it is clear to a reviewer that all important potential overcooling conditions have been considered. Classes of init'iators should be developed, important variations of initiators within each class should be identified, and potential transients resulting from these initiators should be defined.
Development of Classes of Initiators Any class of transients which could lead to overcooling of the reactor vessel should be considered in the analysis. It should, however, be appropriate to use logical arguments to eliminate classes of transients as actual PTS initiatorswheneverjustifiable. Examples of initiators which should be included are:
- LOCAs including steam generator tube rupture accidents ;
- Steamline breaks
- Overfeeds
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- Combinations of these Identification of Important Initiator Variations Once the classes of potential initiators have been identified, it is important to consider variations within any individual class. These ~l variations should include:
(1) Decay heat level - The decay heat level, determined by recent operating history of the plant, can have a major impact on the potential ;
consequences of a given event. Thus various decay heat conditions should <
be considered. Clearly, decay heat associated with a reactor trip from full power (assuming operation at full power for some considerable time) ;
should be examined. Zero decay heat represents the opposite extreme but l for all practical purposes occurs only once at the beginning of life for the plant when PTS is not important. Therefore, the analyst may choose i to use some other level of decay heat which wo'uld cover potential decay l heat conditions after the initial start-up of the plant. The reasons
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for choosing particular decay heat levels for analysis should be documented. Each identified initiator should be examined at all decay heat levels defined whenever appropriate.
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(2) Power level - Power level may be important since certain equipment '!
'I conditions or configurations may only exist at certain power levels, for example hot standby. As in the case of decay heat level identification, the reasons for the selection of specific power levels for analysis purposes should be stated. It should be noted that under certain conditions a reactor system may be at a high power level with a low g decay heat condition.
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(3) Location of event - In many . instances'the location of the event is defined. For. example, an event consisting of a failed open turbine bypass valve has the location defined since it is a specific valve failure. However, for some events such as pipe breaks the location is
'not defined and could have an impact on the progression of the event.
.In the case where location is not defined, all locations that could be significant should be considered. Each location should then be eliminated by logical argument, bounded by consequences associated with another location, or treated as a separate event.
(4) Magnitude of event - Many of the initiators can occur to various degrees. For example, a LOCA can range' from a very small break to a i full guillotine pipe break. Break sizes should be examined to identify {
categories of sizes khich lead to similar system conditions. In.the- ,
case of the LOCA event, special consideration should be given to the :
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identification of break sites that could lead to loop flo'w stagnation. <
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The larger size'LOCAs typically do not contribute to PTS risk since the
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pressure cannot be maintained due to the large flow out the break. j Definition of Potential Transients Resulting from Each Initiator ;
= Once the complete set of significant initiators has been defined, event trees h are required to identify potential sequences resulting from each initiator. j The development of the event tree headings and branches should be done in a. ,
consistent and logical manner. This was done in the ORNL studies (Ref. I i'
1,2,3) by using what have been called system state trees. These trees define the potential states of each plant system of interest conditional on specific thermal- h'draulic y conditions. Initiator specific event trees can then be c developed by examining the system state trees with respect to each initiating l event. A similar or equivalent approach should be used to assure trace-l l l
t f
?
I J
- n. , -~:.___ .u- --
. , - - - . . . ..- : n ..n - -
t t
[
ability of the event trees and to-assure that important sequences are not inadvertently eliminated.
Support system failures must also be presented within some type o'f event tree structure. If the event trees are developed as previously described, any ,
support system failure would most likely lead to a sequence of events which is already mapped out on the event trees, but in many instances with a higher probability of occurrence. In other cases, it may be necessary to define !
event trees resulting from a support system failure. In either case, it is important that the support systems be examined to identify their potential impact on overcooling conditions. The results of this examination should be presented as a separate section with the identification of specific support '
system failure sequences of interest. The support system review should at f least include: ,
- the electrical supply system I
- the compressed air instrument system
- the component and service water systems !
r I '
3.3 Operator Effects The operator effects are analyzed in two separate sections. In this section ;
the potential operator actions are identified. These actions are further ,'
analyzed in Section 3.4 when the probabilities associated with the L performance of an operator action are developed.
The operator can improve, aggravate, or initiate an overcooling transient.
All three of these categories should be discussed in this section.
- 1) Procedures and/or the operators' general knowledge can lead to actions which improve the conditions associated with an overcooling event. j t
w . + . _ - - - -,w,_. .. _- - .. .. -
Explanation should be included as to why it is perceived that this action would be taken. Where appropriate, these operator actions should be either included directly on the event trees or presented as separate operator action trees which can later be coupled with the principal event trees.
- 2) Although the ORNL studies (Refs. 1,2,3) did not include operator init*iated events or events aggravated by operator actions contrary to l procedures, this category of events should also be examined as part of a plant specific analysis.
l
- 3) Events not normally considered to be overcooling events should be examined to identify the potential for operator actions resulting from misdiagnosis or operator error which could lead to an overcooling event.
In addition previously identified initiators should be analyzed for i potential operator actions which could lead to a more severe cooldown.
The ccnfusion matrix approach (Ref 4) used in human reliability analysis could provide a structure for identifying and analyzing these potential operator actions. i 3.4 Sequence Quantification .
Quantify the event trees by using identified initiating event frequencies, appropriate conditional probabilities associated with the success or failure i of specific equipment operations, as well as success and failure probabilities associated with operator actions. Plant specific data should be used whenever appropriate to define these probabilities, including appropriately adjusted simulator studies. This should be supplemented by vendor-specific or PWR generic data bases when plant specific data do not appear to provide an adequate data base. The "Probabilistic Safety Analysis l;
Procedures Guide" (NUREG/CR-2815) includes guidance about treatment of 4
.. , ;.- . -.m_
_ _- m --_m ._ _ .. . .-. .. . _ . -
generic and plant specific data. Its appendices include an updated generic data base that should be used.
Identify by specific reference or provide in appendices all of the reliability data used as input to quantify the event sequences. An explanation should be supplied as to how the data were derived for each data point.
1 Initiating Events Initiating event frequencies'should be developed based on the number of observed events within selected periods of operation for the general type of plant under consideration. -If no failures have been observed and no other information is available with which to estimate a probability, a standard statistical method such as the Poisson distribution can be used to determine a probability, or the technique described in the latest revision of the NUREG/CR 2815 appendices for estimation of plant specific initiating event frequencies can be used. For some initiators, it may be necessary to i estimate the frequency of events in a particular operating mode, e.g. hot zero power. The data should be researched to identify trends associated with the occurrence of the event and the operating mode. In addition, the initiator itself should be examined to identify physical conditions which might favor failure in one mode rather than another. If this examination reveals no evidence of correlation between frequency and operating mode, the fraction of time spent in each operating mode can be used as a weighting factor.
- Equipment Failures 1 Following each initiating event, certain components are designed to perform in a defined manner. Failure of a component to perform its required function i
1
(
r
g __ _ _ _ - . .__ _ _ _ _ . . _ _ . -- m___
Q 9 could lead to PTS considerations. Thus it is necessary to assign a failure and successful operation probability for each component on a per-demand basis. These probabilities can be obtained by estimating the number of failures observed within a period of time, combined with an estimate of the number of demands expected within that same period, or by the development of fault trees. If no failures have been observed and no other information is available with which to estimate a failure-on-demand probability, a standard statistic'ai method as previously discussed' can be used to develop a probability.
As with all event trees, the probability associated with a particular branen ,
is conditional on the prior branches in the sequence. Questions of conditional probability should be carefully considered before a failure probability is assigned.
The potential for coupled or common cause failures must be examined in the analysis. Careful consideration should be give'n to increasing the failure .
potential.of a component given the failure of one or more components of the I same type. As additional components of a particular type are postulated to fail, the probability for the next component of the same type to fail should increase. Based'on the ORNL analysis, a simplified approach would be to l assume that the failure probability of the second component, given that the i first component has failed, might be as high as 0.1. The third component might ,
be assumed to fail with a 0.3 probability, given the failure of two identical components. One could then assume that after the failure of three components of the same type all remaining components of that type would fail with a probability of 1.0. The licensee should discuss how these types of coupled t failures are handled in the analysis.
Common cause failures of a different type may occur, as previously discussed, l through the failure of a support system or a control signal. An analysis.of t
G
.._ _ _ .. ~
these potential failures should be made and the branch _ probabilities should be a'd justed whenever appropriate.
. Operator Actions Operator action probabilities are particularly difficult to determine due to the lack of a data base. The problem is further complicated when time becomes a'n important variable. The procedure outlined below represents one approach to quantify operator actions. This procedure should be conservative for any operator action performed as required by procedures assuming that the equipment required is operational. For operator actions which might not be associated with procedure steps, it is not clear that this simplified approach would produce conservative frequencies. Therefore the approach described would only be recommended for operator actions associated with procedure steps. Regardless of the method used, the human error probabilities used in these analyses should be supported by data validated for the plant being analyzed.
(1) Identify operator actions - In this step the procedures associated with each initiator would be reviewed to identify those operator actions which would have an impact on downcomer temperature.
1 (2) Identify time cor.straint - In the case of each operator action the transient would be reviewed assuming no operator action to identify the time frame available for successful completion of the operator action.
l (3) Assign screening failure probabilities - In this step a conservative l value for the failure of the operator action would be identified. For j operator actions required by procedures to be performed within the first 5 minutes of the transient the time-reliability curve as presented in
. .~. _. .. . . . . . . - . .
NUREG/CR-2815 (Ref. 5) could be used to identify a screening value.
After 5 minutes, a value of 0.9 for success and 0.1 for failure would be assumed for all operator actions. The entire PTS analysis would then be completed using these screening values.
(4) Identify dependency factors - In some instances there may be coupled ,
failures associated with. operator actions just as there were coupled fail'ures associated with equipment failures. In many instances the potential failure of a operator action may be linked, to various degrees, to the success or failure of a previous operator action. Thus, it is recommended that each opnrator action be reviewed with respect to dependency. This can be accomplished using the dependency tables as presented in the Human Reliability Handbook (NUREG/CR-1278) (Ref. 6).
(5) If any of the dominant sequences involve the failure of an operator action, a more comprehensive evaluation of the-failure would he performed for that operator action. When necessary the comprehensive evaluation should be performed using one of the accepted human reliability methodologies.
3.5 Event Tree Collaose Collapse the event trees using a frequency screening criterion to form a list of specific sequences and a set of residual groups to be analyzed. This is j important since the event trees may generate thousands of end states which can not be individually analyzed. A screening value of 1.0E-7/ reactor year ,
is recommended. This value should assure that important sequences are f
! treated individually and it should also help to keep the size of the residual smalL This is particularly important since it may be necessary to treat the !
residual using a bounding consequence condition.
i I
-_- . - + _ _ , ~ . . - - n_ _. -n n - - . . . - . . - - .
. . l Specific Sequences Those sequences which survive the frequency screening should be defined and their frequency noted. It is recommended that some identification be assigned to each sequence to enhance the traceability of each sequence through the remainder of the analysis. Grouping and identifying each sequence with respect to initiator type may also prove helpful.
Residual Groups Those sequences which do not survive the frequency screening must also be considered. They should be grouped together based on transient characteristics to form a set of residual groups. The residual groups should be reviewed to identify sequences which should be grouped with previously defined sequences because of transient similarity or specifically evaluated because of their severe consequence. It is important-to attempt to reduce the size of each residual group since it will be necessary to assign a bounding consequence which would apply within each group. Each residual group should be defined and its frequency noted.
4.0 THERMAL HYDRAULIC ANALYSIS l
This chapter should present the reactor coolant pressures, temperatures, and q heat transfer coefficients at the vessel's interior surface in the beltline
- region for the set of overcooling sequences which envelops the plant's I potential for experiencing a PTS event. A good example is presented in' Chapter 4 of reference 3. It should also present the details of the analysis methods used to obtain these fluid conditions. This chapter should include the following sections:
(1) The thermal-hydraulic analysis plan and logic l (2) A description and evaluation of the thermal-hydraulic models
, t
r----- . -- a - - . + . - - ~-- ~ ~ - .~. -
l l
1 (3) A description of any simplified analysis methods used in the stuoy (4) A description of the methods used to evaluate the effects of thermal stratification and mixing (5) Graphs of all of the best estimate thermal-hydraulic results with their associated uncertainties and a detailed explanation of the transient behavior observed 4.1 Therinal-Hydraulic Analysis Plan This section should outline the logic and identify the subtasks,in the thermal-hydraulic analysis. Subtasks include detailed thermal-hydraulic '
systems analysis, simplified thermal-hydraulic systems analysis, and thermal stratification analysis. The logic should describe the sampling plan used to select sequences for detailed or simplified analysis. ORNL experience favors selectin ~g detailed thermal-hydraulic analysis sequences including at least a few severe examples of each type of postulated overcool,ing transient in order to understand and benchmark the plant behavior for subsequent simplified calculations. The order in which the scenarios are evaluated can result in a considerable reduction in expenditures. By first analyzing the scenarios which are expected to be the bounding cases (i.e., the most severe),
calculations for an entire class of overcooling scenarios may be deemed i unnecessary if the bounding case is not of PTS concern. Similarly, careful !
selection of the first set of scenarios to be evaluated can permit simple !
extrapo1ation or interpolation of the results to other scenarios that share
~
common controlling thermal-hydraulic phenomena. ;
. j During the analysis, the sequence identification analyst and the thermal-hydraulic analyst should coordinate activities to ensure that pertinent l details of the delineated sequences are thoroughly understood. Similarly, close coordination must be maintained between the thermal-hydraulic analyst I
% , _ _ .... m _ __ _ - . .
- . . . . . - ~ _ _ _ _ _
t .
and the' fracture-mechanics analyst so that the transient fluid conditions are calculated at the appropriate vessel locations.
l 4.2 Thermal-Hydraulic Model This section and supporting appendices should present a detailed description of the thermal hydraulic computer models used in this analysis. The models should intlude an accurate representation of the pertinent parts of the primary and secondary systems. This includes the condensate system, the main and auxiliary feedwater systems, and parts of the steam system. The model should include appropriate' secondary side metal heat capacity. Particular attention should be given to the modeling of control system logic and characteristics, such as valve closure times and liquid level measurements.
References 7 through 10 illustrate some of the modeling details included in such a study. The thermal-hydraulic models should be capable of predicting single and two phase flow behavior, and critical flow as required. The models should be capable of predicting plant behavior for 1.0CA's, steamline breaks and steam generator tube ruptures. In general, a one-dimensional code is suitable for most overcooling transient calculations. However, if any of the control systems are dependent solely on the fluid conditions in a single loop (e.g., reactor coolant pump restart criteria), then a method of esti-mating the three-dimensional effects in the downcomer may be necessary for some of the asymmetric cooldown scenarios encountered in the PTS study.
Sensitivity of calculated results to the nodalization schemes used should be discussed.
1 This section of the report must also present the results of benchmarking the computer models against suitable plant data or data from experimental .
facilities. As a minimum, the plant data comparison should fully exercise the acdeling features which are employed'in the thermal-hydraulic computer programs, such as the pressurizer sprays, heaters and liquid level controls, i
, - ~ . - . . . - . - . - - - - _ .. - . ,- _ _
a s the steam generator liquid level controls, and the turbine bypass (i.e.,
steam dump) controls under steady-state and transient conditions. If overcooling transients have occurred at the plant, or at a similar plant, they should be benchmarked against the computer models. The licensee is :
encouraged to use codes and methods accepted by NRC at the time the i calculation is performed.
The model's should be capable of accurately predicting steam condensation in the pressurizer during the repressurization phase of an overcooling transient. Effects of non-condensible gases, if present, on system pressure and temperature calculations should be addressed.
All code input and modeling assumptions must be documented in an audit notebook available for NRC review.
4.3 Simplified Analysis Methods This section should present the technical bases for any simplified analysis methods which are applied in the study. This includes the grouping of similar sequences by controlling phenomena and any extrapolations used to modify existing calculations. If a simplified thermal-hydraulic plant model is used to predict portions of the plant transients, then all of the simplifying assumptions inherent to this model must be stated and justified. ,
Reference 11 provides examples of how to group sequences and develop a simplified thermal-hydraulic model suitable for portions of the analysis.
i 4.4 Thermal Stratification Effects l
Transient thermal-hydraulic computer programs available to analyze LWR ,
response to overcooling scenarios do not model fluid behavior with sufficient !
detail to. predict the onset of HPI thermal fluid stratification in the cold j
i
s s leg and the subsequent cold leg and downcomer behavior. As a result, additional analysis methods may be needed to determine which transients are affected by thermal stratification and the extent of such effects.
This section should describe and justify the thermal fluid mixing analysis methods which have been applied in the study. References 12 through 17 describe the results of recent mixing analyses and experiments. Reference 12-identifids a useful stratification criteria to determine which overcooling transients will require the additional mixing analysis. Particular attention should be given to scenarios which involve HPI injection under very low flow or stagnant loop conditions. When stagnation is partial (i.e., not all loops stagnate), stratification is expected only within the cold legs of the stagnant loops. However, scenarios involving complete loop stagnation will require the evaluation of a transient cooldown in the presence of stratified layers both in the cold legs and in a portion of the downcomer. The mixing model should include the effect of metal heating on the mixing behavior, particul1rly in a stagnant flow situation. Also, the effect of non-condensibles (if present) must be included. References 12 through 16 describe tools which have been used for such an analysis.
This section should also document the heat transfer correlations applied in the mixing analysis. The research efforts described in References 12 through 17 indicated that the downcomer heat transfer coefficients generally exceeded 300 Btu /hr-sqft-F. These values of heat transfer coefficient were generally i high enough to. keep the vessel wall surface temperatures within a few degrees ,
of the downcomer fluid temperature. Furthermore, because the vessel wall .
. cooldown was controlled by conduction processes rathbr than convection l processes, the vessel wall surface temperatures were insensitive to heat [
transfer coefficient variations due to changes in flow and heat transfer f
regimes.
I se %
4.5 Thermal-Hydraulic Analysis Results This section should present graphs of the best-estimate.downcomer pressures, fluid temperatures and heat transfer coefficients and their associated uncertainty ranges as a function of time at the critical weld areas. This includes the results of the detailed thermal-hydraulic model, the simplified model, and mixing analysis calculations.
The duration assumed for each overcooling scenario must be justified.
Assuming a scenario duration of two hours may be reasonable for.many cases since the overcooling transient would probably be identified and mitigated prior to that time. However, there may be scenarios requiring lengthier evaluation periods because the controlling phenomena delay the scenario's evolution.
This section should also include a discussion of the accuracy of the results and how the predicted plant behavior compared to plant history and operator experience. Time dependent uncertainty estimates for the downcomer pressure, fluid temperature and heat transfer coefficients at the critical welds must be provided for each scenario. These uncertainties are often limited by physical phenomena. For example, the pressurizer PORV setpoints will limit :
the system pressure for certain high pressure scenarios. Therefore the i uncertainty is limited by PORV operating characteristics. References 9 and 11 describe some uncertainty analysis techniques. !
5.0 FRACTURE MECHANICS ANALYSIS ;
For each sequence identified in Chapter 3 calculate (or for unimportant !
segments estimate, using bounding conditions) the conditional probability'of f thru-wall crack penetration P(TWCP/E), versus effective full power reactor years. A good example is provided in Chapter 5 of reference 3. Input for these calculations include the primary system pressure, the temcerature of I
~ _. - _ _ _ - - - - ._. -__ _. - -_
i :
the coolant in the reactor vessel downcomer, the fluid-film heat transfer coefficient adjacent to the vessel wall, all as a function of time, and the vessel properties. Although the licensee was required to utilize the -
10CFR50.61(b)(2) specified method of determining RTNDT (RTPTS) when evaluating his vessel properties with respect to the screening limits, in performing these plant-specific calculations, he is encouraged to utilize '
any alternate methods / data / correlations for which ne provides justification of applicibility to this plant. The calculatiens should be performed with a probabilistic fracture mechanics code such as OCA-P or VISA (References 18 l
and 19).
An acceptable procedure to be followed in the fracture-mechanics analysis is as follows. A one-dimensional thermal and stress analysis for'the vessel wall should be performed. The effect of cladding should be accounted for in both the thermal and stress analyses. .The fracture-mechanics model can be based on linear elastic fracture mechanics with a-specified maximum value of KIato account for upper shelf behavior. Plastic instability should be considered in the determination of failure. Acceptable types of material x properties are given in the study of the H. B. Robinson reactor (Ref. 3).
In the Monte Carlo portion of the analysis, as a minimum, each of the following must be assigned distribution functions:
Kl e
- Static crack initiation fracture toughness KI, - Crack arrest fracture toughness RT NDT
- Nil-ductility reference temperature ,
Cu - Concentration of copper, wt. % j Ni - Concentration of nickel, wt % '
F - Fast neutron fluence.
~
The functions used must be justified. Examples of these distributions are found in Reference 3.
The following additional information must be supplied:
l i
- n. - _ ,. . _ . . .. - .. _ . - . - _ - - . . - - .--
t
~
Flaw density - The number of cracks per unit surface area must b'e established for use in the calculations and must be justified. A value of 0.2 flaws per square meter (one flaw / cubic meter) was selected in Refs. 1, 2, and 3.
Flaw depth density function - The flaw depth density distribution must be established. The function to be used can be that specified in Refs.
1, 2', and 3. .
Flaw size, shape and location - Axial flaws with depths less than 20% of the wall thickness and all circumferential flaws should be modeled in '
two dimensions. Axial flaws with depths greater than 20% of the wall thickness should be modeled in two or three dimensions depending on the relative toughness of the weld regions and plate material. For instance, the length of an axial flaw in an axial weld that suffers severe radiation damage relative to the plate can be limited to the length of the weld.
All regions of the beltline must be considered. This includes axial and circumferential walds as'well as the base material.
The following relationships are required:
KIc= f(T,RTNDTo,dRTNDT),and KI, = f(T, RTNDTo,dRTNDT) where: T = the wall temperature i t'
RTNDio = initial Nil-ductility reference temperature 1
If ,
dRTNDT = increase in Nil-ductility reference temperature due to radiation damage, f(Cu,Ni, fluence). If plant surveillance data meet the criteria for credibility given in Ref. 20, they may be used as described therein.
Examples of these functions are described in References 3 and 20.
In reporting the results, the methods used for the probabilistic vessel-integrity analysis should be described, their limitations for this analysis identified, and the impact of uncertainties in the resulting vessel .
failure probabilities estimated. Discussion of the analysis should include a listing of the assumptions used, their bases, and a di:cussion of the sensitivity of the results to variations in the assumptions. Vessel dimensions and material properties used should be given.
The results of each transient should include a set of critical-crack-depth curves (see Refs. 1, 2, and 3), i.e., a plot of crack depths corresponding to initiation and arrest events vs time. This plot should also have curves indicating when warm prestressing is effective, the depth at which upper shelf toughness is effective and the depth at which upper shelf toughness is reached. These results should correspond to -2(sigma) values for KI and c
KI,, +2(sigma) values for RTNDT, mean values of all other parameters and an appropriate number of effective full power years (i.e. , corresponding to the RTNDT screening criterion or greater).
6.0 INTEGRATION OF ANALYSES In this chapter, the event frequencies are coupled with the results of the ;
~
fracture mechanics analysis to obtain an integrated frequency o.f vessel through-wall crack due to PTS. An example of one acceptable method is presented in Chapter 6 of Ref. 3. A table should be provided which supplies t
, - _ - - . , . . . _ _ _ , - _ _ _ _ _ . _ - . . . . _ _ , , _ _ . _ _ _ _ , - _ _ , - _ - - , _ - _ _ _ _ , ___y.-
.. s the following information for each specific sequence and residual group identified in section 3.5. These results should be provided for the operating time at which the reactor will reach the PTS screening criterion and for any additional operation life being requested.
" Sequence identification
- Type of initiator (small break LCCA with low decay heat, large steamline break at full power, etc)
- Estimated sequence frequency
- Method used to determine conditional thru-wall crack penetration probability
- Sequence conditional through-wall crack penetration probability (1) ~
" Frequency of through-wall crack due to sequence obtained by product of sequence frequency and sequence conditional through-wall crack penetration probability.
For each dominant sequence a section and/or table should be provided which supplies: 1) specific reference to the graph of temperature, pressure, and flow as provided in Chapter 4; 2)a time-line description of the accident sequence noting important operator actions, control actions, protective system acticns, equipment faults, and vessel failure; and 3) frequency of through-wall crack penetration as a function of effective full power years.
Results should then be summed within each initiator type to provide a frequency of through wall crack penetration as a function of initiator type.
W The conditional through-wall crack penetration probability is the probability of a through-wall crack as determined by the fracture mechanics analysis, given that the event occurs.
- - - -. . .. ~ . . - - . . ---- - ~ .. . - - - - - - -
t :
The discussion should explain why each initiator type is or is not important to PTS.
Finally, the results should be summed over each initiator type to provide an
~
integrated frequency of through-wall crack for the vessel. This integrated value should be reported as a function of EFPY and plotted with uncertainty ,
values as determined in Chapter 7 included on the plot. The discussion should identify important operator actions, control actions, and plant features that can cause or prevent vessel failure.
7.0 SENSITIVITY AND UNCERTAINTY ANALYSES OF THROUGH-WALL CRACK FREQUENCY .
In order for the results of the probabilistic analysis to be useful for regulatory decision making, the sensitivity of the results to input parameters and assumptions must be determined, the major sources of uncertainty should be identified, and the magnitude of the, uncertainty estimated. in this chapter, the results and the procedures used to perform each of these processes are documented. A good example is given in Chapter 7 of reference 3.
7.1 Sensitivity Analysis Perform a sensitivity analysis to estimate the change in the through-wall '
crack frequency for a known change of a single parameter. Parameters examined in the sensitivity analysis should include: 1)the initiating event and event tree branch frequencies, 2)the thermal-hydraulic variables (temperature, pressure, etc), and 3)the fracture mechanics variables .
(fluence, flaw density, etc). Where appropriate, 68th percentile (1 sigma) values should be used to represent the change in the parameter. This should f provide a sufficient change to illustrate the effects of the change, and the use of the 68th percentile value whenever possible will help to define the
-- ~ . . - - , , - - - , - , , - - - - - . _n-.- - , - . , - - - - - - - - - - - - - , - - - - - - - - - - - .
_ _ ~ - . -.~ . ..
- e. g important variabilities. In the case of temperature and pressure, however, the 68th percentile values may vary from one sequence to another. In tnis case, it may be easier to identify a representative change in the parameter which could then be used for all sequences rather than try to use the 63th percentile values.
Each variable examined in the sensitivity analysis should be listed along with the change in the variable. In the cases where changes are represented by using 68th percentile values, some explanation should be provided to document the reasons the value is considered a 68th percentile value. In those cases where something other than a 68th percentile value is chosen, ,
discussion should center around the reasons for choosing the value used.
Sensitivity factors should be obtained by dividing the through-wall crack frequency obtained with the changed variable by the through-wall crack frequency obtained with each variable at its mean value. Supply the sensitivity factors obtained for both positive and negative changes in each of the variables. The sensitivity factors obtained for changes made in the PTS-adverse direction should be ranked according to magnitude and provided in table form.
7.2 Uncertainty Analysis 7.2.1 Parameter Uncertainties Each step in the probabilistic analysis should include an uncertainty
- analysis. This should include uncertainty in frequency of occurrence of a -
sequence, uncertainty in temperatures and pressures reached during the sequence, and uncertainty in the fracture mechanics model for vessel failure given the transients.
p . _- =- ._ , ._ -- ._v,n _ _ , . _ _ . . ...
e !
For the following reasons, a Monte Carlo simulation is appropriate for PTS uncertainty analysis.
- The temperature and pressure error distributions are not symmetric.
- 'iht fracture mechanics results are non-linear with respect to viriations in' input parameters, particularly the temperature and
. pressure time histories.
.The results of the Monte Carlo analysis can indicate the shape of the output distribution.
The Monte Carlo approach would involve four steps as described below: -
l
- 1) Develop a statistical distribution for each variable used in the ,
calculations- This step will involve the representation of each l
variable as a distribution with 5th and 95th percentiles as pr,3viously identified. The shapes of the distributions selected should be discussed. -
~.
- 2) Select a random value from each distribution - A random sampling code should be used to sample from each of the distributions.
- 3) Calculate a through-wall crack frequency estimate based on values obtained in the previous step - In this step, the through-wall I
crack frequency is obtained based on the randomly selected ,
variables. This requires understanding the form of the !
relationship between each input variable and through-wall crack [
frequencies. For some variables, such as initiating event and l branch frequencies and flaw density, this is simple since the f through-wall crack frequency is directly proportional to the value f f
I
_ . . ~ . -.
of these parameters over the range of variable values considered.
Other variables such as temperature and pressure may require the development of an appropriate relationship. In such cases where the effect of a variable change may be dependent on the value of another variable, response-sur14ce techniques can be used to estimate important interaction effects.
- 4) ' Summarize the resulting estimates and approximate frequency distribution - Steps 2 and 3 are repeated until a statistically valid numDer of trials have been performed. A distribution of through-wall crack frequencies is tnen produced from the results of the trials. The 95th and 5th percentiles and the mean (expected value) of this distribution should be identified and discussed.
7.2.2 Modeling Uncertainties (Biases)
During the process of performing tne PTS analysis, the analyst will make simplifying assumptions in order to make the analysis tractable. Such assumptions include decisions on thermal-hydraulic models, fracture mechanics models, grouping of sequences both for thermal-hydraulic analysis and fracture mechanics analysis, etc. These assumptions can introduce conservative or non-conservative biases into the analysis. These biases should be identified and their potential impact on the results discussed. In this section important assumptions made as part of the analysis should be listed. Each assumption should be identified as being either conservative or !
. non-conservative. A discussion should be supplied for each assumption with i respect to its impact on the overall value of through wall crack frequency.
Whenever excess conservatism or non-conservatism is suspected to be present in an assumption, an alternative assumption should also be used in the full calculative procedure and the impacts on the overall result compared.
l 1
l L
y ,. _ - _ _ _ , _ . _ . ..% _ ,, _ .
8.0 EFFECT OF CORRECTIVE ACTIONS ON VESSEL THROUGH-WALL CRACK FREQUENCY This chapter summarizes the licensee's program of corrective measures. Each corrective measure considered by the licensee should be presented and explained in this chapter. In each case the reasons for considering the action as a corrective measure should be documented and the estimated impact of the action with respect to through-wall crack frequency should be provided.' Corrective actions that should be considered include, but are not limited to, those discussed in the remaining subsections of the chapter. An example can be found in Chapter 8 of reference 3.
8.1 Flux Reduction Program Early analysis and implementation of such flux reductions as are reasonably practicable to avoid reaching the screening criterion are already being required and accomplished in accordance with the PTS rule, 10 CFR50.61.
Further flux reductions to critical areas of the vessel wall should be considered that would reduce the risk of continued operation beyond-the screening criterion. If such additional flux reductions are needed, in view of the irreversibility of embrittlement, the licensee should, consider early implementation before reaching the screening criterion. For licensees who are considering applications to extend the operating license beyond its present expiration date, it may be prudent to implement the reduction as early as possible to avoid the necessity of vessel annealing or replacement.
8.2 Ooeratiro Procedures and Training Program Imorovements Operator actions and associated plant response play a key role in the ,
initiation and mitigation of pressurized thermal shock events. Therefore ensure that the actions are based on approved technical guidelines that include an integrated evaluation of relevant technical considerations, including, but not limited to, PTS, core cooling, environmental releases and
- - . . _ _ - ~ . . - - - . - _ _ _ -
. .o .
X '
containment integrity. The evaluation should adoress the following types of cencerns:
Frequent realistic " team" training should be conducted, exposing ,
the operators to potential PTS transients and their precursor j events. The training should give the operators actual practice in l controlling reactor system pressure and cooldown rates during PTS !
situations. Specific training should include but is not limited f to: RCP pump trip criterion, HPI throttling criterion, control of !
natural circulation, recovery from inadequate core cooling, '
recovery from solid plant operations, and use of power operated ,
relief valves (PORVs) to control primary over pressure.
Instructions should be based on analyses that include consideration of system response delay times (e.g., loop transport time, thermal
', transport time).
I
- Whether or not there is a need for cooldown rate limits for periods shorter than one hour should be evaluated.
- Methods for controlling cooldown rates should be provided.
Reference should be made to these methods with respect to the dominant PTS risk sequences whenever possible. -
Guidance should be provided for the operator if cooldown rates or P-T limits are exceeded. These guidelines should take into account !
.i potential core cooling, environmental release or containment l integrity problems which could exist as a result of responding to the abnormal cooldown rate. These guidelines should leave little i doubt as to when PTS concerns are more important than other safety ?
. ._ - . . _ =- m . .
e 42 -
1 issues, and when other safety issues assume primary importance ove'r PTS concerns.
The desired region.of operation between the pressure-temperature limit and the limit determined by avoidance of saturation condi- [
tions should be evaluated to determine if it can be revised to i minimize total risk due to plant operation from PTS plus non-PTS j
' events.
Instructions for controlling pressure following depressurization transients should be provided.
t Instructions should be available for the condition where natural ;
circulation is lost and the primary system main circulation pumps are not available.
i 8.3. Inservice Inspection and Nondestructive Evaluation (NDE) Program - 1
?
The utilization of state-of-the-art nondestructive evaluation techniques '
could provide an opportunity to decrease any conservatism which might exist !
in the flaw density value used in the analysis. This decrease in f
conservatism, however, may be less important than the decrease in uncertainty {
in the actual flaw density which may result from an evaluation of this type. !
F Existing inservice inspection programs should be reevaluated to consider incorporation of state-of-the-art examination techniques for inspecting *the clad-base metal interface and the near-surface area. This includes plant-unique consideration of the clad surface conditions. Consideration should be given to increased frequency of inspections.
I I
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- The reliability of the NDE method selected to detect small flaws should be documented.
8.4 Plant Modifications t
Plant modifications that may be considered include the following: -
M
'1) Instrumentation, Controls ~, and Operation a) reactor vessel downcomer water temperature monitor b) instantaneous and integrated reactor coolant system cooldown rate monitors -l c) steam dump interlock '
-d) feedwater isolation / flow control logic '
e) reactor coolant system pressure and temperature monitors f)' monitor to measure margin between vessel inner surface temoerature and current RT NDT at that location ;
g) diagnostic instrumentation and displays i h) primary system coolant pump trip logic ,
i) automatic isolation of AFW to broken steamlines/ generators [
t
- 2) Increased Temperature of Emergency Core Cooling Water and Emergency
[
Feedwater i t
1 If plant modifications are proposed to prevent overcooling, the report should include an evaluation of undesirable side effects (i.e., undercooling) and a
. i.
discussion of steps planned to ensure that.the modifications represent a net improvement in safety, when PTS and non-PTS related events are considered.
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-8.5 In-Situ Annealina If in-situ annealing is part of the licensee's program of corrective measures, the licensee should describe the program to ensure that' annealing ,
will achieve the planned increase in vessel toughness, the surveillance {
program to monitor vessel toughness after annealing, and the program to 1 ensure that annealing does not introduce other safety problems.
9.0 PREDICTION OF VESSEL FAILURE MODE-i In development of the PTS rule (10CFR50.61), the NRC staff utilized analyses I that predicted the expected frequency of through-wall crack penetration due U to PTS events. It was implicitly assumed that a through-wall crack would !
likely cause a core melt, which was assumed to be an unacceptable event that should be prevented, regardless of the resulting risk in terms of man-rem, fatalities, etc.
Analyses that were performed subsequent to the rule-development tended to confirm the assumption that a through-wall crack is indeed likely to cause a i vessel failure massive enough to cause core damage (Reference 22 provides an example of these calculations). Therefore, the Acceptance Criteria in Section I.C use the through-wall crack frequency as the primary acceptance criterion, but consequence and risk calculations are also required to judge acceptability of continued plant operation above the screening limit. The required consequence and risk analyses are described in this chapter and in ,
Chapter 10.
i l
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^Q f This chapter should present methods that can be used to determine the mode of
. vessel failure that is expected to occur when a crack penetrates the wall of the vessel. The results to be obtained include the final length and opening area of the crack, the size and velocity of possible missiles, and ;
predictions regarding waether such missiles will be arrested. These results j are needed to predict the core melt conditional probability given i through-wall crack penetration, which in turn is needed as input to the risk calculation presented in Chapter 10. l Input for these analyses include the vessel wall temperature distribution and ,
pressure at the time of occurrence of the through wall crack, the probability .;
of existence of such cracks for each weld in the vessel, initial toughness
- parameters for each weld and plate in the vessel, and the fluence at each f
location in the vessel. All major contributing sequences to vessel failure [
should be addressed. !
l i
The output of the analyses will be the conditional probability for each of I the significant failure modes. Examples of such modes include the range from !
t crack arrest at the end of the affected axial weld to extension of the axial [
crack into adjacent plate material to a complete fracture of a j circumferential weld.
The vessel failure mode calculations should be performed with a probabilistic ,
fracture mechanics code such as described in Reference 22. As an I
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alternative, the licensees may conservatively assume the failure mode which has the greatest consequences on predicted risk.
The following methods and assumptions are described in more detail in Reference 22 and are acceptable to the NRC staff:
The initial length of all axial cracks is the full length of the axial I
weld.
All circumferential cracks should extend the full 360 degrees around the vessel, unless an analysis can justify arrest at a shorter length. ,
The initial state of each through-wall crack in the vessel should be treated as a running crack that must be arrested to preclude further k 4
i crack propagation.
i Thermal stresses can be neglected in analyses of propagation of !
through-wall cracks once the crack has initiated. l t
- The effect of fluence and temperature gradients can be treated.by calculating an average toughness for the vessel wall by a root mean square approach.
The effects of fluid structural interactions and structural dynamics for I opening cracks should be considered. However, simplifying approximations as described in Reference 22 may be used. ,
The effects of irradiation on ductile brittle transition temperature, upper shelf toughness and flaw stress should be included in any l analyses. The methods should be consistent with those used to predict I the growth of part-through flaws. 1 i
l.
I
- Elastic plastic fracture mechanics should be used for growth predictions of through-wall cracks in the vessel. Acceptable methods and approximations are described in Reference 22.
The possibility that axial cracks may turn and follow circumferential welds should be part of the failure mode analyses.
In the calculations of probabilities for vessel failure modes, the random variations in the parameters governing material toughness can be assumed to be independent for the different welds and base metal that make up the vessel.
The masses and velocities of vessel fragments that become missiles should be estimated. Analyses should be performed to determine if such f missiles can escape from the vessel cavity, and if such missiles can .
cause loss of containment integrity. ;
r i
Core movement and deformation caused by missiles should be examined to evaluate core damage and probability of core melt, f ,
The methods and assumptions used to predict vessel failure modes should be {
described. The output of the analyses should describe the location and size '
of the final opening of the through-wall crack.
f, The licensee should document the sources of vessel material characteristics such as chemistry and value of toughness properties for the unirradiated [j vessel. Uncertainties in these properties should be discussed. Conservative '
values should be assumed when data are lacking.
{
Analyscs have shown that axial cracks may very often turn and produce a fracture of circumferential welds. In the analyses of vessel failure modes, ,
i r
crack tip stress intensity factors have been estimated for those cracks that may turn and follow a circumferential weld. The stress intensity factor decreases by a factor of two when the crack changes direction from axial to circumferential.
Fluid-structural interactions can be approximated in the analyses of reactor pressure vessels by the use of static stress intensity factor solutions.
However, one must assume that the full pressure remains acting during the static crack opening process.
For.the long through-wall cracks of concern to the failure mode analyses, the use of elastic plastic fracture mechanics is required. When cracks run -
beyond the irradiated zone into ductile materials with good upper shelf toughness, crack arrest can often be predicted if one takes credit for the increased tearing resistance associated with stable crack growth.
An important part of the evaluation is to assemble data which describe the chemistry and initial toughness of all welds and base metal in the vessel j shell. (These data are generally available to the licensee or through !
industry data bases.) When vessel specific data are lacking, it will be l necessary to make conservative estimates based on data available for similar !
vessels, i
i It will be difficult to estimate the size of vessel fragments that can become missiles. For horizontal missiles, it may be possible to apply ballistic penetration equations to predict missile arrest by the vessel cavity , ,
, concrete. It may be found that the upper bound velocities for the spectrum of missile sizes show consistent arrest. The other class of missile of I
concern is that of the upper head assembly. For these missiles, the restraint provided by the attached primary coolant piping may be the major f
factor contributing to arrest of the missile. !
- _ - _ - _ - I
The prediction of vessel failure modes may be performed on a probabilistic basis using Monte Carlo simulations. The approach may use the probabilistic fracture mechanics codes for prediction of through-wall cracks as a guide.
For the failure mode, the calculations must involve through wall rather than surface cracks. The toughness of each weld and plate in the vessel should be simulated. The step changes in toughness from material to material are the basis for predicting the arrest of the through-wall crack.
10.0 LIKELIHOOD OF CORE MELT AND PREDICTION OF RISK 10.1 Risk Analysis A factor to be considered in determining the acceptability of operation beyond the screening criterion will be the estimated risk in terms of the frequency of core melt due to PTS-induced failures, multiplied by the consequences in terms of public exposure to radiation. To calculate risk in
, these terms, one must understand the core melt frequency due to PTS-induced failure, the effectiveness of containment, the behavior of radionuclides !
within the primary system and containment and the transport of radionuclides !
beyond the site boundary. This chapter gives guidance for the estimation of !
those quantities.
10.2. General Guidance f
The purpose of severe accident analysis is to obtain a realistic estimate of risk. Consequently it is appropriate to use best-estimate methods, data and assumptions. However, because many of the phenomena are not well understood, it is necessary to account for the impact of analysis uncertainties on the conclusions being drawn from the analysis results, t
i r
Analyses should be plant and site specific. Generic results may be used if the generic data are shown to be applicable or bounding for the plant under analysis.
Many of the phenomena that govern severe accident risk are currently undergoing intense study by the NRC Severe Accident Research Program (SARP), -
EPRI and foreign research programs. Because no existing plants are expected to approach the screening criterion in the near future, it is likely that much of the ongoing research will be complete when the first PTS risk calculation is performed, which may result in the availability of better information and/or calculative methods for use in those first calculations.
The discussion below describes the methods and assumptions that currently are
- acceptable to the NRC staff and identifies areas where ongoing research is o likely to change current methods. The licensee should use the methods and assumptions accepted by NRC at the time of the analysis in order to minimize the review and approval process, 10.3. Core Melt Frequency The frequency of through wall cracks in the reactor vessel should be determined from thermal-hydraulic, metallurgical and fracture mechanics analysis according to the guidance described in the sections above. The frequency of through-wall vessel cracks should be estimated as a function of ,
crack location and size, including the case of a complete crack around a circumferential weld. [
- {
Based on break location and size, the licensee should determine whether or not the pressure vessel will depressurize, because reactor vessel pressure can affect containment performance.
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l The licensee may choose to assume that all through-wall cracks lead to core melt. Alternatively, best-estimate ECCS performance analyses may be submitted to demonstrate that one or more classes of cracks would not lead to coce melt.
Based on these analyses, the licensee should determine the frequency of core melts (per reactor year). Given a core melt, the timing and mode of containme'nt failure depend on the nature of the accident sequence and the status of containment protective systems at the time of core melt. Together, these factors determine the plant damage state.
For PWRs, the principal containment safety feature is containment heat removal by_ sprays and fan coolers. Operation of the sprays or fans can delay
)j' or prevent containment overpressure failure. Furthermore, the sprays, and to ,
a lesser extent the fan coolers, can reduce the suspended fission product inventory in containment. In determining the probability of fan and spray operation, the licensee should account for multiple independent failures and -
common cause failures such as sump blockage or clogging of the fan cooler heat exchangers by aerosols. The staff's current perception is that operation of sprays and/or fans is highly likely for those sequences involving potential PTS concerns.
i The possibility exists that containment will be bypassed by a consequential break in an injection line outside of containment or by a steam generator tube rupture. The former may be postulated to result from violent motion of the primary coolant system during blowdown via a through-wall crack. The 4
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l latter can also result from primary system motion, but in addition, may have been the transient which initiated the PTS event.- There is also preliminary evidence that tube ruptures could result from the superheated steam produced during core meltdown. All plausible mechanisms should be accounted for in estimating the likelihood of containment bypass.
The staff's current perception is that containment bypass in'a PTS event is unlikely.,
10.4 Containment Failure Given plant damage state frequencies, the timing of containment failure and -
the magnitude of the radiological release depend on the response characteristics of the containment. These characteristics are numerically embodied in the containment failure matrix (C-Matrix).
Determining the C-Matrix requires a best estimate analysis of the failure pressure of containment. In general, this pressure will be higher than the design pressure.
Threats to containment integrity can come from many sources, including missiles, steam explosions, direct heating, failure to isolate, gradual '
overpressurization, hydrogen burns or detonations, preexisting or induced leakage, basemat meltthrough and the bypass modes discussed above. Each mode of failure should be assessed for each plant damage state. The probability i of gradual overpressure failures and hydrogen burns should be based on ,
plant-specific calculations of accident phenomenology in the reactor vessel and containment, using a methodology acceptable to the staff at the time the analyses are performed. ,
r l
The probability of leakage or failure to isolate can also be based on generic results, and should account for operational experience with containment isolation during normal operation. Basemat meltthrough should be assessed with a model acceptable to the staff at the time the analyses are performed for core-concrete interaction.
The output of the containment failure matrix is a set of probabilities of release categories which characterize the timing of containment failure and severity of the radiological release. In the WASH-1400 study, these core melt release categories are designated PWR-1 through PWR-7. The contributions from several plant damage states and containment failure modes can be lumped into a single release category, if their characteristics are similar.
10.5. Source Term The radiological source term associated with each release category specifies the fraction of each type of radionuclide released, the energy of release, the chemical form, the particle size, the release time, tha warning time and duration of release. Calculations of these parameters are based on estirates of the rate of release of fission products from the fuel, the transport of radionuclides in the reactor vessel and containment, and various mechanisms for deposition and removal from the atmosphere. Currently,thereisamajor research effort devoted to a better understanding of the radionuclide release fractions and release characteristics. It is possible in the future that the .
Commission will adopt a new methodology. ;
For the purpose of risk calculations for pressurized thermal shock, the t licensee should use the source term methods and assumptions that are acceptable to the Commission at the time of the analysis.
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10.6. Site Consequence Analysis ;
For each radiological release category, an assessment of offsite consequences should be performed. The CRAC-2 code for consequence calculations is an example of an acceptable methodology. It has been tested in numerous applicationsandsubjectedtointernationalpeerreview. The licensee should use a code which embodies an acceptable set of models for fission product transport' and deposition, emergency actions and dose response. The licensee is encouraged to use codes and methods accepted by the NRC at the time the calculation is performed. ,
l The licensee should use plant-specific data for weather and other site '
characteristics. A site specific population distribution should be used for the year with maximum population during the remaining projected life of the
! plant.
Consequences should be integrated over an appropriate area. The 1"icensee must justify the choice of warning time, emergency response delay time, radius of limitation, and speed of evacuation. These should be in agreement I
with an evacuation plan and model accepted by the NRC at the time the PTS analyses are performed.
The consequence calculation should be performed for numerous weather conditions and averaged based on the actual plan't meteorology. Results should include conditional point estimates of early fatalities, latent cancer fatalities and public radiation exposure (in person-res) for each release category. Results should be consistent with NRC acceptance criteria in [
effect at the time of the PTS analyses. '
10.7 Risk l
l The risk contributions for each release category should be calculated as the
! product of the release category frequency per reactor year and each of the i
l !
three consequences determined: early fatalities, latent cancer fatalities and person-rem. The overall pressurized thermal shock core melt frequency and risk is the sum of the release category contributions, and should be calculated using models approved by the NRC at the time the analyses are performed, and compared to acceptance criteria in effect at the time.
10.8 Uncertainties o
The uncertainty in risk reflects the uncertainties in every aspect of the sequence of calculations, including the through-wall-crack frequency, vessel failure mode, core melt frequency, containment response, radiological release l fractions and offsite consequences. The licensee is expected to make reasonable estimates of the uncertainties associated with each phase of the
'l '
calculation. Based on those uncertainties, the licensee should place reasonable upper and lower bounds on the risk estimates.
3 11.0 RESULTS AND CONCLUSIONS REGARDING PTS PISK This chapter summarizes the models used, the results obtained, and provides the conclusions reached with respect to continued operation of the plant.
11.1 Summary of Analysis InthissectionthemajorfindingsofeachaspectofthePTSanalvsis,as described in the previous chapters, should be presented. This should [
include:
)
i 5
. e l
l l Expected (mean)value of frequency of reactor vessel failure and of core melt vs time, with uncertainty bound (95th percentile).
l l
Identification of dominant accident sequences. If dominant '
l sequences are different for through-wall crack, core melt, and '
early release of major radioactivity, identify and explain i differences.
j If sensitivity / uncertainty analysis shows that slightly different ;
l assumptions could lead to different dominant sequences, identify ,
l these assumptions and discuss the impact on results given the different assumption. -
l Identification of important operator actions, control actions, and l plant features that can increase or decrease the frequency or .
consequences of overcooling transients.
l 1
- The likelihood that the accident can cause a missile to rupture l containment and cause core damage and/or disable containment l cooling.
i
!
- Major sources and magnitudes of uncertainty in the analysis l
The relative effectiveness of potential alternative corrective ,
measures in reducing the expected (mean value) of vessel failure, core melt, and early major release of radioactivity. ,
l
! i
- The program of planned corrective measures. ;
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11.2 Basis-for Continued Ooeration Finally, as part of the plant-specific analysis package, the licensee will provide a basis for concluding whether or not continued plant operation is ;;
justified. The basis for continued operation must include comparison with !'
t NRC's PTS acceptance criteria given in Section I.C of this Guide. f-i,
,1.
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. . _ . _ _ _ _ . . . _ . - . _ _ _ ~ _ _ . . . _ . . _ - - _ _ _ _ _ . - _ _ _ ~ _ , . , _ , . - . . _ . -
I
. III. REFERENCES
- 1. T. J.' Burns, et al., " Pressurized Thermal Shock Evaluation of the Oconee-1 Nuclear Power Plant", NUREG/CR-3770 (ORNL/TM-9176), to be published. ;
- 2. D. L. Selby, et al. , " Pressurized Thermal Shock Evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant", NUREG/CR-4022 .
(ORNL/TM-9408), to be published.
- 3. D. L. Selby, et al., " Pressurized Thermal Shock Evaluation of the'H. B.
Robinson Unit 2 Nuclear Power Plant", NUREG/CR-4183 (ORNL/TM-9567) to be published.
- 4. L. Potash, " Generation of the Confusion Matrix for Misdiagnoses",
Appendix A of the Oconee PRA report.
- 5. R. E. Hall, J. Fragola, and J. Wreathall, " Post Event Human Decision Errors: Operator Action Tree / Time Reliability Correlation",
- 6. A. D. Swain and H. E. Guttmann, " Handbook of Human Reliability Analysis With Emphasis on Nuclear Power Plant Applications", NUREG/CR-1278, (August 1983).
- 7. B. Bassett, et al., " TRAC Analysis of Severe Overcooling Transients for ;
the Oconee 1 PWR," NUREG/CR-3706. i
- 8. C. D. Fletcher, et al., "RELAP 5 Thermal-Hydraulic Analysis of PTS Sequences for the Oconee 1 PWR," NUREG/CR-3761.
i
(,
t I
i I
l 9. J. Koenig, G.'Spriggs and R. Smith, " TRAC-PF1 Analysis of Pctential PTS Transients at a Combustion Engineering PWR," NUREG/CR-4109.
- 10. C.- D. Fletcher et al., "RELAP 5 Thermal-Hydraulic Analyses of PTS Sequences for the H. B. Robinson Unit 2 PWR," NUREG/CR-3977.
- 11. C. D. Fletcher, C. B. Davis and D. M. Ogden, " Thermal-Hydraulic Analyses of 0'vercooling Sequences for the H. B. Robinson Unit 2 PTS Study,"
- 12. T. G. Theofanous, et al., " Decay of Buoyancy Driven Stratified Layers with Application to PTS," NUREG/CR-3700.
- 13. T. G. Theofanous, et al., " REMIX: Computer Program for Temperature Transients Due to High Pressure Injection in a Stagnant Leop,"
- 14. T. G. Theofanous, et al., " Buoyancy Effects on Overcooling Transients Calculated for the USNRC Pressurized Thermal Shock Study,"
- 15. Bart Daly, "Three Dimensional Calculations of Transient Fluid Thermal Mixing in the Downcomer of the Calvert Cliffs-1, Plant using SOLA-PTS,"
- 16. Martin Torrey and Bart Daly, "SOLA-PTS: A Transient 3-D Algorithm *for Fluid Thermal Mixing and Wall Heat Transfer in Complex Geometries," l NUREG/CR-3822.
[
- 17. F. X. Dolan, et al., " Facility and Test Design Report: 1/2 Scale .
Thermal Mixing Project," NUREG/CR-3426, Volumes I and II.
i i
I
m.
. 18. R. D. Cheverton and D. G. Ball, "0CA-P, A Deterministic and Probabilistic Fracture-Mechanics Code for Application to Pressure Vessels", NUREG/CR-3618 (ORNL-5991), (May 1984).
D. L. Stevens, F. A. Simonen, J. Strosnider, R. W. Klecker,
- 19. D. W.
Engel, and K.. I. Johnson, " VISA - A Computer Code for Predicting The
~
Probability of Reactor Vessel Failure", NUREG/CR-3384, (1984).
- 20. 'USNRC Regulatory Guide 1.99.
- 21. " Clarification of TMI Action Plan-Requirements. Requirements for Emergency Response (I.C.1, Guidance for the Evaluation and Development of Procedures of Transients and Accidents", NUREG-0737, (1983).
- 22. F. A. Simonen, " Vessel Behavior Following a Through Wall Crack," PNL SA12547, October 1984 (Presented at the 12th Water. Reactor Safety Research Information Meeting, October 1984, Gaithersburg, Md). .
- 23. SECY-82-465, Pressurized Thermal Shock (PTS), November 23, 1982.
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ENCLOSURE 2 1
REGULATORY ANALYSIS OF PROPOSED REGULATORY GUIDE FOR LICENSEE PREPARATION AND NRC STAFF REVIEW OF PLANT-SPECIFIC ANALYSES REQUIRED BY THE PRESSURIZED THERMAL SHOCK RULE, 10CFR50.61 The pressurized thennal shock (PTS) rule,10CFR50.61, requires collection and reporting of material properties data, analyses of f, lux reduction options, and detailed plant-specific PTS risk analyses for those plants that reach the' rule-specified, RTNDT
- based screening criterion during the term of the operating license. The proposed Regulatory Guide addresses the detailed plant-specific risk analysis requirement, providing recommendations regarding hcw licensees should perform, and how the NRC staff should review, those analyses.
Neither the PTS .ule nor the proposed Regulatory Guide requires specific corrective actions. The proposed Guice merely provides guidance for the performance of the analyses required by the rule to identify and select necessary corrective actions. Therefore, in accordance with the Comission's Regulatory Analysis Guidelines (NRUEG/BR-0058, Revision 1) this Regulatory Analysis does not provide extensive and detailed assessment of required, specific corrective actions.
i The background material, nature of the problem, objectives and costs, etc. of the PTS rule's recuirements are covered in the Regulatory Analysis prepared as part of the rulemaking proceeding (Enclosure 3 to SECY-83-258, proooted Pressurized Thermal Shock (PTS) Rule, July 15. 1983, and Enclosure O to SECY-85-60 Final Pressurized Themal Shock (PTS) Rule. February 20, 1585),.
This Regulatory Analysis therefore addresses only: (1) the ,need for publishing guidance regarding how licensees should perform tne PTS rule-required plant-specific analyses; (2) the apprcpriateness of this particular guidance, and (3) the basis for the NRC-staff-acceptance criteria provided in the subject Guice.
' Reference Temperature for the Ni? Cuctility Transitice. a treasure of the temperature range where tne matterials' ductild ty cnanges most P!picly with changes in temperature.
p '
l (1) Need for Guidance The NRC staff has obtained considerable experience concerning PTS risk :
analyses. This experience has come from performance of analyses by the staff, from prototype plant-specific analyses performed by National =t Laboratories and sponsored by NRC, and from review of industry-sponsored f) analyses. The proposed Regulatory Guide reflects the lessons learned from this experience. Availability of the Guide will aid licensees in ':
perfoming analyses that will efficiently derive risk estimates in the ,
form the NRC needs for use in evaluating their conformance with the regulations. .l,
'i This need for guidance is particularly acute since the plant-specific l PTS analyses should utilize a probabilistic risk analysis (PRA) I approach, as opposed to the more traditional design basis accident'(DBA) ,
approach, as explained below. l The PTS risk is developed as the sum of the small risks resulting from j each of a large number of possible (but unlikely) PTS events. The ;
proposed Regulatory Guide accordingly describes acceptable :nethods to l identify as many as possible of the potential PTS events, group them, i
- calculate tne frecuencies and consequences of each group, determine the risk due to each grcup by multiplying the predicted frequency by the i calculated consequences, and then sum the results from all groups to ;
obtain total PTS risk estimates that can be compared with the acceptance (!
criteria given in the proposed Regulatory Guide. ,
The 08A approach, on the other hand, would attempt to define a worst credible event (the " design-basis-accident)" and then show that: (1) consecuences frem that event are acceptable and (2) 411 other crscible d I
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events are less severe and therefore acceptable. The staff has detemined that this DBA approach is not appropriate for plant-specific PTS analyses, because the total risk from all credible PTS events can be significant even though each event individually is less severe then the "DSA." The NRC staff therefore believes that it is necessary to publish this Guide so that licensees will utfifre the acceptable PRA approach and not waste time and resources on the more traditional DBA approach.
(2) iustification l of this Particular Guidance The NRC staff has performed prototype plant-specific analyses for three plants. They constitute the most detailed, thorough analyses perfomed j
to date and the lessons learned in their perfomance is reflectec in the Guide. The NRC staff has incorporated into the Guide descriptions of l the best methods found regarding how to assemble details of a plant's design (and to what level those details should be included), how to use event tree methodologies to identify and group potential PTS events, how to calculate severity of the events, how to integrate the resulting risk, anc many other subjects. The staff believes that the benefit of this experience is presented in this Guide, and its use by licensees will enable them to avoid many of the falso starts and errors made by the staff and their contractors in per#oming the prototype analyses, ,
thereby saving time and resources.
e (3) Justification of Acceptance Criteria The Guide states that in judging the acceptability of continued operation beyond the PTS screening criterion, the staff will assess wnether the results of the analyses: (1) indicate a predicted I
- in 'recuency of reacter vessel tnrcugn wall crack eenetratici less tnan 5 x '0 6 per reactor year, sne W are consiste*: with .a '1C's rica
~
criteria 1.1 use at :ne time the aral /sis is cerfor-ed ea7seding ccas roit frecuency, person-rsn exposure and fatalities.
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The mean frequency of reactor vessel tnrough-wall crack penetration is included as the principal acceptance criterion because the staff's analyses predict that there is a high likelihood of core damage in the event of such cracks. Core damage events have potential public health and safety consequences that are difficult to analyze with certainty, and also would have severe economic impacts upon the licensee and the public who will pay for cleanup and replacement power. For all of these reasons, reactor vessel tnrcugh-wall crack penetration frecuency is included as the principal acceptance criterion. The particular value of S x 10'0 mean frecuency per reactor year was ,
selected as an achievable, realistic goal that will result in an acceptable level of risk. It is believed that this value is acceptably low considering that pressure vessel failure is not part of the design basis of the plant, and therefore must have a frequency low enough to be consid2 red incredible. When the various (unquantifiable) biases that are inherent in the analyses are taken into account at least qualitatively, such as the implicit assumption that " core damage" is equivalent to
" core melt", this value probably results in a core melt mean frequency c1cse to one per million reactor years.
Additionally, the staff believes that in orcer to make a thorougn assessment of risk due to PTS events, the analyses shculd include prediction of risk in terms of predicted core melt frequency ,
person-rem exposure, and early and late fatalities. This is bec1use, given a reactor vessel threugh-wall crack, the vessel failure ecces will
<ary from plant-to-plant, and given vessel failure and core melt, ,
contattrent performance and resulting aisk to the surrounding site !
specific population density will be different. Therefore a risk ;
assessment in those standard terms is required. Since the flRC's safety I l
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j geals are currently being ceveloped, it is possible at this tir.e enty to state that the results of the plant-specific PTS risk analyses must be acceptable under whatever goals are in place or in general use by the NRC at the time the plant-specific PTS analyses are required.
j In the opinion of the NRC staff, there are no other practical quantities on which to base the acceptance criteria other than reactor vessel through-wall cracks (i.e., vessel failure) and/or the standard risk ,
- reasures of core melt frequency, person-rem exposure, and fatalities.
Analysis Pecuired to Support Imposition of '
Infonnation Collection Requirements.
The pressurized thermal shock (PTS) rule,10CFR50.61, requires collection and repcrting of material properties data, analyses and reporting of flux' reduction options, and submittal of detailed plant-specific PTS risk analyses for those plants that reach the rule-specified RTND7* based screening criterion during plant life. The subject Regulatory Guide only addresses the detailed plant-specific risk analysis requirement, providing recomendations regarding the exact details of how licensees should perfom and NRC staff Shculd review those analyses. ,
- Reference Temperature for the Nil Ductility Transition, a measure of tne temperature range where the materials' ductility changes nest rapidly with j changes in tempera *ure.
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L All three of the above listed, rule specified data reporting requirements, g
including the detailed plant-specific analysis submittal addressed by the subject Regulatory Guide, were discussed in the Analysis Required to Succort
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,' l Imposition of Information Collection Reouiwts portion of the Regulatory ~
Analysis written earlier for the proposed PTS rule (Enclosure 8 to SECY '
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288, Presssed Pressurized Thermal Shock (PTS) Rule. July 15,1983). That f document resulted in OMB approval of all thme data collection requirements of the rule (September 12, 1983. OMB No. 3150-0011). .
'l The subject Regulatory Guide provides detailed guidance regarding the ,
performance of the detailed plant-spircific PTS analyses required by the FTS
't rule (10CFR50.61) and whose imposition was already approved as stated above.
l The detailed guidance provided in the subject Regulatory Guide will result in a submittal from the licensee that 1: well within the cost and time estintes ( )
pmvided in the above referenced dccument that resulted in the OM8 approval.
Consecuently, we incorporate Enclosure 8 to SECY-83-288 into this document i[
as the (already acproved) justification for the infomation reporting [
requirements of the subject Regulatory Guide.
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LNCLOSURE 3 Background Infomation for CRGR Review of U5I A-49 Final Resolution ,
The following informati:n is provided in the tormat specified in NNN Office '
2 Letter No. 39. Revision 1. "NRR Procedures for Control and Review of Generic Nequirements," dated December 15, 1982. For each item, the request for ,
infomation is given followed by a discussion of the response or a reference I Further supporting infomation is to where tne infomation is provided.
contained in Enclosure 2 " Regulatory Analysis for USI A-49."
(1) The proposed generic requirement as it is proposed to be sent out to "l Itcensees. :
1 The generic requirements are set forth in tne Pressurfred Themal Shock .
(PTS) rule, 10 CFR 50.61. The enclosed proposed Regulatory Guide !
(Enclosure 1) provides guidance regarding how to perfom the analyses ;
requirac by 10 CFR 50.61(b)(4). t (2) Draft staff papers or other underlying staff documents supporting the f
proposed requirements.
The relevant technical info matien is contained in the Procesed Reg. l Guice (Enclosure 1), the Regulatory Analysis (Enc 1csure 2) and related y references listed therein. Copies of any references will be provided !
upon request. ,
(3) A brief description of each of the steps anticipated that licensees i must carry out in order to complete the requirements.
I Are there' separate st. ort-:em requirements? {
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There are no separate short-term requirements associated with the subject Reg. Guide. (Short term requirements of other portions of the PTS rule, not related to this Reg. Guide, were approved by .
CRGR at meetings #38 and 70.) b l
Is it the definitive, comprehensive position on the subject or is ,
e it the first of a series of requirements to be issued in the future?-
The suDject Reg. Guide ano the PTS rule for which it is intended ,
to provide guidance represent the final statt position on USI :
A 49. It is possible, however, that significant new infonnation 4 could, at some future date, require consideration of amendments to i the rule itself or issuance of supplements to the Reg. Guide.
-k How does this requirement affect other requirements? Does this i requirement mean that other items or systems or prior analyses
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need to be reassessed? l h
This reouirement applies to all PWRs. It will not affect other I' I
items or systems of prior analysis. t I.
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,Is it only computation? Or does it require or may it entai: l engineering design of a new system or modification of any existing [
systems. [
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% i Depending on the outcome of the plant-specific analyses outlined l by the subject Reg. Guide, the PTS rule may require corrective I actions entailing engineering design of a new system or '!)
7.odification of existing systems. However, the subject Guide only l recu1res c:moutation. ;
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-3 Is plant shutdown necessary? How long?
Plant shutdown is not necessary to perform the analyses outlined by ,the subject Guide, e
t Does design need NRC approval?
- i The analyses outlined by the Guide (required by the PTS rule) need NRC approval. The NRC will review any equipment modifications identified as needed by the analyses on a case-by-case basis.
Does it require new equipment? Is it available for purchase in sufficient quantity by all affected licensees or must such equipment be designed? What is the lead time for availability?
This question will be answered by the analyses outlined in the
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1 (4) Identification of the category of reactors to which the generic requirement is to apply. 7 I
The proposed generic requirements apply to all FWR plants.
t (5) For each such category-3:
A risk reduction assessment performed us ng a cata base and ,
methodology ccmmonly accepted within NRC (for example, similar to f
'that outlined in SECY-81-513). f t
i Risk analyses are discussed in Regulatory Analysis (Enclosure 2).
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4-An assessment of costs to NRC, and assessment of cost to licensees, including resulting occupational dose increase or decrease, added plant and operational complexity, and total financial costs.
An estimate of utility costs to comply with the proposed analysis
, requirement and an estimate of costs to the NRC are presented in the Regulatory Analysis (Enclosure 2). Costs to implement any necessary corrective actions identified by the analysis will vary from plant to plant depen, ding on the results of the required ;
analyses.
- Consistent with the first two items above, provide the basis for ;
requiring or permitting implementation by a' given date or on a I particular schedule.
The implementation schedule was previously approved by CRGR at meetings #38 and 70, as it is part of the rule itself. I t
i Other suggested implementation schedule and the basis therefor. [
This should include sufficient infonnation to demonstrate that the j.
schedule is realistic and provides sufficient time for indepth p engineering, evaluation, design, procurement, installation, p testing, development of operating procedures, and tra1ning of L operators.
I Previously discussed, as above. No alternate implementation schedule is suggested.
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e Schedule for staff actions involved in completion of requirement (based on hypothesi 6d effective date of approval).
The staff will review the analyses within a reasonable time following submittal as required by 10 CFR 50.61. Analyses will be ,
due 3 years before each plant reaches the RTPTS screening
, cri terion. Exact dates are not now known - short term requirements ;
of the rule will provide a better estimate.
Prioritization of the proposed requirement considered in light of all other safety-related activities underway at all affected ,
facilities. This prioritizaticn shall be based on the guicance I
and direction issued from time to time by DEDROGR. Until such time '
as such advice is provided, each proposing office shall use its best technical judgment and explain the' basis therefor. ,
I Imolementation of the proposed requirement is judged to be high ,
priority based on staff assessment of the safety significance. !
t For proposed requirements involving reports and/or record keeping, i an assessment of whether such reporting or record keeping is tne j best means of implementation and the apprcpriate degree of i fonnality and detail to oe imposed. {
The staff believes that the requirements involving reports and/or record keeping as specified in the guide are kept to the minimom to involve only those which are absolutely necessary.
To the extent that the category contains plants of different types t
or vintages, the items listed above shall be provided for each .
. .ype and vintage, or justification snall be provioed cemonstrating that the analysis of eacn item is valid f r all types and vintages e
covered.
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Inese requirements apply to all PWR plants.
l (6) Each proposed requirement shall contain the sponsoring Office's position as to whether the requirement implements existing regulations or goes beyond them.
The proposed Reg. Guide provides guidance regarding how licensees should meet the existing regulations in 10 CFR 50.61. Requirements are within the existing regulations fcund in 10 CFR 50.61.
(7) The proposed method of implementation along with the cencurrence (and any coments) of ELD on the method proposed.
ELD will be asked to review the final version of the proposeo Guide after resolution of public comments and preparation of the final CRGR
) package. ;
i (8) Regulatory Analysis sufficient to adcress the Paperwork Reduction Act, I the Regulatory Incemnability Act, and Executive Order 12291.
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CMB clearance has been obtained as noted in the Regulatory Analysis i (Enclosure 2). f
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