Regulatory Guide 1.154

(Task SI 502-4) Format & Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors

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U.S. NUCLEAR REGULATORY COMMISSION

January 1987

/"

"REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 1.154 (Task SI 502-41 FORMAT AND CONTENT OF PLANT-SPECIFIC

PRESSURIZED THERMAL SHOCK SAFETY ANALYSIS

REPORTS FOR PRESSURIZED WATER REACTORS

USNRC REGULATORY GUIDES

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Table o f Contents INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Background and Purpose o f This Guide . . . . . . . . . . . . . .

Objectives o f Plant-Specific PTS Safety Analysis Reports . . . .

Staff Review o f Plant-Specific PTS Safety Analysis Reports and Acceptance C r i t e r i a Tor Continued Operation . . . . . . . . .

Recommended Format . . . . . . . . . . . . . . . . . . . . . . .

CHAPTER 1 OVERALL APPROACH. SCOPE OF ANALYSIS. AND REPORT

ORGANIZATION . . . . . . . . . . . . . . . . . . . . . .

CHAPTER 2 PLANT DATA . . . . . . . . . . . . . . . . . . . . . . . .

2.1 Systems Pertinent t o PTS . . . . . . . . . . . . . . .

2.2 Reactorvessel . . . . . . . . . . . . . . . . . . . .

. . . . . . . . . . . . . . . . . . . . . . .

2.3 Fluence

. . . . . . . . . . . . .

2.4 Inservice Inspection Results

2.5 Plant Operating Experience . . . . . . . . . . . . . . .

2.6 Operating Procedures . . . . . . . . . . . . . . . . .

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CHAPTER 3 DETERMINATION OF DETAILED PTS SEQUENCES FOR

ANALYSES . . . . . . . . . . . . . . . . . . . . . . . . .

3.1 Approach Used . . . . . . . . . . . . . . . . . . . . .

3.2 Sequence Delineation . . . . . . . . . . . . . . . . .

. . . . .

3.2.1 Development o f Classes o f I n i t i a t o r s

3.2.2 I d e n t i f i c a t i o n o f Important I n i t i a t o r Variations . . . . . . . . . . . . . . . . . .

3.2.3 Definition o f Potential Transients

. . . . . . . . .

Resulting from Each I n i t i a t o r

3.3 Operator Effects . . . . . . . . . . . . . . . . . . .

3.4 Sequence Quantification . . . . . . . . . . . . . . .

3.4.1 I n i t i a t i n g Events . . . . . . . . . . . . . . .

3.4.2 Equipment Failures . . . . . . . . . . . . . .

3.4.3 Operator Actions . . . . . . . . . . . . . . .

. . . . . . . . . . . . . . . . .

3.5 Event Tree Collapse

. . . . . . . . . . . . . .

3.5.1 Specific Sequences

3.5.2 Residual Groups . . . . . . . . . . . . . . . .

Page v

v i v i i i iii

Tab1 e o f Contents (Continued)

Page CHAPTER 4 THERMAL-HYDRAULIC ANALYSIS . . . . . . . . . . . . . . . .

4.1 Thermal-Hydraulic Analysis Plan . . . . . . . . . . .

4.2 Thermal-Hydraulic Model

. . . . . . . . . . . . . .

4.3 Simp1 i f i e d Analysis Methods . . . . . . . . . . . . .

4.4 Thermal S t r a t i f i c a t i o n Effects . . . . . . . . . . . .

4.5 Thermal -Hydraul i c Analysis Results . . . . . . . . . .

CHAPTER 5 FRACTURE MECHANICS ANALYSIS . . . . . . . . . . . . . . .

CHAPTER 6 INTEGRATION OF ANALYSES . . . . . . . . . . . . . . . . .

CHAPTER 7 SENSITIVITY AND UNCERTAINTY ANALYSES OF THROUGH-WALL CRACK

FREQUENCY . . . . . . . . . . . . . . . . . . . . . . . .

7.1 S e n s i t i v i t y Analysis . . . . . . . . . . . . . . . . .

7.2 Uncertainty Analysis . . . . . . . . . . . . . . . . .

7.2.1 Parameter Uncertainties . . . . . . . . . . . .

7.2.2 Model ing Uncertainties (Biases) . . . . . . . .

CHAPTER 8 EFFECT OF CORRECTIVE ACTIONS ON VESSEL THROUGH-WALL

CRACK FREQUENCY . . . . . . . . . . . . . . . . . . . . .

8.1 Flux Reduction Program . . . . . . . . . . . . . . . .

8.2 Operating Procedures and Training Program

. . . . . . . . . . . . . . . . . . . . .

Improvements

8.3 Inservi ce Inspecti on and Nondestructive Examination Program . . . . . . . . . . . . . . . . .

8.4 Plant Modifications . . . . . . . . . . . . . . . . .

8.5 I n S i t u A n n e a l i n g . . . . . . . . . . . . . . . . . .

CHAPTER 9 FURTHER ANALYSES . . . . . . . . . . . . . . . . . . . .

CHAPTER 10 RESULTS AND CONCLUSIONS REGARDING PTS ANALYSES . . . . . .

10 .. 1 Summary o f Analysis . . . . . . . . . . . . . . . . .

10.2 Basis f o r Continued Operation . . . . . . . . . . . .

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

REGULATORY ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . .

INTRODUCTION

/

Background and Purpose o f This Guide The pressurized thermal shock (PTS) rule, § 50.61 o f 10 CFR Part 50

issued on July 23, 1985 (50 FR 29937), establishes a screening c r i t e r i o n based on reactor vessel n i l - d u c t i l i t y - t r a n s i t i o n temperature (RTNDT)

The screening c r i t e r i o n was establ ished a f t e r extensive industry and NRC analyses regarding the likelihood o f vessel f a i l u r e due t o PTS events i n pressurized water reactors (PWRs).

The analyses were applied generically and contained conservative assumptions t o make the results bounding f o r any PWR.

Based on the results, the NRC concluded t h a t the r i s k due t o PTS events i s acceptable a t any p l a n t so long as the RTpTSX o f the reactor pressure vessel remains be1 ow the screening c r i t e r i o n .

Extensive safety analyses are required by the r u l e for any p l a n t t h a t wishes t o operate with RTpTS values above the screening c r i t e r i o n .

The recom- mended methods t o be used i n performing the analyses are outlined i n t h i s guide.

The purpose o f the analyses i s t o assess the r i s k due t o PTS events f o r proposed operation o f the p l a n t w i t h reactor vessel RTpTS above the screen- ing c r i t e r i o n .

Effective 1 year a f t e r the pub1 i c a t i o n o f t h i s regulatory guide, Section 50.61 requires t h a t these analyses be completed 3 years before the screening c r i t e r i o n would be exceeded t o allow adequate time f o r implementation on the p l a n t o f any corrective actions assumed i n the analyses before the plant operates above the screening c r i t e r i o n .

,

This regulatory guide describes a format and content acceptable t o the NRC

s t a f f f o r these plant-specific PTS safety analyses and describes acceptance c r i t e r i a t h a t the NRC s t a f f w i l l use i n evaluating licensee analyses and pro- posed corrective measures.

The references l i s t e d i n t h i s guide include a set o f analyses sponsored by the NRC that, taken together, constitute an example o f the analyses described by t h i s guide.

The s t a f f recommends t h a t these references be extensively used, along w i t h t h i s guide, by those performing the plant-specific PTS analyses re- quired by the PTS rule, § 50.61.

References 1, 2, and 3, f o r example, each represent an analysis by the Oak Ridge National Laboratory (ORNL) predicting through-wall crack frequency f o r one PWR.

These references w i l l provide guid- ance through the analyses.

Reference 3 (analysis o f H. B. Robinson) should be most helpful because it was the l a s t one performed and includes the experience gained i n performing the two e a r l i e r analyses.

Objectives o f Plant-Specific PTS Safety Analysis Reports Paragraph 50.61(b)(4)

requires t h a t a licensee whose p l a n t w i l l exceed the screening c r i t e r i o n before expiration o f the operating license submit safety analyses t o determi ne what, i f any, modi f i c a t i ons t o equipment, systems, and

• To avoid confusion among several (preexisting) s l i g h t l y d i f f e r e n t d e f i n i t i o n s of RTNDT, !$

50.61 contains i t s own d e f i n i t i o n o f an RTNDT (called RTpTS) t o be used when comparing plant-speci f i c vessel materi a1 properties w i t h the PTS

screening c r i t e r i o n .

operation are necessary t o prevent potential f a i l u r e o f the reactor vessel as a r e s u l t o f postulated PTS events i f continued operation beyond the screening c r i t e r i o n i s a1 lowed.

These analyses must include the effects o f a1 1 corrective actions the licensee believes necessary t o achieve an acceptable PTS-related r i s k f o r continued operation o f the plant.

The f i n a l objective o f the plant- specific PTS study, therefore, i s t o j u s t i f y continued operation o f the plant by demonstrating t h a t the 1 i kel i hood o f a through-wal 1 crack during continued operation i s acceptably low.

The study must include calculation, f o r the re- mainder o f plant 1 i f e , o f the expected frequency o f through-wall cracks due t o PTS .

I n calculating these results, it w i l l be necessary to:

O

I d e n t i f y the dominant accident sequences.

O

I d e n t i f y operator actions, control actions, and plant features impor- t a n t t o PTS.

O

Estimate the effectiveness o f potential corrective actions i n reduc- i

ng the expected frequency o f through-wal 1 cracks.

O

I d e n t i f y the sources and approximate magnitude o f the major uncertain- t i e s and t h e i r effects on the conclusions.

O

Present and j u s t i f y the licensee's proposed program f o r corrective measures.

I

\\

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Present and j u s t i f y the licensee's proposed basis f o r continued opera- t i o n a t embrittlement level s above the screening criterion.

This must include comparison with the acceptance c r i t e r i a described be1 ow o f the PTS-related through-wall crack frequency with corrective actions implemented as necessary.

S t a f f Review o f Plant-Speci f i c PTS Safety Analysis Reports and Acceptance C r i t e r i a f o r Continued Operation The PTS r u l e specifies a screening c r i t e r i o n based on RTNDT (called RTpTS

f o r use as defined within the rule) o f 270°F f o r axial weld and plate materials and 300°F f o r circumferential weld materials.

As detailed i n SECY-82-465 (Ref. 4), these values were selected based on generic studies o f the expected frequency and character o f a wide spectrum o f transients and accidents t h a t could cause pressurized overcool i n g o f the reactor vessel (PTS events) and on operating experience data.

The r i s k due t o PTS events was assessed i n terms o f probabilistic fracture mechanics calculations o f the expected frequency o f through-wall crack penetration o f the pressure vessel due t o the PTS events.

I n selecting the screening c r i t e r i o n based on those calculations, the conserva- t i v e assumption was made t h a t any through-wall crack could r e s u l t i n severe core degradation o r melt.

Core melt i t s e l f was viewed as an event t o be avoided even though r i s k t o the public due t o such an event i n terms o f person- rems and early and l a t e f a t a l i t i e s was not calculated with any certainty.

The estimated through-wall crack frequency developed as a function o f RTNOT f o r axial welds (Fig. 8.3 o f Ref. 4) i s shown i n Figure 1.

LONGITUDINAL CRACK EXTENSION NO ARREST

1 o - ~

SECY-82-465 PRA RESULTS

I

1

1

1 I

1 k

LEGEND:

MEAN SURFACE RTNDT(*F)

Figure 1

The RTpTS screening criterion selected by the staff corresponds to a mean I

(or average) "best estimate surface RTpTS of 210°F.

The staff used a "2-sigma"

-L

value (spread between "best estimate" and "upper limit") of 60°F;* thus the screening criterion expressed in terms of RTpTS, which, by definition, i s this upper limit value, was selected a t 210 + 60 = 270°F.

For axial weld and plate materials, Figure 1 gives a through-wall crack frequency of about 5 x per reactor year a t 210°F, which corresponds w i t h an RTpTS of 270°F.

For circum- ferential welds, the same frequency i s believed to be bounded by an RTpTS of approximately 300°F (Ref. 4).

The Commission concluded that the PTS-re1 ated risk a t any PWR i s acceptable so long as the RTpTS values remain below the specified screening criterion.

I t was realized that there are many unknowns and uncertainties inherent in the probabilistic calculations; thus it was w i t h deliberate intent that conser- vative assumptions such as those stated above were made.

The expectation was that the true risk a t any plant due to PTS events would in all likelihood be considerably below that derived from Figure 1 and would therefore be acceptable.

Also contributing to the belief that the real PTS risk a t any given plant was lower than that resulting from the analysis in Reference 4 was the belief that many of the generic plant assumptions made i n Reference 4 (e.g., material properties, system performance, crack distribution) would prove to be overcon- servative for analysis of a specific plant and that the resulting plant-specific analysis, when performed, i s likely t o result in a reduced prediction of PTS

risk.

If the plant-specific PTS analyses submitted by licensees in accordance with § 50.61 using the methodology described in this guide (or acceptable equi- valent methodology) predict that the PTS-related, through-wall crack penetration mean frequency will remain less than 5 x per reactor year for the requested period of continued operation, such operation would be acceptable to the staff.

In a1 1 the analyses performed, the licensee must justify that the impor- tant input values used are valid for the remaining 1 ife of the plant.

Recommended Format The recommended content of plant-specific PTS safety analyses i s presented i n Chapters 1 through 10 of this guide.

Use of this format by 1 icensees will help ensure the completeness of the information provided, dill assist the NRC

staff i n locating the information, and will aid in shortening the time needed for the review process.

If the Ticensee chooses to adopt this format, the numbering system of this guide should be followed a t least down to the section level.

Certain sections may be omitted i f they are clearly unnecessary to pro- vide for comprehension of the analysis or if they are repetitive.

RTeDT data from many plants (see Table P . l of Enclosure A

to Ref. 4).

viii

Additional guidance on style, composition, and specifications of safety i

analysis reports is provided in the Introduction of Revision 3 to Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Pl ants (LWR Edition). "

The Advisory Committee on Reactor Safeguards has been consulted concerni ng this guide and has concurred in the issuance of this regulatory guide.

Any information collection activities mentioned in this regulatory guide are contained as requirements in 10 CFR Part 50, which provides the regulatory basis for this guide. The information collection requirements in 10 CFR Part 50

have been cleared under OMB Clearance No. 3150-0011.

1. OVERALL APPROACH, SCOPE OF ANALYSIS, AND REPORT ORGANIZATION

,

This chapter is to describe the overall approach to the analysis and out-

1 i ne 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 inter- relationship of the tasks should be described and should be illustrated by a flow chart. How the analysis tasks are integrated to achieve the results and conclusions is to be described.

Major emphasis should be placed on analyzing event sequences leading to vessel through-wal 1 cracking and corrective actions to prevent this from occurri ng.

The report should include both probabilistic and deterministic fracture mechanics analyses. The probabi 1 i stic analyses should be used to determine the statistical 1 i kel i hood of vessel through-wal 1 crack penetration assuming a crack size distribution appropriately justified for the vessel being analyzed and appropriate uncertainties and distribution of the significant input param- eter such as material properties. The deterministic analyses should be used to evaluate the critical time interval in the transient during which mitigat- ing action can be effective. The deterministic analyses should be carried out using the two sigma upper and lower bounding values of the appropriate param- eters such as fluence, copper content, nickel content, fracture initiation toughness, fracture arrest toughness, and ductile fracture toughness.

The input to the probabilistic analysis should be best estimates based on appropriate assumptions. Uncertainties and conservatisms should be explicitly

,

presented in the decision rationale for the 1 icensee' s proposed corrective mea- sures and basis for continued operation.

The analysis should include effects of operator actions, control system interactions, and support systems such as electric power, instrument air, and service water cooling.

The report should be organized by starting with a description in Chapter 1 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 anal- yses performed, present the results obtained and conclusions drawn, and present and justify the licensee's proposed program of corrective measures.

Appendices should contain data, detai 1 ed models , sample calculations , and detailed results needed to support the various chapters of the report. Appen- dices should contain 1 i ttle supporting text. Instead, the nature and relevance of material in the appendices should be described in the pertinent chapters of the main report.

Throughout the guide, wherever it is specified or suggested that detailed descriptive materials should be submitted as part of the licensee's analyses, these detai 1 ed materials may be provided by incorporation of reference material already submitted to the NRC (for example, in the final safety analysis report).

It remains the responsibility of the licensee to provide a coherent, readable

document that does not unduly burden a reviewer with collecting extensive references before proceeding with the review. Therefore, care should be exercised in limiting such material provided by reference to the reviewer who is conducting an extensive, detai 1 ed eval uation of the submitted work.

Certain details (noted in Chapter 1 and in Section 4.3 of this regulatory guide) that have not been previously submitted to the NRC may be made available for NRC inspection and may also be referenced by the submitted analyses.

2.

PLANT DATA

This chapter i s t o b r i e f l y describe plant systems and operations pertinent t o PTS.

Chapter 2 o f Reference 3 (the H. 0. Robinson analysis by ORNL) provides a good example.

Supporting appendices o r references are t o present the design and operating data used i n the analysis o r needed t o understand the analysis.

References t o other dbcuments (e. g. , the f i n a l safety analysis report (FSAR))

should indicate specific sections.

( R e l i a b i l i t y data, however, are t o be i n Section 3.4, "Sequence Quantification," or i t s supporting appendices and references. )

2.1 Systems Pertinent t o PTS

Summarize design and operating features o f systems pertinent t o PTS.

I l l u s t r a t e each system with a simplified process and instrumentation diagram o r a single l i n e diagram.

I d e n t i f y on each i l l u s t r a t i o n any interfaces w i t h other systems.

For each system, include a table summarizing key design and operating data.

Give the maximum, minimum, and nominal values f o r those cases i n which design data may vary with time ( f o r example, high-pressure i n j e c t i o n (HPI) water temperature may vary with season).

Such values used i n the analysis should be i d e n t i f i e d and j u s t i f i e d .

Refer t o appendices o r other documents (e. g. , specific sections o f the FSAR) as necessary f o r more details.

Systems t o be considered should include pertinent portions of:

Reactor cooling system Condensate and main feedwater systems Steam system Auxi 1 i ary feedwater system Reactor protection system Chemical and volume control system Emergency core cooling systems Instrumentation and, control systems Support systems

- Electric power

- Instrument a i r

- Service cooling water

2.2 Reactor Vessel Summarize the reactor vessel construction and i t s material properties.

Use tables, drawings, or graphs t o show:

O

Vessel design (including weld locations and hot leg and cold

1 eg penetrations).

Vessel materials and chemical composition i n the b e l t l i n e region (including both base and weld material properties).

O

Vessel fabrication procedures, p a r t i c u l a r l y welding and cladding.

0

Vessel properties (e. g. , RTNDT, i n i t i a l RTNDT, appropriate fracture C

toughness data, including the upper-shelf regime, residual stresses, flaw density distribution, etc. ). Describe and j u s t i f y methods used t o calculate o r otherwise determine properties.

Available information on the vessel properties should be reexamined i n detail t o f i l l any gaps i n the supporting data f o r making an estimate o f RTNDT

and t o support resolution o f any disagreements about the v a l i d i t y o f values used.

Few data are currently available and validated t o support the selection o f a value f o r the i n i t i a l RTNDT The confidence t h a t can be placed i n estimates of the i n i t i a l RTHDT depends not only on material tests but also on the accu- rate documentation of we1 d i ng technique, weld wire used, and weld f l u x used.

The c r e d i b i l i t y o f such estimates could be enhanced by performing more tests on archival material, by discovering previously unreported t e s t results on weld specimens from the particular plant, o r by evaluating properties o f welds considered typical o f the p l ant-speci f i c we1 d.

2.3 Fluence Present (or incorporate by reference t o a submitted report) the current and projected fluence on the vessel using benchmarked computer programs and methodology and information from neutron f l u x surveillance dosimetry.

Use the weld locations and fluence values t o i d e n t i f y the c r i t i c a l welds.

Show how the

1. ,

fluence varies along the length and depth o f the c r i t i c a l welds.

Describe the basi s f o r these estimates and t h e i r uncertainty.

These f 1 ucnce val ues should be benchmarked, f o r example, through use o f ENDF/B-IV or V 1;ross sections, t o quantify the error.

Inservice Inspection Results To the extent pertinent t o the probabilistic analysis and proposed correc- t i v e actions, summarize:

O

Results - The number, size, depth, and location o f any flaws found should be we1 1 defined and described.

O

Methods used - The method used t o perform the inspection should be we1 1 described with documentation o f any val i d a t i on informati on.

Note:

Only those inservice inspections (ISIs) that have actually been per-

-

formed should be discussed i n t h i s section.

Improved I S 1 programs as proposed by the licensee should be described under corrective measures i n Chapter 8,

"Effect o f Corrective Actions on Vessel Through-Wall Crack Frequency."

2.5 P l ant Operati ng Experience Summarize overcool ing transients t h a t have occurred a t t h i s station and similar stations.

A1 so, summarize lessons learned from these and other tran- s i ents, and indicate actions taken t o prevent recurrence or m i nimize severity o f overcooling transients.

2.6 Operating Procedures

/

This section provides procedural data, e.g., what the operator i s supposed to do and when.

This section, for example, should present and describe the important operator actions as defined by existing procedures associated w i t h potential overcool i ng transients. A1 so emphasize how the procedures were evaluated and optimized i n light of any competing risks that might arise from events other than PTS events t o ensure that overall plant safety i s appropriately balanced.

The conditions under which the operator takes each action, the expected time for performing the action, and how the time was derived should be identified.

Some examples of these operator actions are:

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Trip reactor coolant pumps.

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Throttlehermi nate' emergency core cool ant.

O

Throttl e/termi nate main and emergency feedwater.

0

Restore main and emergency feedwater.

O

Isolate break (primary or secondary).

Supply a summary of training materials associated w i t h overcooling events in general and with respect to principal initiators.

In addition, a summary of simulator exercises associated with potenti a1 overcool i ng events should be provided.

Note:

Proposed improvements in procedures, diagnostic instrumentation, display

-

systems, and operator training should be presented in Section 8.2 under the

1 i censee' s program of corrective measures.

3.

DETERMINATION OF DETAILED PTS SEQUENCES FOR ANALYSES

This chapter i s t o present the methods and analyses used t o i d e n t i f y those transient sequences t h a t could contribute significantly t o the PTS r i s k .

A

good example i s presented i n Chapter 3 o f Reference 3.

The scope includes iden- t i f y i ng i n i t i a t i n g events, developing event trees, model ing and quantifying the r e l i a b i l i t y o f relevant systems and operator actions, and collapsin'g the event trees t o i d e n t i f y speci f i c re1 evant sequences.

Detai 1 ed models , data, and sample calculations should be included i n appendices o r referenced, However, the l o g i c o f the analysis, c r i t e r i a used, results, and insights gained are t o be described i n the main report.

3.1 Approach Used Describe how the material presented i n t h i s chapter f i t s i n t o the overall PTS study.

Provide a general description o f the process used t o i d e n t i f y PTS

sequences.

It should be made clear how the approach used w i l l r e s u l t i n com- pleteness o f i d e n t i f i c a t i o n o f a l l classes o f events t h a t could contribute sig- n i f i c a n t l y t o PTS risk, how specific events are selected f o r more detailed anal- y s i s t o represent each class, and f i n a l l y how the events so analysed are used t o determine t o t a l PTS r i s k a t the plant.

3.2 Sequence Del i neati on Identi f y potential overcool i ng transients i n a we1 1 -def i ned manner, and document them i n such a way t h a t it i s clear t o a reviewer t h a t a l l important potential overcool ing conditions have been considered.

Classes o f i n i t i a t o r s

1 should be developed, important variations o f i n i t i a t o r s within each class should be identified, and potential transients resulting from these i n i t i a t o r s should be defined.

Operating experience a t the specific plant and a t similar plants should be carefully examined t o a i d i n the i d e n t i f i c a t i o n o f potentially significant PTS i n i t i a t o r s , contributing f a i 1 ures, and potential corrective actions.

The ORNL contribution t o Systematic Eva1 uation Program reviews (Ref. 5, f o r example)

i s a technique that can be used f o r t h i s purpose.

3.2.1 Development o f Classes o f I n i t i a t o r s Any class o f transients t h a t could lead t o overcool ing o f the reactor ves- sel should be considered i n the analysis. It should, however, be appropriate t o use logical arguments t o eliminate classes o f transients as actual PTS

i n i t i a t o r s whenever justifiable. . Examples o f i n i t i a t o r s that should be included are:

O

Loss-of-coolant accidents ( LOCAs) , i ncl udi ng steam generator tube rupture accidents.

Steam 1 i ne breaks.

O

Overfeeds.

O

Combinations of these, i ncl uding possible return t o c r i t i c a l i ty.

3.2.2 I d e n t i f i c a t i o n o f Important I n i t i a t o r Variations

-

After the classes o f potential i n i t i a t o r s have been identified, it i s i m -

portant t o consider variations within any individual class. These variations should include:

1.

Decay heat level - The decay heat level, determined by recent operat- ing history o f the plant, can have a major impact on the potential consequences o f a given event.

Thus, various decay heat conditions should be considered.

Clearly, decay heat associated with a reactor t r i p from f u l l power (assuming operation a t f u l l power f o r some considerable time) should be examined.

Zero decay heat represents the opposite extreme but f o r a l l practical purposes occurs only once a t the beginning o f l i f e f o r the plant when PTS i s not important.

Therefore, the analyst may choose t o use some other level o f decay heat t h a t would cover potential decay heat conditions a f t e r the i n i t i a l startup o f the plant.

The reasons f o r choosing particular decay heat levels f o r analysis should be documented.

Each i d e n t i f i e d i n i t i a t o r should be examined a t a l l decay heat levels defined whenever appropriate.

2.

Power level - Power level may be important since certain equipment conditions o r configurations may only e x i s t a t certain power levels, e.g.,

hot standby.

As i n the case of decay heat level identification, the reasons f o r the selection o f specific power levels f o r analysis purposes should be stated.

It should be noted t h a t under certain conditions a reactor system may be a t a high power level with a low decay heat condition.

3.

Location o f event - I n many instances the location o f the event i s defined.

For example, an event consisting o f a f a i l e d open turbine bypass valve has the location defined since it i s a specific valve failure.

However, f o r some events such as pipe breaks, the location i s not defined and could have an impact on the progression o f the event.

I n the case i n which location i s not defined, a l l locations that could be s i g n i f i c a n t 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 o f event - Many o f the i n i t i a t o r s can occur t o various degrees.

For example, a LOCA can range from a very small break t o a f u l l g u i l -

l o t i n e pipe break.

Break sizes should be examined t o i d e n t i f y categories o f sizes t h a t lead t o similar system conditions.

I n the case o f the LOCA event, special consideration should be given t o the i d e n t i f i c a t i o n o f break sizes t h a t could lead t o loop flow stagnation.

The larger-sized LOCAs t y p i c a l l y do not contribute t o PTS r i s k since the pressure cannot be maintained because o f the large flow out o f the break.

3.2.3 D e f i n i t i o n o f Potential Transients Resulting from Each I n i t i a t o r After the complete set o f s i g n i f i c a n t i n i t i a t o r s has been defined, event trees are required t o i d e n t i f y potential sequences resulting from each i n i t i a -

tor.

The development o f the event tree headings and branches should be done i n a consistent and logical manner.

This was done i n the ORNL studies (Refs. 1,

2, and 3) by using what have been called system state trees.

These trees define the potential states o f each plant system o f interest conditional on specific thermal-hydraulic conditions.

I n i t i a t o r - s p e c i f i c event trees can then be developed by examining the system state trees with respect t o each i n i t i a t i n g

event.

A similar or equivalent approach should be used to ensure traceability of the event trees and to ensure that important sequences are not inadvertently el imi nated.

Support system failures should also be presented within some type of 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 that i s already mapped out on the event trees, b u t in many instances with a higher pro- bability of occurrence.

In other cases, i t may be necessary t o define event trees resulting from a support system failure.

In either case, i t i s important that the support systems be examined to identify their potential impact on over- cooling 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 a t least include:

"

The electrical supply system.

O

The compressed a i r instrument system.

O

The component and service water systems.

Operator Effects The operator effects are analyzed in two separate sections.

In this sec- tion the potential operator actions are identified.

These actions are further analyzed in Section 3.4 in which the probabilities associated w i t h the perfor- mance o f an operator action are developed.

I

The operator can improve, aggravate, or initiate an overcooling transient.

1 All three of these categories should be discussed in this section.

.

1.

Procedures and/or the operators1 general knowledge can lead t o actions that improve the conditions associated with an overcooling event.

Explanation should be included as to why i t i s 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 that can later be coupled w i t h the principal event trees.

2.

Although the ORNL studies (Refs. 1, 2, and 3) did not include operator- initiated events or events aggravated by operator actions contrary to procedures, this category of events should also be examined as part of a plant-specific analysis.

3.

The analyses should include a quantitative approximation of the PTS

risk resulting from operator acts of commission.

Also included should be the possibility that an operator could initiate or exacerbate some milder event into a more severe PTS-type event.

Since there i s no generally accepted way t o perform such analyses, the approximation used by the licensee for this purpose should be discussed and justified for appl icabi 1 ity to this particular plant.

The "confusion matrix" approach (Ref. 6) used i n human reliability analysis could provide an acceptable 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 ttie success or failure of specific equipment operations, and success and failure probabilities asso- ciated 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. Reference 7 includes guidance about treatment of 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 the reliability data used as input to quantify the event sequences. An explanation should be suppl ied as to how the data were derived for each data point.

3.4.1 Initiating Events Initiating event frequencies should be developed based on the number of observed events within selected periods of operation for similar plants under consideration. If no failures have been observed and no othe-r 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 Appendix B to Reference 3 for estimating plant-specific initiating event frequencies can be used. For some initiators, it may be neces- sary to 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 that might favor failure iti 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.

3.4.2 Equipment Fai 1 ures Following each initiating event, certain components are designed to perform in a defined manner. Failure of a component to perform its required function 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 o f demands expected within that same period, or by developing fault trees. If no failures have been observed and no other information is available with which to estimate a fai 1 ure-on-demand probabi 1 i ty , a standard statistical method can be used to develop a probability.

As with all event trees, the probability associated with a particular branch is conditional on the prior branches in the sequence. Questions of conditional probabi 1 i ty should be careful ly considered before a fai 1 ure probabi 1 i ty i s assigned.

The potential for coupled or common cause failures within a system or between systems should be examined in the analysis. Careful consideration

should be given t o increasing the f a i l u r e potential o f a component, given the f a i l u r e o f one o r more components o f the same type i n the same system o r i n other systems being subjected t o the same environment o r f a u l t causes.

As additional components o f a p a r t i c u l a r type are postulated t o f a i l , the proba- b i l i t y f o r the next component o f the same type t o f a i l should increase.

Based on the ORNL analysis, a simplified approach would be t o assume t h a t the f a i l u r e p r o b a b i l i t y o f the second component, given t h a t the f i r s t component has f a i l e d ,

might be as high as 0.1.

The t h i r d component might be assumed t o f a i l w i t h a

0.3 probability, given the f a i l u r e o f two identical components.

One could then assume that, a f t e r the f a i l u r e o f three components o f the same type, a l l remaining components o f t h a t type i n the same o r i n other systems being subjected t o the same environment o r f a u l t causes would f a i l w i t h a p r o b a b i l i t y o f 1.0.

The licensee should discuss how these types o f coupled f a i l u r e s are handled i n the analysis.

Common cause f a i l u r e s o f a d i f f e r e n t type may occur, as previously dis- cussed, through the f a i l u r e o f a support system or a control signal.

An anal- y s i s o f these potential f a i l u r e s should be made and the branch p r o b a b i l i t i e s should be adjusted whenever appropriate.

3.4.3 Operator Actions Operator action p r o b a b i l i t i e s are p a r t i c u l a r l y d i f f i c u l t t o determine because o f the lack o f a data base.

The problem i s f u r t h e r complicated when time becomes an important variable.

The procedure outlined below represents one approach t o quantifying operator actions.

This procedure shoul d be conser- vative f o r any operator action ~erformed as required by procedures assuming i,

t h a t the equipment required i s operational.

For operator actions t h a t might not be associated w i t h procedural steps, it i s not clear t h a t t h i s s i m p l i f i e d approach would produce conservative frequencies.

Therefore, the approach described would only be recommended f o r operator actions associated w i t h proce- dural steps.

Regardless o f the method used, the human e r r o r p r o b a b i l i t i e s used i n these analyses should be supported by data validated f o r the p l a n t being analyzed.

1. I d e n t i f y operator actions - I n t h i s step the procedures associated w i t h each i n i t i a t o r would be reviewed t o i d e n t i f y those operator actions t h a t

7 would have an impact on downcomer temperature.

2.

I d e n t i f y time constraint - I n the case o f each operator action, the transient would be reviewed assuming no operator action t o i d e n t i f y the time- frame available f o r successful comp~etion' o f the operator action.

3.

Assign screening f a i l u r e p r o b a b i l i t i e s - I n t h i s step a conservative value f o r the f a i l u r e o f the operator action would be identified.

For operator actions required by procedures- t o be performed w i t h i n the f i r s t 5 minutes o f the transient, the t i m e - r e l i a b i l i t y curve as presented i n NUREGKR-2815 (Ref. 7)

could be used t o i d e n t i f y a screening value.

After 5 minutes, a value o f 0.9 f o r success and 0.1 f o r f a i l u r e would be assumed f o r a l l operator actions.

The e n t i r e PTS analysis would then be completed using these screening values.

4.

I d e n t i f y dependency factors - I n some instances, there may be coupled f a i l u r e s associated w i t h operator actions j u s t as there were coupled f a i l u r e s

associated with equipment failures.

I n many instances, the potential failure of an operator action may be linked, t o various degrees, to the success or fail- ure of a previous operator action.

Thus, i t is recommended that each operator action be reviewed w i t h respect to dependency.

T h i s can be accomplished using the dependency tables as presented i n the human reliability handbook (Ref. 8).

5. If any of the dominant sequences involve the failure of an operator action, a more comprehensive evaluation of the failure would be performed for that operator action. When necessary, the comprehensive evaluation should be performed using a human reliability methodology.

The acceptability of this methodology for the purpose should be justified by the licensee (Refs. 9 through 13).

3.5 Event Tree Col lapse Collapse the event trees using a frequency screening criterion to form a l i s t of specific sequences and a set of residual groups to be analyzed.

T h i s i s important since the event trees may generate thousands of end states that cannot be individual ly analyzed.

A screening value of 1.0E-7/reactor year i s recommended.

This value should ensure that important sequences are treated individually, and i t should also help to keep the size of the residual small.

This i s particularly important since i t may be necessary t o t r e a t the residual using a bounding consequence condition.

3.5.1 Specific Sequences Those sequences that survive the frequency screening should be defined and their frequency noted.

I t i s recommended that some identification be assigned t o each sequence t o enhance i t s traceability through the remainder of the anal- ysis.

Grouping and identifying each sequence w i t h respect to initiator type may also prove helpful.

3.5.2 Residual Groups Those sequences that do not survive the frequency screening must also be considered.

They should be grouped together based on transient characteristics to form a s e t of residual groups.

The residual groups should be reviewed to identify sequences that should be grouped with previously defined sequences because of transient similarity or should be specifically evaluated because of their severe consequence.

I t i s important to attempt to reduce the size of each residual group since i t will be necessary to assign a bounding consequence that would apply within each group.

Each residual group should be defined and i t s frequency noted.

4. THERMAL-HYDRAULIC ANALYSIS

\\[

This chapter is to present the reactor coolant pressures, temperatures, and heat transfer coefficients at the vessel's interior surface in the beltline region for the set of overcooling sequences that envelops the plant's potential for experiencing a PTS event. A good example is presented in Chapter 4 of Reference 3. Also the chapter is to present the details of the analysis methods used to obtain these fluid conditions and is to include the following sections:

1.

The thermal-hydraul ic analysis plan and 1 ogic.

2.

A description and evaluation of the thermal-hydraul ic models.

3.

A description of any simplified analysis methods used in the study.

4.

A description of the methods used to evaluate the effects of thermal stratification and mixing.

5 .

Graphs of a1 1 the best-estimate thermal -hydraul ic results with their associated uncertainties and a detailed explanation of the transient behavior observed.

Thermal-Hydraulic Analysis Plan This section should out1 ine the logic and identify the subtasks in the thermal-hydraul ic analysis. Subtasks incl ude detai led thermal- hydraul ic systems

1 analysis, simp1 ified thermal-hydraul ic systems analysis, and thermal stratifica- tion analysis. The logic should describe the sampling plan used to select sequences for detailed or simplified analysis. ORNL experience favors selecting detailed thermal-hydraulic analysis sequences, including at least a few severe examples of each type of postulated overcooling 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 reduc- tion in expenditures. By first analyzing the scenarios that are expected to be the bounding cases (i. e. , the most severe), calculations for an entire class of overcooling scenarios may be deemed 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 extrapolation or interpolation of the results to other scenarios that share common controlling thermal-hydraulic phenomena.

During the analysis, the sequence identification analyst and the thermal -

hydraulic analyst should coordinate activities to ensure that pertinent details of the delineated sequences are thoroughly understood. Similarly, close coor- dination must be maintained between the thermal-hydraulic analyst and the frac- ture mechanics analyst so that the transient fluid conditions are calculated at the appropriate vessel locations.

4.2 Thermal-Hydraulic Models Thjs section and supporting appendices should present a detailed descrip- tion of the thermal-hydraulic computer models used in this analysi

s. The models

should include an accurate representation of the pertinent parts of the primary and secondary systems. This includes the condensate system, the main and auxil- iary 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 14 through 17 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 LOCAs, 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 crite- ria), a method of estimating 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. The thermal-hydraulic models should be coupled, where appropriate, with neutronic models that have the capability to analyze pressure surges resulting from any relevant sequences involving recriticality.

This section of the report must also present the results of benchmarking the computer models against suitable plant data or data from experimental facilities or incorporate this information by reference to an NRC-approved topical report. As a minimum, the plant data comparison should fully exercise the modeling features that are employed in the thermal-hydraulic computer pro- grams such as the pressurizer (including heaters and sprays), feedwater heaters and liquid level controls, the steam generator liquid level controls, and the turbine bypass (i . e. , steam dump) controls under steady-state and transient con- ditions. 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 the NRC at the time the calcula- tion is performed.

The models should be capable of accurately predicting condensation at all steam-water interfaces in the primary system, especially in the pressurizer during the repressurization phase of an overcool ing event or during refi 11 ing of the primary system with cold safety-injection water. The effects of noncon- densible gases, if present, on system pressure and temperature calculations should be addressed.

All code input and modeling assumptions should be documented and available for NRC review during the analysis review period (normally starting 3 years before the plant exceeds the screening limit and continuing until the evaluation results and any requisite actions are approved by the Commission).

Simplified Analysis Methods This section should present the technical bases for any simplified analysis methods that are applied in the study. This includes the grouping of similar sequences by controlling phenomena and any extrapolations used to modify exist- ing calculations. If a simplified thermal-hydraulic plant model is used to pre- dict portions of the plant transients, all the simplifying assumptions inherent

t o t h i s model should be stated and j u s t i f i e d .

Reference 18 provides examples of how t o group sequences and develop a simplified thermal-hydraulic model suitable i

f o r portions o f the analysis.

i

4.4 Thermal S t r a t i f i c a t i o n Effects Transient thermal-hydraulic computer programs available t o analyze LWR

response t o overcooling scenarios do not model f l u i d behavior with s u f f i c i e n t d e t a i l t o predict the onset o f H P I thermal f l u i d s t r a t i f i c a t i o n i n the cold leg and the subsequent cold l e g and downcomer behavior.

As a result, additional analysis methods may be needed t o determine which transients are affected by thermal s t r a t i f i c a t i o n and the extent o f such effects.

This section should describe and j u s t i f y the thermal f l u i d mixing analysis methods t h a t have been applied i n the study.

References 19 through 24 describe the results o f recent mixing analyses and experiments.

Reference 19 i d e n t i f i e s a useful s t r a t i f i c a t i o n c r i t e r i o n t o determine which overcooling transients w i l l require the additional mixing analysis.

Particular attention should be given t o scenarios t h a t involve H P I under very low flow o r stagnant loop conditions.

When stagnation i s p a r t i a l (i.e.,

not a l l loops stagnate), s t r a t i f i c a t i o n i s expected only w i t h i n the cold legs o f the stagnant loops.

However, scenarios involving complete loop stagnation w i l l require the evaluation o f a transient cooldown i n the presence o f s t r a t i f i e d layers both i n the cold legs and i n a portion o f the downcomer.

The mixing model should include the e f f e c t o f metal heating on the mixing behavior, p a r t i c u l a r l y i n a stagnant flow situation.

Also, the e f f e c t o f noncondensible gases, i f present, should be included.

References 19 through 23 describe tools t h a t have been used f o r such an analysis.

I _

This section should also document the heat transfer correlations applied i n the mixing analysis.

The research e f f o r t s described i n References 18 through

23 indicated t h a t the downcomer heat transfer coefficients generally exceeded

300 Btu/hr-ft2-OF.

These values o f heat transfer c o e f f i c i e n t were generally high enough t o keep the vessel wall surface temperatures w i t h i n a few degrees o f the downcomer f l u i d temperature.

Furthermore, because the vessel wall cool- down was controlled by conduction processes rather than convection processes, the vessel wall surface temperatures were insensitive t o heat transfer coef- f i c i e n t variations due t o changes i n flow and heat transfer regimes.

4.5 Thermal-Hydraulic Analysis Results This section should present graphs o f the best-estimate downcomer pressures, f 1 u i d temperatures, and heat transfer coefficients and t h e i r associated uncer- t a i n t y ranges as a function o f time a t the c r i t i c a l weld areas.

This includes the results o f the detailed thermal-hydraulic model, the simplified model, and mixing analysis calculations.

The duration assumed f o r each overcooling scenario should be j u s t i f i e d .

It i s assumed t h a t a scenario duration o f 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> may be reasonable f o r many cases since the overcool ing transient would probably be i d e n t i f i e d and m i ti gated p r i o r t o t h a t time.

However, there may be scenarios requiring lengthier evaluation periods because the control 1 i ng phenomena delay the scenario's evolution.

I

Also provide a discussion o f the accuracy o f the results, including a demonstration t h a t nodalization and error estimation methods chosen are appro- p r i a t e , and how the predicted p l a n t behavior compared t o p l a n t h i s t o r y and oper- ating experience.

Time-dependent uncertainty estimates f o r the downcomer pres- sure, f l u i d temperature, and heat transfer coefficients a t the c r i t i c a l welds should be provided f o r each scenario.

These uncertainties are often l i m i t e d by physical phenomena.

For example, the pressurizer power-operated r e l i e f valve (PORV) setpoints w i l l l i m i t the system pressure f o r certain high-pressure sce- narios.

Therefore, the uncertainty i s l i m i t e d by PORV operating character- i s t i c s .

References 16 and 18 describe some uncertainty analysis techniques.

5. FRACTURE MECHANICS ANALYSIS

For each sequence identified in Chapter 3, "Determination of Detailed PTS

Sequences for Analyses," calculate (or for unimportant sequences, estimate using bounding conditions) the conditional probabi 1 i ty of through-wal 1 crack penetra- tion given the occurrence of the event versus fluence or RTNDT (Although licensees were required to use the method of determining RTNDT (RTpTS) specified in paragraph 50.61(b)(2)

when evaluating their vessel properties with respect to the screening limits, in performing these plant-specific calculations, they are encouraged to use any alternative methods/data/correlations for which they provide justification of applicability to their specific plant.) Specific sequences identified in Section 3.5.1 should be calculated individually in detail. Less important events such as the residual groups identified in Sec- tion 3.5.2 may be conservatively bounded without a calculation for each sequence in the group. A good example is provided in Chapter 5 of Reference 3. Input for these calculations includes the primary system pressure, the temperature of 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. The calculations should be performed with a probabilistic fracture mechanics code such as OCA-P or VISA-I1 (Refs. 25 arid 26).

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 KIc and I

KIa to account for upper-shelf behavior. Plastic instability should be consid- ered in the determination of failure. Warm prestress should not be assumed in evaluations of the postulated transients. Acceptable types of material pro- perties 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 should be assigned distribution functions:

KIc = Static crack initiation fracture toughness KIa = Crack arrest fracture toughness RTNDT = Ni 1 -ducti 1 i ty reference temperature Cu = Concentration of copper, wt-%

Ni = Concentration of nickel, wt-%

F = Fast neutron fluence The functions used should be justified. Examples of these distributions are found in Reference 3.

The following additional information should be supplied:

/

1.

Flaw density - The number o f cracks per u n i t surface area should be established f o r use i n the calculations and should be j u s t i f i e d .

A value o f

0.2 flaw per square meter o f 8-inch-thick material (one flaw/cubic meter) was selected i n References 1, 2, and 3.

2.

Flaw depth density function - The flaw depth density d i s t r i b u t i o n should be established.

The function t o be used can be t h a t specified i n References 1, 2, and 3.

3.

Flaw size, shape, and location - Axial flaws w i t h depths less than

20 percent o f the wall thickness and a l l circumferential flaws should be modeled i n i n f i n i t e length.

Axial flaws with depths greater than 20 percent o f the wall thickness may be modeled i n i n f i n i t e o r f i n i t e length depending on the r e l a t i v e toughness o f the weld regions and plate material.

For instance, the length o f an axial flaw i n an axial weld t h a t suffers severe radiation damage r e l a t i v e t o the plate can be l i m i t e d t o the length o f the weld.

The flaws should be assumed t o be located a t the inner surface o f the vessel and should extend through the cladding t o the inner surface o f the vessel.

Reference 20 provides a comprehensive discussion o f recommendations f o r input distributions t o be used i n probabi 1 i s t i c fracture mechanics calculations.

4.

A l l regions o f the b e l t l i n e should be considered.

This includes axial and circumferential welds as well as the base material.

The f o l low

-

K ~ c

-

-

K ~ a

-

where T = Val 1 temperature R

T

~

~

~

o

= I n i t i a l n i l - d u c t i l i t y reference temperature Exampl

= Increase i n n i l - d u c t i l i t y reference temperature due t o radiation damage, f(Cu,Ni,fluence).

I f plant surveillance data meet the c r i t e r i a f o r c r e d i b i l i t y given i n Reference 27, they may be used as described therein.

es o f these functions are described i n References 3 and 27.

I n reporting the results, the methods used f o r the p r o b a b i l i s t i c vessel- i n t e g r i t y analysis should be described, t h e i r limitations f o r t h i s analysis identified, and the impact o f uncertainties i n the resulting vessel f a i l u r e probabilities estimated.

Discussion o f the analysis should include a l i s t i n g of the assumptions used, t h e i r bases, and a discussion o f the s e n s i t i v i t y o f the results t o variations i n the assumptions.

Vessel dimensions and material properties used should be given.

For each transient of interest, a deterministic analysis that includes a i

set of critical crack-depth curves as functions of time (see Refs. 1, 2, and 3),

i - e . , a plot of crack depths corresponding to initiation and arrest events versus L

time, should be carried out.

This plot should also have curves indicating the depth of crack a t which upper-shelf toughness i s effective.

These results should correspond to minus two sigma values for KIc and KIa, plus two sigma values for RTNDT, and plus two sigma values for the copper and nickel contents as well as plus two sigma for the fluence value.

These curves, which graphically represent the worst-case condition for each transient of interest, will be used i n the evaluation of the critical time interval from the initiation of the transient during which mitigating action can occur.

6.

INTEGRATION OF ANALYSES

-

I n t h i s chapter, the event frequencies are coupled with the results o f the fracture mechanics analysis t o obtain an integrated frequency o f vessel through- wall cracking due t o PTS.

An example o f one acceptable method i s presented i n Chapter 6 o f Reference 3.

A table t h a t supplies the following information f o r each specific sequence and residual group i d e n t i f i e d i n Section 3.5 should be provided.

These results are t o be provided f o r the operating time a t which the reactor w i l l reach the PTS screening c r i t e r i o n and f o r any additional operation

1 i f e bei ng requested:

O

Sequence identification.

O

Type o f i n i t i a t o r (smal 1-break LOCA with low decay heat, large steamline break a t f u l l power, etc.).

O

Estimated sequence frequency.

O

Method used t o determine conditional through-wal 1 crack penetration probabi 1 i ty.

O

Sequence conditional through-wal 1 crack penetration probabi l i ty."

O

Frequency o f through-wall cracking due t o sequence obtained by the product o f sequence frequency and sequence conditional through-wall crack penetration probabi 1 i ty.

For each dominant sequence, a section o r table should be provided t h a t sup- p l i e s (1) specific reference t o the graph o f temperature, pressure, and flow as provided i n Chapter 4, "Thermal-Hydraulic Analysisn; (2) a time-line description o f the accident sequence noting important operator actions, control actions, protection system actions, equipment faults, and vessel f a i l ure; and (3) fre- quency of through-wall crack penetration as a function o f fluence o r RTNDT-

Results should then be summed within each i n i t i a t o r type t o provide a fre- quency o f through-wall crack penetration as a function o f i n i t i a t o r type.

The discussion should explain why each i n i t i a t o r type i s or i s not impor- t a n t t o PTS.

Finally, the results should be summed over a l l i n i t i a t o r types t o provide an integrated frequency o f through-wall cracking f o r the vessel.

This inte- grated value should be reported as a function o f fluence, or RTNDTy and p l o t t e d with uncertainty values as determined i n Chapter 7, "Sensitivity and Uncertainty Analyses o f Through-Wall Crack Frequency," and included on the p l o t .

The dis- cussion should i d e n t i f y important operator actions, control actions, and p l a n t features t h a t can cause o r prevent vessel failure.

he conditional through-wall crack penetration probability i s the probability of a through-wall crack as determined by the fracture mechanics analysis, given t h a t the event occurs.

7.

SENSITIVITY AND UNCERTAINTY ANALYSES OF THROUGH-WALL CRACK FREQUENCY

I n order f o r the results o f the probabilistic analysis t o be useful f o r regulatory decisionmaking, the s e n s i t i v i t y o f the results t o input parameters and assumptions should be determined, the major sources o f uncertainty should be identified, and the magnitude o f the uncertainty should be estimated.

I n t h i s chapter, the results and the procedures used t o perform each o f these processes are t o be documented.

A good example i s given i n Chapter 7 o f Reference 3.

Portions o f t h a t analysis, or other analyses, may be referenced i n l i e u o f por- tions o f the analysis described i n t h i s chapter, provided the licensee demon- strates the appl icabi 1 i t y o f the referenced analyses t o the specific plant.

7.1 Sensitivity Analysis Perform a s e n s i t i v i t y analysis t o estimate the change i n the through-wall crack frequency f o r a known change o f a single parameter.

Parameters examined i n the s e n s i t i v i t y analysis should include (1) the i n i t i a t i n g event and event tree branch frequencies, (2) the thermal -hydraul i c variabl es (temperature, pres- sure, etc. ), and (3) the fracture mechanics variables (fluence, flaw density, etc.).

Where appropriate, 68th percentile (1-sigma) values should be used t o represent the change i n the parameter.

This should provide a s u f f i c i e n t change t o i l l u s t r a t e the effects o f the change, and the use o f the 68th percentile value whenever possible w i l l help t o define the important v a r i a b i l i t i e s .

I n the case o f temperature and pressure, however, the 68th percentile values may vary from one sequence t o another.

I n t h i s case, it may be easier t o i d e n t i f y a representative change i n the parameter t h a t could then be used f o r a l l sequences rather than t o t r y t o use the 68th percentile values.

1.

Each variable examined i n the s e n s i t i v i t y analysis should be l i s t e d along w i t h the change i n the variable.

I n the cases i n which changes are represented by using 68th percentile values, some explanation should be provided t o document the reasons the value i s considered a 68th percentile value.

I n those cases i n which something other than a 68th percentile value i s chosen, discussion should center around the reasons f o r choosing the value used.

Sensitivity factors should be obtained by dividing the through-wal 1 crack frequency obtained with the changed variable by the through-wall crack frequency obtained with each variable a t i t s mean value.

Supply the s e n s i t i v i t y factors obtained f o r both positive and negative changes i n each o f the variables.

The s e n s i t i v i t y factors obtained f o r changes made i n the PTS-adverse direction should be ranked according t o magnitude and provided i n table form.

Uncertainty Analysis

7.2.1 Parameter Uncertainties Each step i n the p r o b a b i l i s t i c analysis should include an uncertainty anal- ysis.

This should include uncertainty i n frequency o f occurrence o f a sequence, uncertainty i n temperatures and pressures reached during the sequence, including t h a t resulting from the nodal i z a t i o n scheme chosen as discussed i n Section 4.5, and uncertainty i n the fracture mechanics model for vessel f a i l u r e given the transients.

For the following reasons, a Monte Carlo simulation i s appropriate for

,I

portions of the PTS uncertainty analysis.

O

The temperature and pressure error di s t r i buti ons are not symmetric.

O

The fracture mechanics results are nonlinear with respect to variations i n input parameters, particularly the temperature and pressure time hi stories.

O

The results of the Monte Car10 analysis can indicate the shape of the output distribution.

The Monte Carlo approach would involve four steps as described below:

1.

Develop a statistical distribution for each variable used in the calculation - T h i s step will involve the representatioh of each variable as a distribution w i t h 5th and 95th percentiles as previously 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 crack frequency i s obtained based on the randomly selected variables.

This requires under- standing the form of the relationship between each input variable and through- wall crack frequencies.

For some variables such as initiating event and branch frequencies and flaw density, this is simple since the through-wall crack frequency i s directly proportional to the value of these parameters over the range of variable val ues considered.

Other vari abl es such as temperature and pressure may require the development of an appropriate relationship.

In such cases i n which the effect of a variable change may be dependent on the value of another variable, response-surface techniques may be used to estimate important interaction effects.

4.

Summarize the resulting estimates and approximate frequency distribu- tion - Steps 2 and 3 are repeated until a statistically valid number of t r i a l s have been- performed.

A distribution of through-wall crack frequencies i s then 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 Model i ng Uncertai nties (Bi ases)

During the process of performing the PTS analysis, the analyst will make simplifying assumptions i n order to make the analysis tractable.

Such assump- tions include decisions on thermal-hydraulic models, fracture mechanics models, grouping of sequences both for thermal-hydraulic analysis and fracture mechanics analysis, nodalization i n the thermal-hydraulic models, etc.

These assumptions can introduce conservative or nonconservative 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

1 isted.

Each assumption should be identified as being either conservative or nonconservative.

A discussion should be supplied for each assumption w i t h respect t o i t s impact on the overall value of through-wall crack frequency.

Whenever excess conservatism or nonconservatism i s suspected to be present i n an assumption, an alternative assumption should also be used in the full calcu- lation procedure and the impacts on the overall result compared.

8.

EFFECT OF CORRECTIVE ACTIONS ON VESSEL THROUGH-WALL CRACK FREQUENCY

This chapter i s t o summarize the licensee's program o f corrective measures.

Each corrective measure considered by the licensee should be presented and ex- plained.

I n each case, the reasons f o r considering the action as a corrective measure are t o be documented, and the estimated impact o f the action with respect t o through-wall crack frequency provided.

Corrective actions t h a t are t o be considered include, but are not l i m i t e d to, those discussed i n the remaining sections o f the chapter.

An example can be found i n Chapter 8 o f Reference 3.

8.1 F l ux Reduction Program Early analysis and implementation o f such flux reductions as are reasonably practicable t o avoid reaching the screening c r i t e r i o n are already being required and accomplished i n accordance with the PTS rule, § 50.61.

Further f l u x reduc- tions t o c r i t i c a l areas o f the vessel wall t h a t would reduce the r i s k o f con- tinued operation beyond the screening c r i t e r i o n should be considered.

I f such additional f l u x reductions are needed, i n view o f the i r r e v e r s i b i l i t y o f embr-ittlement, the .Ticensee should consider early implementation before reaching the screening criterion.

For licensees who are considering applications t o extend the operating license beyond i t s present expiration date, i t may be pru- dent t o implement the reduction as early as possible t o avoid the necessity o f vessel annealing or replacement.

8.2 Operating Procedures and Training Program Improvements Operator actions and associated p l a n t response play a key r o l e i n the i n i t i a t i o n and mitigation o f PTS events.

Therefore, ensure t h a t the actions are based on approved technical guidelines t h a t include an integrated'evaluation o f relevant technical considerations, including, but not l i m i t e d to, PTS, core cooling, environmental releases, and containment i n t e g r i t y .

The evaluation should address the following types o f concerns:

Frequent real i s t i c "team" t r a i n i n g should be conducted, exposing the operators t o potential PTS transients and t h e i r precursor events.

The t r a i n i n g should give the operators actual practice i n controlling reactor system pressure and cooldown rates during PTS situations.

Specific t r a i n i n g should include, but not be l i m i t e d to, reactor cool- ant pump t r i p c r i t e r i a , the HPI t h r o t t l i n g c r i t e r i o n , control o f natural circulation, recovery from inade,quate core cooling, recovery from s o l i d plant operations, and the use o f PORVs t o control primary overpressure.

O

Instructions should be based on analyses t h a t include consideration o f system response delay times (e. g. , loop transport time, thermal transport time).

O

Whether o r not there i s a need f o r cooldown rate l i m i t s f o r periods shorter than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> should be evaluated.

O

Methods f o r control 1 i n g cooldown rates should be provided.

Reference should be made t o these methods with respect t o the dominant PTS

r i s k sequences whenever possible.

O

Guidance should be provided for the operator if cool down rates or i

pressure-temperature limits are exceeded. These guidelines should L

take into account potential core cooling, environmental release, or containment integrity problems that could exist as a result of respond- ing to the abnormal cooldown rate. These guidelines should leave little doubt as to when PTS concerns are more important than other safety issues and when other safety issues assume primary importance over PTS concerns. It should be emphasized how the guidelines were evaluated and optimized in light of any competing risks that might arise from events other than PTS events to ensure that plant safety is appropriately balanced.

O

The desired region of operation between the pressure-temperature 1 imi t and the limit determined by avoidance of saturation conditions should be evaluated to determine if it can be revised to minimize total risk due to plant operation from PTS plus non-PTS events.

O

Instructions for control 1 ing pressure following depressurization transients should be provided.

O

Instructions should be available for the condition where natural

,circulation is lost and the primary system main circulation pumps are not available.

Portions of the above may be provided by incorporation by reference, for example, to the plant-specific Emergency Response Guidelines. However, a sum- mary discussion re1 ati ng the referenced material to the overall subject should be provided.

8.3 Inservice Inspection and Nondestructive Examination Program The use of state-of-the-art nondestructive examination (NDE) techniques could provide an opportunity to decrease any conservatism that might exist in the flaw density value used in the analysis. This decrease in conservatism, however, may be less important than the decrease in uncertainty in the actual flaw density that may result from an examination of this type.

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. Considerati on should be given to increased frequency of inspections.

The reliability of the NDE method selected to detect small flaws should be documented.

8.4 Plant Modifications All plant modifications should be evaluated and optimized in light of any com- peting risks that might arise from events other than PTS events to ensure that overall plant safety is appropriately balanced. PI ant modifications that may be considered include the following:

1.

Instrumentation, Controls, and Operation a.

Reactor vessel downcomer water temperature monitor.

b.

Instantaneous and integrated reactor coolant system cooldown rate monitors.

c.

Steam dump interlock.

d.

Feedwater i sol ation/f low control 1 ogic.

e.

Reactor coolant system .pressure and temperature monitors.

f.

Control system to prevent repressurization of the reactor primary coolant system during overcooling events.

g.

Monitor to measure margin between vessel inner-surface temperature and current RTNDT at that location.

h.

Diagnostic instrumentation and displays.

i.

Primary coolant system pump trip logic..

j.

Automatic isolation of auxiliary feedwater to broken steam

1 i nedgenerators.

2.

Increased Temperature of Emergency Core Cooling Water and Emergency Feedwater If plant modifications are proposed to prevent overcooling, the report should include an evaluation of undesirable side effects (i.e., undercooling)

and a discussion of steps planned to ensure that the modifications represent a net improvement in safety when PTS and non-PTS related events are considered.

8.5 In Si tu Anneal i ng If in situ annealing is part of the licensee's program of corrective mea- sures, 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, the program directed toward code requalification after annealing, and the program to ensure that annealing does not introduce other safety problems.

9.

FURTHER ANALYSES

The PTS rule (Q

50.61 of 10 CFR Part 50) requires Commission approval for plant operation with RTpTS values above 270°F.

This regulatory guide out1 ines the analyses that should be performed in support of any request to operate a t R

T

~

~

~

values in excess of 270°F, as required in paragraphs 50,6l(b)(4) and

50.61(b)(5), and states that the s t a f f ' s primary acceptance criterion wi 11 be licensee demonstration that expected through-wall crack frequency will be below

5 x per reactor year for such operation.

In the event that a licensee i s unable to meet this primary acceptance criterion, he may request Commission approval for continued operation under the provisions of paragraph 50.61(b)(6), which allows the submittal of further anal- yses.

The content of these further analyses would be determined by the licensee and might include topics such as overall plant risk analyses that are beyond the scope of the vessel failure analyses covered by this regulatory guide.

10.

RESULTS AND CONCLUSIONS REGARDING PTS ANALYSES

1 This chapter i s to summarize the models used and the results obtained and provide the conclusions reached with respect to continued operation of the plant.

10.1 Summary of Analysis In this section the major findings of each aspect of the PTS analysis, as described i n the previous chapters, should be presented.

These should include:

O

Expected (mean) value of frequency of reactor vessel through-wall crack penetration versus time, w i t h uncertainty bound (95th percentile).

O

Identi f ication of dominant accident sequences.

O

If sensitivi ty/uncertainty analysis shows that slightly different assumptions could lead to different dominant sequences, identification of these assumptions and discussion of the impact on results given the different assumptions.

O

Identification of important operator actions, control actions, and plant features that can increase or decrease the frequency or severity of overcooling transients, and whether these have been appropriately balanced to ensure optimum overall plant safety.

O

Major sources and magnitudes of uncertainty i n the analysis.

O

The re1 ative effectiveness of potential a1 ternati ve corrective measures in reducing the expected (mean) value of through-wall crack penetration.

O

The program of planned corrective measures.

10.2 Basis for Continued Operation Finally, as part o f the plant-specific analysis package, the licensee should provide a basis for concluding whether or not continued plant operation i s justified.

The basis for continued operation should include comparison with NRC's PTS acceptance criteria given in the Introduction to this guide.

REFERENCES

I -i

1.

T. 3. Burns e t a1 . , "Preliminary Development o f an Integrated Approach t o the Evaluation o f Pressurized Thermal Shock Risk As Applied t o the Oconee Unit 1 Nuclear Power Plant," Oak Ridge National Laboratory, U.S.

Nuclear Regulatory Commission (USNRC) Report NUREG/CR-3770 (ORNL/TM-9176),

May 1986.

2.

D. L. Selby e t al., "Pressurized Thermal Shock Evaluation o f the Calvert C l i f f s Unit 1 Nuclear Power Plant," Oak Ridge National Laboratory, USNRC

Report NUREG/CR-4022 (ORNL/TM-9408),

November 1985.

3.

D. L. Selby e t al., "Pressurized Thermal Shock Evaluation o f the H. B.

Robi nson Unit 2 Nuclear Power Plant ,'I Oak Ridge National Laboratory, USNRC Report NUREG/CR-4183 (ORNL/TM-9567),

November 1985.

4.

USNRC, "Pressurized Thermal Shock (PTS) ,'I SECY-82-465, November 23, 1982.

5.

Appendix F t o "Integrated Plant Safety Assessment Report, Systematic Evaluation Program, San Onofre Nuclear Generating Station Unit 1," USNRC

Report NUREG-0829, Apri 1 1985.

6.

L. Potash, "ConfusionMatrix,"

SectionC.1.2of Appendix C inl'OconeePRA,"

Electric Power Research I n s t i t u t e , Palo Alto, CA, and Duke Power Co.,

Char1 otte, NC, NSAC/60, Vol . 4, 1984.

7.

R.

A.

Bari e t al. , "Probability Safety Analysis Procedures Guide,"

k Brookhaven National Laboratory, Revision 1 t o USNRC Report NUREGKR-2815, August 1985.

8.

A. D. Swain and H. E. Guttmann, "Handbook o f Human R e l i a b i l i t y Analysis with Emphasis on Nuclear Power Plant Applications ,I1 Sandia National Laboratories, USNRC Report NUREGKR-1278 (SAND80-0200),

October 1983.

9.

M.

K. Comer e t al., "Generating Human R e l i a b i l i t y Estimates Using Expert Judgment ,It General Physics Corporati on, USNRC Report NUREGAR-3688, Vol s. 1 and 2, January 1985.

10.

D.

E. Embrey, "The Use o f Performance Shaping Factors and Quantified Expert Judgment i n the Evaluation o f Human R e l i a b i l i t y :

An I n i t i a l Appraisal , " Broo khaven National Laboratory, USNRC Report NUREGKR-2986 (BNL-NUREG-51591),

October 1983.

11.

0. A.

Seaver and W.

G. S t i l l w e l l , "Procedures f o r Using Expert Judgment To Estimate Human Error Probabi 1 i t i e s i n Nuclear Power Plant Operations ,"

Sandia Nattonal Laboratories, USNRC Report NUREGAR-2743 (SAND82-7054),

April 1983.

12.

Organi sation f o r Economic Co-operation and Development, Nuclear Energy Agency, Committee on the Safety o f Nuclear I n s t a l lations, "Assessing Human Re1 i a b i l i t y i n Nuclear Power Plants ," May 1983.

13.

Organisation f o r Economic Co-operation and Development, Nuclear Energy Agency, Committee on the Safety o f Nuclear I n s t a l 1 ations , "Expert Judgment o f Human Reliability," CSNI Report No. 88, January 1985.

0. Bassett e t a1 . , "TRAC Analyses o f Severe Overcool i n g Transients f o r the Oconee 1 PWR,"

Los Alamos S c i e n t i f i c Laboratory (LASL),

USNRC Report NUREG/CR-3706, August 1985.

C. D. Fletcher e t al., "RELAP 5 Thermal-Hydraulic Analysis o f PTS Sequences f o r the Oconee 1 PWR,"

EG&G, USNRC Report NUREG/CR-3761, July 1984.

3. Koenig, G. Spriggs, and R. Smith, "TRAC-PF1 Analyses o f Potential PTS

Transients a t a Combustion Engineering PWR,"

LASL, USNRC Report NUREGKR-4109, Apri 1 1985.

C. D. Fletcher e t al., "RELAP 5 Thermal-Hydraulic Analyses o f PTS Sequences f o r H. B. Robinson Unit 2 PWR,"

EG&G, USNRC Report NUREG/CR-3977, A p r i l 1985.

C. D. Fletcher, C. B. Davis, and D. M. Ogden, "Th6rmal-Hydraulic Analyses o f Overcooling Sequences f o r the H. B. Robinson Unit 2 PTS Study,"

EG&G,

USNRC Report NUREGAR-3935, July 1985.

T. G. Theofanous e t al., "Decay o f Buoyancy Driven S t r a t i f i e d Layers w i t h Appl ication t o PTS ,I1 Purdue University , USNRC Report NUREG/CR-3700,

May 1984.

T. G. Theofanous e t al. , "REMIX:

Computer Program f o r Temperature Transients Due t o High Pressure I n j e c t i o n i n a Stagnant Loop," Purdue University , USNRC Report NUREGKR-3701, May 1986.

T. G. Theofanous e t al., "Buoyancy Effects on Overcooling Transients Calculated f o r the USNRC Pressurized Thermal Shock Study," Purdue University , USNRC Report NUREG/CR-3702, May 1986.

Bart Daly, "Three-Dimensional Calculations o f Transient Fluid-Thermal Mixing i n the Downcomer o f the Calvert C l i f f s - 1 Plant Using SOLA-PTS,"

LASL, USNRC Report NUREG/CR-3704, Apri 1 1984.

Martin Torrey and Bart Daly, "SOLA-PTS:

A Transient 3-D Algorithm f o r F l uid-Thermal Mixing and Wall Heat Transfer i n Complex Geometries," LASL,

USNRC Report NUREG/CR-3822, July 1984.

F. X.

Do1 an e t a1 . , "Faci 1 i ty and Test Design Report:

1/2-Scal e Thermal Mixing Project ," USNRC Report NUREGAR-3426, Vols. 1 and 2, September 1985.

R. D. Cheverton and D. G. Ball, "OCA-P, A Deterministic and Probabilistic Fracture-Mechanics Code f o r Application t o Pressure Vessels ,I1 Oak Ridge National Laboratory, USNRC Report NUREG/CR-3618 (ORNL-5991),

July 1984.

F. A. Simonen e t al., "VISA-I1 - A Computer Code f o r Predicting the Proba- b i 1 i t y o f Reactor Vessel Fai 1 ure ,I1 Battel l e Pacific Northwest Laboratories, USNRC Report NUREG/CR-4486, Apri 1 1986.

USNRC Regulatory Guide 1.99, "Effects o f Residual Elements on Predicted Radiation Damage t o Reactor Vessel Materials. "

REGULATORY ANALYSIS

The pressurized thermal shock (PTS) rule, § 50.61 of 10 CFR Part 50

(July 23, 1985--50 FR 29937), requires collection and reporting of material properties data, analyses of flux reduction options, and detailed plant-specific PTS risk analyses for those plants that reach the screening criterion based on RTNDT,* as specified in the rule, during the term of the operating 1 icense.

The regulatory guide addresses the detailed plant-specific risk analysis requirement, providing recommendations regarding how licensees should perform and how the NRC

staff should review those analyses.

Neither the PTS rule nor the regulatory guide requires specific corrective actions.

The guide merely provides guidance for the performance of the analyses required by the rule to identify and select necessary corrective actions.

There- fore, i n accordance w i t h the Commission's Regulatory Analysis Guide1 ines (NUREG/

BR-0058, Revision l), this regulatory analysis does not provide extensive and detai led assessment of required, specific corrective actions.

The background material, nature of the problem, objectives, and costs, etc.,

of the PTS rule's requirements are covered in the regulatory analysis prepared as part of the rulemaking proceeding (Enclosure B to SECY-83-288, Proposed Pressurized Thermal Shock (PTS) Rule, July 15, 1983, and Enclosure D to SECY-85-60, Fi nal Pressurized Thermal Shock (PTS) Rul e, February 20, 1985).

This regulatory analysis therefore addresses only (1) the need for publishing guidance regarding how licensees should perform the required plant-specific analyses, (2) the appropriateness of this particular guidance, and (3) the basis i

for the NRC staff acceptance criteria provided in the subject guide.

,

Need for Guidance The NRC staff has gained 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 laboratories and sponsored by NRC, and from review of industry-sponsored analyses.

The regula- tory guide reflects the lessons learned from this experience and will aid

1 icensees in performing analyses that wi 11 efficiently derive risk estimates in the form the NRC needs for use in evaluating their conformance with the regulations.

This need for guidance i s particularly acute since the plant-specific PTS

analyses should use a probabi 1 i s t i c risk analysis (PRA) approach, as opposed to the more traditional design basis accident (DBA) approach, as explained be1 ow.

The PTS risk i s developed as the sum of the small risks resulting from each of a large number of possible (but unlikely) PTS events.

The regulatory guide accordingly describes acceptable methods to identify as many as possible of the potential PTS events, group them, calculate the frequencies and conse- quences of each group, determine the risk due to each group by multiplying the predicted frequency by the calculated consequences, and then sum the results

"Reference Temperature for the Nil Ductility Transition, a measure of the temperature range i n which the materials' ductility changes most rapidly with changes i n temperature.

from a1 1 groups t o obtain t o t a l PTS r i s k estimates t h a t can be compared w i t h the acceptance c r i t e r i a given i n the regulatory guide.

The DBA approach, on the other hand, would attempt t o define a worst cred- i b l e event (the "design basis accident") and then show t h a t (1) consequences from t h a t event are acceptable and (2) a l l other credible events are less severe and therefore acceptable.

The s t a f f has determined t h a t t h i s DBA approach i s not appropriate f o r plant-specific PTS analyses because the t o t a l r i s k from a l l credible PTS events can be s i g n i f i c a n t even though each event i n d i v i d u a l l y i s less severe than the DBA.

The NRC s t a f f therefore believes t h a t t h i s guide w i l l encourage licensees t o use the acceptable PRA approach and not waste time and resources on the more t r a d i t i o n a l DBA approach.

2.

J u s t i f i c a t i o n o f This.Particular Guidance The NRC staff has performed prototype plant-specific analyses f o r three plants.

They constitute the most detai 1 ed, thorough ana.lyses performed t o date, and the lessons learned i n t h e i r performance are r e f l e c t e d i n the guide.

The NRC s t a f f has incorporated i n t o the guide descriptions o f the best methods found regarding how t o assemble d e t a i l s o f a p l a n t ' s design (and t o what level those d e t a i l s should be included), how t o use event t r e e methodologies t o i d e n t i f y and group potential PTS events, how t o calculate severity o f the events, how t o integrate the r e s u l t i n g r i s k , and many other subjects.

The s t a f f believes t h a t the benefit o f t h i s experience i s presented i n t h i s guide, and i t s use by li- censees w i l l enable them t o avoid many o f the false s t a r t s and errors made by the s t a f f and t h e i r contractors i n performing the prototype analyses, thereby saving time! and resources.

3.

J u s t i f i c a t i o n o f Acceptance C r i t e r i a The guide states that, i n judgi ng the acceptabi 1 i t y o f conti nued operation beyond the PTS screening c r i t e r i o n , the s t a f f w i l l accept any analyses performed w i t h acceptable methods such as those described i n the subject regulatory guide t h a t predict a through-wall crack penetration frequency less than 5 x per reactor year.

The mean frequency o f reactor vessel through-wall crack penetration i s used as the principal acceptance c r i t e r i o n because the s t a f f ' s analyses p r e d i c t t h a t there i s a high l i k e l i h o o d o f core damage i n the event o f such cracks.

Core damage events have potential public health and safety consequences t h a t are d i f f i c u l t t o analyze w i t h certainty.

They would also have severe economic impacts upon the licensee and the public who w i l l pay for cleanup and replace- ment power.

For a l l these reasons, reactor vessel through-wall crack penetra- t i o n frequency i s used as the p r i n c i p a l acceptance c r i t e r i o n .

The p a r t i c u l a r value o f 5 x mean frequency per reactor year was selected as an achievable, r e a l i s t i c goal t h a t w i l l r e s u l t i n an acceptable level o f r i s k . It i s believed t h a t t h i s value i s acceptably low considering t h a t pressure vessel f a i l u r e i s not p a r t o f the design basis o f the p l a n t and therefore must have a frequency low enough t o be considered incredible.

When the various (unquantifiable)

biases t h a t are inherent i n the analyses are taken i n t o account a t l e a s t q u a l i t a t i v e l y , such as the i m p l i c i t assumption t h a t "core damage" i s equivalent

to "core melt," this value probably results in a core me1 t mean frequency close I

I

to one per mil 1 ion reactor years.

k In the opinion of the NRC staff, there are no practical quantities on which t o base the acceptance criteria other than reactor vessel through-wall cracks (i.

e., vessel failure).

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555 OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

USNRC

PERMIT No. G-67

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