RELEASEDTOTHE PDRI 

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POLICY ISSUE 

December 20, 1999 (Information) SECY-99-289 

FOR: The Commissioners 

FROM: William D. Travers 

Executive Director for Operations 

SUBJECT: STATUS REPORT ON ACCIDENT SEQUENCE PRECURSOR PROGRAM AND 

RELATED INITIATIVES 

PURPOSE: 

To inform the Commission of the staff's activities and progress with respect to the Accident 

Sequence Precursor (ASP) Program and related initiatives. 

BACKGROUND: 

Established by the NRC in 1979 in response to the Risk Assessment Review Group report (Ref. 

1), the primary objective of the Accident Sequence Precursor (ASP) Program is to 

systematically evaluate U. S. nuclear plant operating experience to identify, document, and rank 

those operating events that were most significant in terms of the potential for inadequate core 

cooling and core damage (precursors). In addition, the ASP Program has the following 

secondary objectives: (1) to categorize the precursors for plant-specific and generic 

implications, (2) to provide a measure that can be used to trend nuclear plant core damage risk, 

and (3) to provide a partial check on PRA-predicted dominant core damage scenarios. The 

program is also used to monitor the agency's performance against the following Strategic Plan 

goal for maintaining safety: "No more than one event per year which is a significant precursor of 

a nuclear reactor accident." Since its inception, the ASP Program has published sixteen reports 

documenting the results of its review of operational experience for precursors covering the 

years 1969-1997. These reports have been issued on a yearly basis since 1986. 

Accident sequences of interest to the ASP Program are those that would have resulted in 

inadequate core cooling that could have caused severe core damage, if additional failures had 

occurred. Events or conditions considered to be potential precursors are analyzed, and a 

conditional core damage probability (CCDP) is calculated by mapping failures observed during 

the event onto accident sequences in risk models. 

Contact: 

Patrick O'Reilly, RES 

415-7570 

P_ •~ - I CI, - C9 D5 I
The Commissioners

The NRC staff uses the ASP methodology and models as follows: 

(1) To make prompt assessments of the risk significance of operational events to support 

regulatory decisions by senior management. 

(2) For Phase 3 of the significance determination process (SDP) to evaluate the significance 

of inspection findings as part of the agency's improved reactor oversight process. 

(3) To evaluate the change in risk associated with licensing amendments submitted by 

licensees requesting changes in surveillance frequencies or allowed outage times. 

(4) To determine the need for generic communications (such as information notices). 

(5) In the systematic screening, review, and analysis of operational experience data for 

accident sequence precursors. 

(6) To evaluate the generic implications of precursors, trend industry performance, and 

check against PRAs. 

(7) In regulatory analyses performed in association with the resolution of generic issues. 

(8) To evaluate the risk associated with a specific technical issue identified at an individual 

plant. 

Important points and key findings in the discussion below are identified in bold. The numbers in 

brackets associated with each point or finding explicitly link it to the appropriate use(s) listed 

above. 

DISCUSSION: 

Since the last status report, SECY-98-298, dated December 22, 1998, the staff has: 

* Per the Commission's direction given in response to COMSECY-98-031, transferred 

functional responsibility for the ASP Program to the Office of Nuclear Regulatory 

Research. 

0 Evaluated 1998 events for precursors and published the results of the analyses. 

0 Evaluated and assessed trends in the precursor data. 

* Evaluated the risk significance of issues regarding the D. C. Cook plant. 

* Evaluated 1999 events for precursors and published the results of the analyses. 

0 Redirected the coordination of the ASP Program.

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* Continued the development of the Standardized Plant Analysis Risk Models (SPAR). 

Transfer of ASP Program Functional Responsibilities to RES 

In January 1999, the Office for Analysis and Evaluation of Operational Data (AEOD) was 

abolished and many of its functions were transferred to the Office of Nuclear Regulatory 

Research (RES). Effective March 29, 1999, RES was reorganized, and, in concert with its 

charter to apply risk insights to the evaluation of operational experience data, the Operating 

Experience Risk Analysis Branch (OERAB) in the Division of Risk Analysis and Applications 

(DRAA) assumed responsibility for management of the ASP Program. In addition, OERAB 

became responsible for the development of the models and methods that support the ASP 

Program and other regulatory activities. 

Responsibility for development of the computer codes [e.g., the agency's suite of PRA codes 

(SAPHIRE) and the GEM analysis tool] that support the ASP Program remained with the 

Probabilistic Risk Analysis Branch (PRAB) in DRAA. 

1998 ASP Event Analysis 

The screening, review, and preliminary analysis of 1998 operational events are complete. The 

preliminary analyses identified 11 potential precursors affecting ten different units for 1998. 

Peer and licensee review of the 11 analyses is complete, and preparation of responses to 

comments and final analysis of the last four 1998 potential precursors are in progress. The 

review of licensee comments on one event led to a reanalysis of the event that yielded a 

conditional core damage probability (CCDP) which was below the ASP Program threshold 

(CCDP=1.0x10 6 ). As a result, ten operational events affecting nine different units are under 

consideration as precursors for 1998. Although this is an increase in the number of 

precursors over 1997 (when there were 5 precursors), the 1998 results are consistent 

with the period 1993-96, when the average number of precursors per year was 12. [6] 

Eight of the 1998 precursors involved the unavailability of equipment; two involved initiators. 

This result is consistent with 1993-1997 results (with the exception of 1996), in which 

conditional unavailability events outnumbered initiators. [6] 

Historically, the percentage of precursors occurring at BWRs has been less than the 

percentage of BWRs in the operating reactor population. One of the 1998 precursors 

occurred at a BWR; there were none in 1997. Results from the staff's review of IPEs (NUREG

1560) indicate that the BWR core damage frequencies estimated in the IPEs generally were 

less than those for PWRs. NUREG-1560 attributed the difference to the larger number of 

injection systems in the BWR design and the capability of the BWR to more easily depressurize 

to allow use of low pressure injection systems. Since the ASP Program uses conditional core 

damage probability (CCDP) as the screening criterion for precursors, this result is consistent 

with expectations relating to core damage frequency (CDF) from the body of BWR risk 

assessments. 

Precursors involving degradations or failures in the auxiliary/emergency feedwater (AFW/EFW) 

at PWRs typically have been contributors to the number of precursors in a given year. As in 

1997, none of the 1998 precursors involved problems with AFW/EFW systems. An update of 

the analysis first presented last year of the number of precursors involving failures or

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degradations of the AFW/EFW system over the period 1984-1997 revealed a statistically 

significant, decreasing trend. The 1998 data were consistent with this trend. [6] The 

observed decrease is attributed in part to the decrease in the number of unexpected AFW/EFW 

system demands and a slight reduction in the number of failures per demand. 

Three of the 1998 precursors involved problems with electrical equipment (two at Davis-Besse 

and one at Byron 1 - See Attachment 1). Although this result is consistent with the relative 

contribution of electrical equipment problems in the 1996 and 1997 results, it differs 

from the previous 5-year period (1990-95), when about 60 percent of the precursors each 

year involved electric power issues. [6] One 1998 precursor also involved a loss of offsite 

power (LOOP) - the tornado-caused LOOP at Davis-Besse in June. 

Four of the ten 1998 precursors involved loss-of-coolant-accident (LOCA)-related issues, but 

none involved an actual loss of reactor coolant. Three of these involved calibration and 

calculational errors in emergency sump and borated water storage tank level instrumentation 

which compromised the capability to transfer the emergency core cooling system from the 

injection mode to the recirculation mode of operation. These three precursors involved 

conditions at the three reactors at the Oconee site. 

Precursors with CCDPŽ_ 1 .Ox104 are considered important with respect to risk significance. 

There was one precursor identified for 1998 that had a conditional core damage probability 

(CCDP) Ž1.0x10.4 - the tornado-caused LOOP at Davis-Besse. Over the last eleven years, 

approximately 25% of the precursors have had a CCDP_> 1.0x10 4 . Since 1984, the 

occurrence rate for this group of precursors has exhibited a statistically significant 

decreasing trend. [6] 

The 1998 precursor results are presented in Attachment 1. Detailed discussion of the analysis 

results and the insights summarized above is found in the 1998 Precursor Report (NUREG/CR

4674, Volume 27, which is scheduled to be issued in January 2000). 

Evaluation of Risk Trends in Precursor Data 

The ASP results were reviewed to obtain insights about industry risk. This review consisted of 

analysis of trends in the occurrence of precursors, comparison of an Annual ASP Index with 

core damage frequencies (CDFs) from individual plant examinations (IPEs), and comparison of 

the modes and causes of precursors from the ASP data with those modeled in probabilistic risk 

assessments (PRAs) and IPEs. An updated chart showing CCDP "probability bins" for ASP 

results from 1984 through 1998 is provided as Attachment 2. Precursor results represent but 

one indication of the overall risk of the industry. 

A recent staff technical paper (Ref.2) reported the results of a statistical analysis of trends in 

the occurrence rate of precursors between 1984 and 1994. This paper updates that analysis 

with four more years of data (Attachment 3). Statistically significant decreasing trends 

were found for all of the ASP CCDP bins, except for precursors with CCDPM1.0x10 3 . 

Although, there is a decreasing trend for this bin, it is not statistically significant. 

Absent a precursor with CCDPI.0x10 3 in 1999, the trend identified will become

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statistically significant. [6] Precursors with CCDP_Ž 1 .Ox103 have occurred, on the average, 

about once every 2-3 years. The events in this group appear to exhibit no common (generic) 

characteristics with respect to the nature, modes, causes, and systems affected by the events. 

The last precursor with CCDP__ 1O.x10.3 was the loss-of-offsite power event with an emergency 

diesel-generator out of service for maintenance at Catawba 2 in February 1996. 

The following presents an update of the indication of industry-wide CDF risk implications based 

on operating experience from the ASP Program. Using CCDPs from ASP results to estimate 

CDF is difficult because: (1) the actual mathematical relationship requires a great level of detail, 

(2) statistics for frequency of occurrence of specific precursor events are sparse, and (3) events 

that did not occur also need to be accounted for in the assessment. The ASP models and 

process do not explicitly cover all core damage frequency scenarios and are therefore 

incomplete for estimating total core damage frequency. Also, using CCDP to estimate CDF can 

overestimate the frequency due to double counting. Due to these and other limitations, the 

CCDPs have been used primarily as a relative trending indication. 

The above limitations notwithstanding, a link between ASP results and CDF can be made using 

an Annual ASP Index based on the sum of the CCDPs divided by the number of reactor 

operating years. This index can be used for order of magnitude comparisons with industry 

average CDF estimates derived from PRAs and IPEs. For the last seven years, the index is: 

Year Annual ASP Index 

1998 5.8x10 6/RY 

1997 4.6x1 0 7/RY 

1996 2.4xl 05/RY 

1995 2.2xl 06 /RY 

1994 3.0xl 05 /RY 

1993 8.9xl 0 6/RY 

1992 1.5xl 05 /RY 

1.2xl 05/RY Average 

The estimated CDFs in the IPEs range from 1.2x10 6/RY to 3.7x104/RY, with an average value 

of 6.2x10-5/RY. They also provide incomplete estimates of total CDF, but IPEs are reasonably 

similar in scope to the current ASP Program. On an order of magnitude basis, the ASP 

Index over the last six years is consistent with the order of magnitude of estimates of 

CDFs from the IPEs. [6] However, due to the limitations discussed above, the ASP results are 

not sufficient to verify the IPE CDF results. 

In last year's report, we presented the results of a review of precursor results for the period 

1994-1997 and showed that a number of these events involved event initiators or conditions 

that are not included in the IPEs. These events comprise approximately 15 percent of the 

precursors for this period. While CDF implications from the ASP Program are in general 

agreement with overall estimates of CDF, occasionally key contributors to precursors differ from 

those identified in PRAs. Of the ten precursors currently under consideration for 1998, 

the following conditions occurred that are not typically considered in the IPEs: 

* Potential failure of the recirculation mode of ECCS operation because of 

calibration and calculational errors in emergency sump and borated storage tank

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level instrumentation. This specific condition was cited in Information Notice 98

40. [4,6] 

0 Potential failure of all component cooling water pumps due to steam intrusion 

resulting from a postulated high energy line break. This preliminary result is still 

undergoing technical review, and issuance of an information notice will be 

considered based on the results of the review. [4,6] 

Evaluation of Risk Associated with Issues at D. C. Cook Plant 

As a result of the large number of potential safety issues identified at D. C. Cook Units 1 and 2 

(Cook) and the increased regulatory activities since August 1997, RES conducted an analysis 

of the risk significance of the issues using the Accident Sequence Precursor (ASP) Program 

methodology. To date, we have completed the analysis of 94 out of 117 individual issues. 

Preliminary results indicate that one of these is risk-significant (potential ASP 

precursor). [4,5,8] A draft report detailing the results of the analyses of the individual issues 

and the integrated analysis of all issues will be issued for review and comment in December. 

The final report will be prepared after all comments on the draft report have been resolved. 

Issuance of the final report is currently scheduled for April 2000. 

The one issue at Cook that has been identified as a potential precursor involves the potential 

failure of all component cooling water pumps due to steam intrusion resulting from a postulated 

high energy line break identified above. The results of the preliminary analysis of this issue 

were sent to the Cook licensee for review and comment. When all comments on the 

preliminary analysis have been resolved, the final analysis will be prepared and issued. 

1999 ASP Event Analysis 

The ASP screening, review, and analysis of 1999 operational events began in May 1999. We 

have received 628 1999 LERs, which represents about 40% of the total number of LERs (1500) 

anticipated for the whole year. We selected 240 of these for engineering review as potential 

precursors. Our review identified 23 individual events for detailed analysis to determine if they 

are precursors. We completed eleven preliminary analyses and identified one event as a 

potential precursor. This event is currently receiving peer review by the licensee and the staff. 

Redirected ASP Program Coordination 

Recent developments in the nuclear regulatory arena have resulted in new or different agency 

activities: 

* The NRC's transition to a risk-informed regulatory framework. 

0 Development and implementation of the improved reactor oversight process and the 

associated significance determination process. 

* Approval and implementation of Regulatory Guide 1.174. 

These developments necessitated a reevaluation of the coordination of the ASP Program and 

its associated methods and model development efforts. In addition, the efforts involving

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development of SPAR models and other analytical tools which had been initiated before these 

activities were undertaken are being reassessed to make sure that they meet current and future 

needs of agency users. 

In an effort to ensure continuation of user input to the development efforts that support the ASP 

Program, OERAB formed the SPAR Models Users Group (SMUG). This group consists of 

representatives from each of the organizations within the agency's program offices that use risk 

models in the performance of their regulatory activities. This group includes a senior reactor 

analyst (SRA) from each regional office, staff from each branch within RES and NRR that is 

involved with PRA applications, and staff from each branch involved in the implementation of 

the agency's new reactor oversight process. The SMUG was formed to provide the 

coordination of activities that the former ASP Program Technical Coordination Group 

(ASPPTCG) was not designed to handle. The SMUG has met three times - on September 29, 

1999, on October 21, 1999, and on November 17, 1999 - to agree on the objectives and charter 

for the group, review the current status of the development of methods and model capability in 

each of the major areas identified above, and to identify the methods and model needs of each 

of the participating organizations in SMUG along with their associated priorities. 

One of the functions of the former ASPPTCG had been the preparation of the annual update of 

the ASP Program Plan (originally issued as Enclosure 1 to SECY-94-076 in March 1994; the 

most recent update, Revision 4, was issued in December 1998). Scheduling and tracking of 

SPAR model development are now part of the RES Operating Plan. This eliminates the need 

for a separate ASP Program Plan. 

Model and Methods Development 

As mentioned above, OERAB is now responsible for development of methods and models that 

support the ASP Program and other regulatory activities. The staff currently uses 72 simplified, 

plant-specific, Level 1 models [the Revision 2QA standardized plant analysis risk (SPAR) 

models] in analyses performed during its risk-informed regulatory activities. Ongoing 

development activities include: 

* The Level 1 SPAR models. 

0 Low power/shutdown models. 

* Capability to perform Level 2/3 precursor analysis of events. 

* Capability to estimate the large early release frequency (LERF) associated with 

operational events. 

* Capability to estimate the relative risk associated with external events. 

Since last year, the following accomplishments were achieved in these areas: 

Standardized Plant Analysis Risk (SPAR) Models

* Continued maintenance of the existing Revision 2QA SPAR models.

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* Developed a preliminary on-site review process for the Revision 3 SPAR models. 

The Revision 3 SPAR models have been developed from the Revision 2QA 

models by adding: 

The modeling of support systems (e.g., electrical power systems, component 

cooling water, service water). 

A recovery model that includes an improved human performance analysis 

method. 

An improved method for treating common cause failures. 

A method for treating the parameter uncertainty associated with each basic 

event in the models. 

Models for additional initiating events (e.g., large LOCA, medium LOCA). 

* Completed the on-site review (using the preliminary process) and generated 

three draft Revision 3 SPAR models (Calvert Cliffs, Duane Arnold and 

Millstone 2). 

Large Early Release Frequency (LERF) Models 

* Completed the initial draft of 6 PWR and 2 BWR LERF models and containment 

parameter databases for the remaining plants. 

* Completed containment phenomenology scoping which identified the 

phenomenology to be updated based on recent research/studies. 

* Completed code changes for the decomposition of LERF model results to 

identify the contribution from the Level 1 portion of the plant model. 

FUTURE STATUS REPORT: 

The 1998 Precursor Report, "Precursors to Severe Core Damage Accidents: 1998 A Status 

Report," NUREG/CR-4674, Volume 27, is scheduled to be issued by mid-January 2000. The 

next report to the Commission on the status of the ASP Program will be provided in December 

2000. 

REFERENCES: 

1. U. S. Nuclear Nuclear Regulatory Commission, "Risk Assessment Review Group 

Report," NUREG/CR-0400, Washington, D. C., September 1978. 

2. Dale M. Rasmuson and Patrick D. O'Reilly, "Analysis of Annual Accident Sequence 

Precursor Occurrence Rates for 1984-94," Proceedings of the International Topical

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Meeting on Probabilistic Safety Assessment, Park City, Utah, September 29 - October 

3,1996, Volume III, pp.1645 -1652. 

William D. Travers 

Executive Director 

for Operations 

Attachments: 

1. ASP Program Precursors for 1998 

2. Conditional Core Damage Probability Results from ASP Program (Chart) 

3. Update of Statistical Analysis of Trends in Precursor Occurrence Rate (Four Charts) 

DISTRIBUTION: 

Commissioners 

OGC 

OCAA 

01G 

OPA 

OCA 

ACRS 

CIO 

CFO 

EDO 

REGIONS 

SECY

9
1998 At-Power Precursors Involving Initiating Events Sorted by Conditional Core Damage Probability 

CCDP Plant Plant Event Identifier Description Event 

Type Date 

5.6 x 10'4 Davis-Besse PWR LER 346/98-006 A Tornado Touchdown Causes a Complete 6/24/98 

Loss of Offsite Power 

* 1.5 x 10-5 Davis-Besse PWR LER 346/98-011 Manual Reactor Trip Due to Component 10/14/98 

Cooling System Leak and the De

Energizing of Safety-Related Bus D I and 

Non-Safety-Related Bus D2

* Preliminary results. Final analysis review in progress.

1998 At-Power Precursors Involving Unavailabilities Sorted by Increase in Core Damage Probability 

ACDP Plant Plant Event Identifier Description Event 

Type Date 

*2.2 x 10' D. C. Cook 1 and 2 PWR LER 316/98-005 Potential for High Energy Line Break to 7/15/98 

Degrade Component Cooling Water System 

1.1 x 10"1 Big Rock Point PWR LER 155/98-001 Long-Term Unavailability of Liquid Poison 3/27/98 

Control System 

7.2 x 10-6 San Onofre 2 PWR LER 361/98-003 Inoperable Sump Recirculation Valve 2/5/98 

5.6 x 10-6 Byron 1 PWR LER 454/98-018 Long-Term Unavailability of an Emergency 9/12/98 

Diesel-Generator 

1.7 x 10-6 Oconee 1, 2, and 3 PWR LER 269/98-004 Calibration and Calculational Errors 2/12/98 

(Units 1, 2) Compromise Emergency Core Cooling 

1.4 x 10-6 System Transfer to Emergency Sump 

(Unit 3) 1 _ 1_1 

* Preliminary results. Final analysis review in progress.

Attachment 1
CONDITIONAL CORE DAMAGE PROBABILITY RESULTS

FROM ASP PROGRAM

NO. OF PRECURSORS

1984 1985 1986 1987 1988 1989 1990* 1991 1992 1993 1994 1995 1996 1997 1998"*

YEAR

* THE 3/20/90 VOGTLE EVENT HAS BEEN ROUNDED UP FROM 9.7E-4 AND PLOTTED AS 1.OE-3. 

PRELIMINARY RESULTS.

20

15

10

5

0

CD 

=>
19-01

1E-02 

1E-03

8 

0

19-4 8

Year

Figure 1. Trend for events with exponent = -3. Trend is not 

statistically significant (p-value = 0.1307)

l -- - 5% ufr0Ity L9,8 -- . .% Unre(i0ty

0 

8

19-02

84 85 88 87 88 89 90 91 92 93 04 85 96 97 98 

Year

Figure 3. Trend for events with exponent = -5. Trend is 

statistically significant (p-value = 0.0001)

19I00 Q Yo."yU, Jt~ L .I.k 84(81 

- ~ k..~Y ~ k888 W ----- h.011 9

0

1E-01 

19-0

1E-8 85 88 87 88 89 90 91 82 93 94 95 98 87 98 

Year

Figure 2. Trend for events with exponent = -4. Trend is 

statistically significant (p-value = 0.0001)

1E-00

0

19-01

19.02

Year

Figure 4. Trend for events with exponent = -6. Trend is 

statistically significant (p-value = 0.0001)

0 

(D 

Z),

84 Be