December 23, 1997 SECY-97-296

FOR: The Commissioners

FROM: L. Joseph Callan /s/

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 and to present (1) the results of the final ASP analyses of 1996 operational events, (2) the results of the preliminary analyses of 1997

operational events completed to date, and (3) an update of the staff’s progress in ASP-related methods and model development efforts. This

paper also responds to a request from a former Commissioner (Rogers) to present a yearly update of annual quantitative ASP results.

SUMMARY:

Since the last status report, SECY-96-241, dated November 26, 1996, the staff has made progress in the following areas: (1) ASP Program

coordination, (2) 1996 and 1997 ASP event analysis, (3) 1982 and 1983 ASP event analysis, (4) evaluation of ASP results and trending,

(5) improvements to existing models, and (6) development of new models. Budget redirections have dictated that completion of improved ASP

models be postponed until November 2001. Other deferrals or cancellations may be necessary; decisions will be made in the near future. 

DISCUSSION:

The Office for Analysis and Evaluation of Operational Data (AEOD), the Office of Nuclear Reactor Regulation (NRR), and the Office of Nuclear

Regulatory Research (RES) have continued to make progress in a number of areas, which are discussed below.

Contact:

Patrick O'Reilly, AEOD

415-7570
The Commissioners 2

ASP Program Coordination

The interoffice (AEOD, NRR, and RES) ASP Technical Coordination Group continued to meet regularly to discuss experiences with the

analysis of operational events. The group also provided guidance regarding the technical direction of several other efforts. These include 

(1) systematic quality assurance and review of the Revision 2 simplified plant analysis risk (SPAR) models used in ASP analyses, (2)

development of improved (Revision 3) SPAR models, and (3) development of Level 2/3 analysis capability. In addition, the group produced the

annual update (Revision 3) of the Integrated ASP Program Plan for the ASP-related activities conducted by each office.

1996 ASP Event Analysis

The screening, review, and analysis of 1996 operational events have been completed. A total of 11 precursors affecting 14 units were

identified for 1996. Eight of the 1996 precursors involved initiators; six involved the unavailability of equipment. This result differs from each of

the past 4 years, in which conditional unavailability events greatly outnumbered initiators. Five of the 1996 precursors (~36 percent) involved

problems with electrical equipment. This result differs from the previous 5-year period (1990695), when about 60 percent of the precursor

events each year involved electric power issues. Besides electrical problems, the 1996 precursors involved problems with auxiliary feedwater,

design problems, and human performance issues. The 1996 precursor results are presented in Attachment 1. 

The following three 1996 precursor events had conditional core damage probabilities (CCDP) $1.0×10-4. 

 a loss of offsite power (LOOP) with emergency diesel generator B unavailable at Catawba 2 on February 6, 1996 (The estimated CCDP

for this event was 2.1 x 10-3, which was the highest CCDP for any of the 1996 precursor events.); 

 a reactor trip with loss of train A of essential service water due to frazil ice formation and loss of the turbine-driven auxiliary feedwater pump

at Wolf Creek on January 30, 1996 

(The estimated CCDP for this event was 2.1×10-4.); and

 a condition that was discovered at the Haddam Neck plant on August 8, 1996, involving potentially inadequate reactor heat removal system

pump net positive suction head following a postulated large- or medium-break loss-of-coolant accident. (This event had an estimated

CCDP of 1.1×10-4.)

Evaluation of ASP CCDP Results and Trending

The ASP results were reviewed to obtain insights about industry risk. This review consisted of analysis of trends in ASP events, comparison of

an Annual ASP Index based on the past 5 years of ASP events with core damage frequencies (CDF) from individual plant examinations (IPE),

and comparison of the modes and causes of ASP events with those modeled in probabilistic risk assessments (PRA) and IPEs. An updated

chart showing CCDP “probability bins” for ASP results from 1984 through 1996 is contained in Attachment 2. 

A paper was presented by the staff at the “PSA ‘96" conference in Park City, Utah, in October 1996, reporting the results of a statistical

analysis of trends in the occurrence rate of precursors identified by the ASP Program between 1984 and 1994. Statistically significant

decreasing trends were found for all of the ASP CCDP bins except the events with CCDP $1×10-3. There was no statistically discernible trend

in the occurrence rate of this group of precursors. The occurrence rate of precursors with CCDP $1.0×10-3 based on this data is approximately

one event every 2 years, although there have been years in which two such events occurred. The Catawba 2 LOOP in February 1996

occurred 2 years after the previous 10-3 precursor (the blowdown of the reactor coolant system to the refueling water storage tank during hot

shutdown at Wolf Creek in 1994), and it is consistent with previous experience with the occurrence of 10-3 precursors. A review of the nature,

modes, causes, and systems affected by the precursors with CCDP $10-3 revealed that the events in this group appear to exhibit no common

(generic) characteristics. 

The following provides an 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. In addition, 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 (RY). For the last 5 years, the index is the following:

Year Annual ASP Index

1996 2.4x10-5/RY

1995 2.2x10-6/RY

1994 3.0x10-5/RY

1993 8.9x10-6/RY

1992 1.5x10-5/RY

 1.6x10-5/RY Average

This average index is consistent with an industry average CDF in the range of about 10-5/RY to 10-4/RY.

The estimated CDFs in the IPEs range from 1.2x10-6/RY to 3.7x10-4/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. It is observed that the order of magnitude of the

index over the last five years is consistent with the order of magnitude of estimates of CDFs from the IPEs. On an industry-wide level, the ASP

results do not provide indication that the IPEs have underestimated their plants’ core damage frequencies. However, due to the limitations

discussed above, the ASP results are not sufficient to verify the IPE CDF results.
3

In addition, a review of the precursor events for the period 1994-1996 revealed that a number of these events involved event initiators or

conditions that are not included in the IPEs. These include the following:

 blowdown of the reactor coolant system to the refueling water storage tank during hot shutdown,

 a reactor trip with loss of one train of essential service water due to frazil ice formation and the unavailability of the turbine-driven auxiliary

feedwater pump, and

 discovery of the potential for inadequate reactor heat removal pump net positive suction head following a large-break or medium-break loss

of coolant accident.

Other precursor events included specific failure causes that could have led to failure probabilities that are higher than are typically assumed,

either explicitly or implicitly, in the IPE models. Examples include the following:

 the failure of a safety-relief valve to close and failure of a suppression pool strainer and

 the fouling of cooling water systems due to concrete sealant injection.

These events comprise approximately 15 percent of the precursors for this period and indicate that, while CDF implications from the ASP

Program are in general agreement with overall estimates of CDF, occasionally key contributors to these results may be somewhat different

from those identified in PRAs. 

1997 ASP Event Analysis

The ASP screening, review, and analysis of 1997 operational events began in May 1997. Of the more than 1,000 licensee event reports 

(LER) received to date, 377 LERs received engineering review as potential precursors. Of these, 23 individual events have been identified for

detailed analysis to determine if they are potential precursors. The preliminary analyses of two events have been completed, and the results

have been transmitted to the respective licensees and to the NRC staff for review and comment. 

198261983 ASP Event Analysis

The review and analysis of 198261983 LERs to obtain the 2 years of precursor data which had previously been missing from the ASP results

were completed during the first quarter of 1997. The 198261983 Precursor Report was published in April 1997 as NUREG/CR-4674, Volume

24. In this effort, over 2,100 LERs were reviewed for precursors. Detailed analyses were performed on 435 LER with 109 precursors

identified — 54 for 1982 and 55 for 1983. There were no new issues identified as a result of this effort. Fifteen of the 109 events had CCDPs 

$1.0×10-4. There was one event with a CCDP $1.0×10-3 (a reactor trip with failure of automatic reactor trip capability at Salem 1 in February

1983). Other important precursors for 198261983 include (1) a reactor trip with a degraded auxiliary feedwater system and an inoperable

pressurizer power-operated relief valve at Robinson 2 in April 1983, (2) a steam generator tube rupture with one power-operated relief valve

failed open at Ginna in January 1982, and (3) a scram with loss of both residual heat removal service water loops for 8 hours due to common

cause failure at Brunswick 2 in February 1982. Most of the boiling-water reactor precursors involved the loss of long-term decay heat removal

capability. The pressurized-water reactor precursors were dominated by LOOP events (emergency diesel-generator and other electrical

problems) and reactor trips with a degraded auxiliary feedwater system.

Model and Methods Development

Progress in model and methods development by RES included the continuation of work on the initial Windows version of the SAPHIRE

(SAPHIRE 6.0) suite of PRA codes. Specific members of the ASP Technical Coordination Group were selected to form a beta testing group

for the hardware and software required to run SAPHIRE 6.0. A workshop on the composition and use of SAPHIRE 6.0 was held for staff

analysts in January 1997. Additionally, initial steps in the construction of a data link with Idaho National Engineering and Environmental

Laboratory were completed, and staff analysts were tied in to the data link for checkout purposes. This data link will result in easier

accessibility to the PRA codes and the Simplified Plant Analysis-Risk (SPAR) and detailed PRA plant models for staff analysts.

The systematic quality assurance (QA) review of the 73 Revision 2 SPAR models continued, with completion of this effort scheduled for the

end of calendar year 1997. In order to provide technical support for some of the modeling approaches, thermal-hydraulic calculations were

performed using RELAP5 with human performance data input obtained from the results of runs performed on the reactor simulators at the

NRC’s Technical Training Center. 

The ASP Technical Coordination Group reviewed the results from the QA of the Revision 2 SPAR models and the original scope for the

Revision 3 SPAR models to identify those issues which should be addressed during the development of the 73 Revision 3 models. The Group

determined that better models were necessary for certain systems (e.g. support systems) and that more detailed QA is required. RES revised

the scope of the Revision 3 model development effort accordingly. The revised work scope and funding limitations have resulted in the

preliminary schedule for completion of the Revision 3 models (including QA and checkout) extending to November 2001, instead of August

1998, as projected in the November 1996 ASP Program Plan. RES is investigating options to accelerate completion of the models.

Due to funding and staff limitations, the status of two key projects sponsored by RES that support the ASP Program is currently under review. 

These two projects are (1) development of models to analyze events occurring during low power/shutdown operation and flooding events

occurring during full power operation and (2) models to analyze internal fires, flooding, and seismic events. A decision on the future direction

of these projects will be made by RES in January 1998.
4

FUTURE STATUS REPORT:

The 1996 Precursor Report is scheduled to be published by December 31, 1997. The next report to the Commission on the status of the ASP

Program will be provided in November 1998. 

L. Joseph Callan

Executive Director 

 for Operations

Attachments:

1. ASP Program Precursors for 1996

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

cc w/atts.:

SECY

OGC

OCA

OPA

CFO

CIO

OIP
6

FUTURE STATUS REPORT:

The 1996 Precursor Report is scheduled to be published by December 31, 1997. The next report to the Commission on the status of the ASP

Program will be provided in November 1998. 

L. Joseph Callan

Executive Director 

 for Operations

Attachments:

1. ASP Program Precursors for 1996

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

cc w/atts.:

SECY

OGC

OCA

OPA

CFO

CIO

OIP

Distribution w/atts.:

File Center FCongel ERodrick, RES SRichards, NRR

Public KRaglin MDrouin, RES JTappert, NRR

OEDO R/F DHickman ELois, RES GHolahan, NRR

(WITS No. 92-00101) JRosenthal JFlack, NRR

AEOD R/F DRasmuson SLong, NRR

SPD R/F WHodges, RES DO’Neal, NRR

JBlaha, AO HVandermolen, RES MPohida, NRR

HThompson, DEDR EChow, RES DSkeen, NRR

TOMartin, OEDO MCunningham, RES RDennig, NRR

C:\1997-296scy.wpd 

*See previous concurrence *See previous concurrence+

To receive a copy of this document, indicate in the box "C" copy w/o attach/encl "E" copy w/attach/encl "N" no copy

OFFICE RRAS/RRAB E RRAS/RRAB E RRAB E SPD E AEOD

NAME PO’Reilly:mmk* SMays* PBaranowsky

*

CERossi* TTMartin

DATE 11/18/97 11/18 /97 11/18/97 11/18/97 12/ /97

OFFICE NRR RES E DEDE EDO

NAME SCollins MKnapp* AThadani LCallan 

DATE 12/ /97 11/18/97 12/ /97 12/ /97

OFFICIAL RECORD COPY
Attachment 1

1996 At-Power Precursors Involving Initiating Events Sorted by Conditional Core Damage Probability

CCDP Plant Plant

Type

Event Identifier Description Event

Date

2.1 x 10-3 Catawba 2 PWR LER 414/96-001 Loss of Offsite Power with Emergency

Diesel-Generator B Unavailable

2/6/96

2.1 x 10-4 Wolf Creek PWR LER 482/96-001, 

-002

Reactor Trip with Loss of Train A of

Essential Service Water and the TurbineDriven Auxiliary Feedwater Pump

Unavailable

1/30/96

5.3 x 10-5 Prairie Island 1

and 2

PWR LER 282/96-012 Loss of Power to Safeguards Buses on Both

Units

7/29/96

7.0 x 10-6 LaSalle 1 and 2 BWR LER 373/96-007,

-008

Concrete Sealant Fouls Cooling Water

Systems

6/28/96

*1.1 x 10-5 Arkansas Nuclear

One, Unit 1

PWR LER 313/96-005 Reactor Trip and Subsequent Steam

Generator Dryout

5/19/96

*Preliminary results, Final results under review.

1996 At-Power Precursors Involving Unavailabilities Sorted by Increase in Core Damage Probability

CDP Plant Plant

Type

Event Identifier Description Event

Date

1.1 x 10-4 Haddam Neck PWR LER 213/96-016 Potentially Inadequate RHR Pump NPSH

Following a Large- or Medium-Break LOCA

8/1/96

4.6 x 10-5 Seabrook PWR LER 443/96-003 Turbine-Driven Emergency Feedwater Pump

Unavailable Because of a Mechanical Seal

Failure

5/21/96

5.8 x 10-6 Salem 1 PWR LER 272/96-002 Charging Pump Suction Valves from the

RWST Potentially Unavailable Because of

Pressure Locking

1/10/96

2.9 x 10-6 Haddam Neck PWR LER 213/96-024 After a Residual Heat Removal Pump

Seized, It Was Determined to Be Susceptible

to Failure Since Being Overhauled in 1987

9/1/96

1.8 x 10-6 McGuire 2 PWR LER 370/96-002 2B Emergency Diesel Generator Inoperable

Due to Slow Instrumentation Response

3/6/96

1996 Shutdown Precursors Involving Initiating Events Sorted by Conditional Core Damage Probability

CCDP Plant Plant

Type

Event Identifier Description Event

Date

1.7×10-5 Byron 1 PWR 454/96-007 Transformer Bus Fault Causes a LOOP 5/23/96
Attachment 2
ROUTING AND TRANSMITTAL SLIP Date 12/13/99

TO: (Name, office symbol, room #, building, agency/post) Initials Date

1. P. O’Reilly - Concur 

2. S. Mays - Concur

3. P. Baranowsky - Concur

4. C. E. Rossi - Concur

6. T. T. Martin - Concur

7. S. Collins - Concur

8. M.Knapp - Concur

9. A. Thadani - Concur 

10. L. Callan - Signature

11. Nancy - Distribute

Action File Note and Return

Approval For Clearance Per Conversation

As Requested For Correction Prepare Reply

Circulate For Your Information See Me

Comment Investigate X Concurrence/Signature

Coordination Justify

REMARKS

Status Report on Accident Sequence Precursor Program and Related Initiatives

FROM: (Name, org. symbol, Agency/Post)

Pat O’Reilly, AEOD

Room # - Bldg. M.S. T-4A9

Phone # 415-7570