NUREG-1780

Regulatory Effectiveness of

the Anticipated Transient

Without Scram Rule

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Washington, DC 20555-0001
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NUREG-1780

Regulatory Effectiveness of

the Anticipated Transient

Without Scram Rule

Manuscript Completed: April 2001

Date Published: September 2003

Prepared by

W.S. Raughley, G.F. Lanik

Division of Systems Analysis and Regulatory Effectiveness

Office of Nuclear Regulatory Research

US. Nuclear Regulatory Commission

Washington, DC 20555-0001
ABSTRACT

The U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research is reviewing

selected regulations to determine if the requirements are achieving the desired outcomes. This

initiative is part of an evolving program to make NRC activities and decisions more effective,

efficient, and realistic. The goal of this evaluation is to determine whether the anticipated

transient without scram (ATWS) rule and the recommendations issued with it were effective.

The effectiveness of the ATWS rule was determined by comparing regulatory expectations to

outcomes. The report concluded that the ATWS rule was effective in reducing ATWS risk and

that the cost of implementing the rule was reasonable. However, uncertainties in reactor

protection system reliability and mitigative capability may warrant further attention to ensure the

expected levels of safety are maintained.

i
CO NTENTS

ABSTRACT ............................ iii

EXECUTIVE SUMMARY ............................ vii

FOREWORD ............................ xi

ABBREVIATIONS......................................................... xiii

1 INTRODUCTION ............................ 1

2 BACKGROUND ............................. 1

2.1 ATWS Risk. 2

2.1.1 Boiling-Water Reactor ATWS Sequence and Event Trees. 2

2.1.2 Pressurized-Water Reactor ATWS Sequence and Event Trees. 4

2.1.3 Operator Action .5

2.2 Commission Recommendations for Reducing the Risk From An ATWS. 6

2.3 NRC Regulatory Analysis for the ATWS Rule .7

3 ASSESSMENT OF THE ATWS RULE ............................ 7

3.1 Method for Assessing Regulatory Effectiveness of the ATWS Rule.. 7

3.2 Comparison of Expectations and Outcomes .10

3.2.1 Modifications and Operating Limitations .10

3.2.2 Comparison of ATWS Rule Risk Expectations and Outcomes .11

3.2.3 Reactor Protection System Reliability Validity .13

3.2.4 Risk Insights From Licensee Probabilistic Risk Assessmentindividual

Plant Examinations . 14

3.2.5 Insights From NRC Reliability Studies and Operating Experience .... 15

3.2.6 Changes in Fuel Management May Affect Pressurized-Water Reactor

ATWS Mitigating Capability. 1 6

3.3 Value-Impact .19

4 CONCLUSIONS ............ 20

5 REFERENCES ............ 21

APPENDICES

A Anticipated Transient Without Scram Rule Event Trees .A-1

B Plant-Specific and General ATWS Information by Reactor Group .B-1

C Resolution of Comments .C-1

v
FIGURES

A-1 ATWS Rule Event Tree for General Electric Reactor Group ..... .............. A-1

A-2 ATWS Rule Event Tree for General Electric Reactor Group Automatic

Standby Liquid Control System ........... ............................... A-3

A-3 ATWS Rule Event Tree for Westinghouse Reactor Group . . A-4

A-4 ATWS Rule Event Tree for Combustion Engineer/Babcock & Wilcox Reactor Group A-6

TABLES

1 Summary of ATWS Rule Expectations and Outcomes ........................ 10

2 ATWS Rule Modifications ................. ............................. 11

3 Summary of ATWS Rule Risk Expectations and Outcomes .................... 12

4 Alternate Estimates of Reactor Protection System Reliability ................... 14

5 Control Rod Insertion Events .............. .............................. 17

6 ATWS Moderator Temperature Coefficient and Peak Pressure

for Pressurized-Water Reactors ......................... 17

7 ATWS Rule Value-impact Summary ......................... 19

A-1.1 Data for Boiling-Water Reactor ATWS Isolation Transient .A-2

A-1.2 Data for Boiling-Water Reactor ATWS Nonisolation Transient .A-2

A-2.1 Data for Boiling-Water Reactor ATWS With Automatic

Standby Liquid Control System Initiation .A-3

A-3.1 Data for Westinghouse Pressurized-Water Reactor ATWS Turbine Trip Transient . A-5

A-3.2 Data for Westinghouse Pressurized-Water Reactor

ATWS Non-Turbine Trip Transient .A-5

A-4.1 Data for Combustion Engineering/Babcock & Wilcox ATWS Transient .A-6

B-1 Operating Pressurized-Water Reactor Data ............................... B-1

B-2 Operating Boiling-Water Reactor Data ................................... B-9

B-3 Reactor Protection System Unreliability Uncertainties ....................... B16

B-4 General Electric Reactor Protection System Common-cause Failure Insights

(NUREG/CR-5500, Volume 3) ............ ............................. B-16

B-5 Westinghouse Reactor Protection System Common-cause Failure Insights

(NUREG/CR-5500, Volume 2) ......................................... B-16

B-6 Babcock & Wilcox Reactor Protection System Common-cause Failure Insights

(NUREG/CR-5500, Volume 4) ........... ............................. B-17

B-7 Combustion Reactor Protection System Common-cause Failure Insights

(NUREG/CR-5500, Volume 10) ........... ............................ B-17

B-8 Value-impact Data for Each Reactor Group .............................. B-18

vi
EXECUTIVE SUMMARY

As part of the U.S. Nuclear Regulatory Commission program to address regulatory

effectiveness, the Office of Nuclear Regulatory Research (RES) is reviewing selected

regulations to determine if the requirements are achieving the desired outcomes.

SECY-97-180, Response to Staff Requirements Memorandum of May 28, 1997, Concerning

Briefing on IPE Insight Report," August 6, 1997, describes a plan for the RES staff to assess

the effectiveness of several major safety issue resolution efforts.

An anticipated transient without scram (ATWS) is an anticipated operational occurrence

followed by failure of the reactor trip portion of the protection system. The likelihood of core

damage from an ATWS depends on three factors: (1) the initiating event frequency, (2) the

reliability of the reactor protection system (RPS), and (3) the reliability of ATWS mitigation

systems.

During ATWS rule development there was considerable disagreement about the reliability of the

RPS. Compared to other systems, the RPS is quite reliable - the failure rate is likely less than

one in ten thousand demands. However, the strong dependence of ATWS risk on RPS

reliability and the uncertainty associated with the value of RPS reliability were major factors in

the decision to adopt the ATWS rule.

The ATWS rule required the installation of hardware to improve the nuclear plant's capability to

prevent an ATWS and mitigate its consequences. The Commission also issued two

recommendations with the ATWS rule to (1) reduce the number of automatic scrams, and

(2) improve RPS reliability.

The goal of this assessment is to determine whether the ATWS rule and other relevant

Commission recommendations issued with the ATWS rule were effective in achieving the

desired outcome and whether certain areas may need attention. For the purposes of this

assessment, the regulatory documents are considered effective if the expectations (desired

outcomes) are being achieved. The expectations were established from objective measures

stated in the ATWS rule and accompanying regulatory documents and compared to the

outcomes in the areas of system modifications and operating limitations, risk, and value-impact.

The outcomes were obtained from reviews of documents and operating experience after the

issuance of the ATWS rule. The value-impact assessment determines if the industry's costs to

implement the ATWS rule were reasonable.

The assessment concludes that the ATWS rule has been effective in installing modifications,

reducing ATWS risk, and implementing the rule at reasonable cost. However, uncertainties in

RPS reliability and mitigative capability warrant continued attention consistent with the NRC

performance goals to maintain the expected levels of safety and to improve effectiveness. To

elaborate:

Hardware modifications and operating limitations required by the ATWS rule were

implemented. All pressurized-water reactors (PWRs) installed diverse means to trip the

turbine and initiate auxiliary feedwater. Combustion Engineering, Inc. and Babcock &

Wilcox Co. plants installed a diverse scram system. Westinghouse plants generally

maintain an unfavorable moderator temperature coefficient (MTC)" of 1 percent.

vii
Boiling-water reactor (BWR) plants implemented diverse recirculation pump trip,

alternate rod insertion circuitry, and upgraded emergency operating procedures; or

installed high capacity standby liquid control systems.

* The mean frequency of automatic scrams (initiating events for ATWS) decreased from

approximately 4/reactor-years in 1983 to 0.5/reactor-years since 1997. This alone

accounts for a reduction of nearly one order of magnitude in the frequency of an

ATWS - probability of an ATWS.

* RPS reliability dominates the risk from an ATWS. There have been no total failures of

the RPS system since the ATWS rule was issued in 1984. Point estimates of RPS

reliability, based on operating experience since 1984 show that the mean RPS

unreliability (one minus RPS reliability) expectations have been met for all four reactor

groups and are approximately an order of magnitude better than the RPS reliability

estimates before the ATWS rule. These generic estimates were developed using a

probabilistic risk assessment model for the RPS and failure rates of components.

* PWR scram system reliability is related to reactor trip breaker reliability. As evidenced

by NRC generic communications and industry group activities, circuit breaker problems

continue to occur. Industry programs to maintain scram system reliability continue to be

useful in limiting risk from ATWS.

* During the ATWS rulemaking the NRC staff set a goal that probability of an ATWS

should be no more than 1 .OE-05/reactor-year. Probability of an ATWS was defined as

the annual frequency of an ATWS leading to plant conditions that exceed certain design

parameters that can result in core melt, containment failure, and the release of

radioactivity and can be viewed as the expected CDF of an unmitigated ATWS.

Updating the original generic ATWS regulatory analysis, using operating data since the

ATWS rule was implemented, found that on a generic basis, all four reactor types

achieved the ATWS rule risk goal. A few individual plants had somewhat higher risk as

a result of using lower levels of RPS reliability rather than the industry average or

assumed additional equipment failures.

* Comparison of the estimated value-impact expectations in the original ATWS regulatory

analysis to the corresponding outcome indicates that the costs to implement the ATWS

rule were less than expected. This is largely due to fewer than expected spurious

scrams caused by ATWS equipment than assumed in the original analysis.

* The higher fuel bumup has resulted in previously unpredicted oxide growth and fuel

assembly distortion. In some cases this has resulted in slow or incomplete control rod

insertion. These failures and degradations are new phenomena and were not

considered during the development of the ATWS rule.

Although past data indicates that the risk from ATWS is in the range foreseen when the ATWS

rule was issued, several issues have the potential to erode past achievements. Attention to

these issues and regulatory actions that maintain compliance with current regulations can

assure that the risk from ATWS remains acceptable. These issues are:

viii
* RPS reliability estimates are subject to large uncertainties. Current point estimates

developed using RPS probabilistic risk assessment models show upper and lower

bounds of unreliability ranging from 1 .8E-6 to 5.7E-5. RPS reliability requirements are

so high and ATWS events are so rare that many more years of operating experience

are needed to generate sufficient system demands to reduce current estimates of the

uncertainty. Licensee's risk calculations in support of licensing actions that could affect

ATWS risk should address these uncertainties.

* ATWS mitigation capability on a PWR is highly dependent on the MTC. Mitigative

functions are considered by the ATWS rule regulatory basis to be non-viable if the

ATWS peak pressure exceeds 3200 psig; and a sufficiently negative MTC will limit the

ATWS peak pressure. Fuel design to achieve longer cycles and higher power ratings

may result in less negative MTCs at full power for a larger fraction of the cycle time,

during which time ATWS mitigation may be less effective. CE and Babcock &

Wilcox Co. reactors installed a diverse scram system to compensate for large exposure

times. Further fuel cycle changes and power upgrades that could affect the ATWS risk

may require compensatory measures (e.g., hardware or procedural), consistent with the

underlying regulatory basis behind the ATWS rule.

* ATWS mitigation on a boiling-water reactor (BWR) is highly dependent on operator

actions. Although improvements in design, procedures, and training since the ATWS

rule was issued should have contributed to improved mitigative response capability,

BWR operator response to an ATWS continues to be a challenge. Probabilistic risk

assessment/individual plant examinations for BWRs indicate large variations in the

assumptions for reliability of human actions in response to an ATWS. Similarities in

design, procedures, and training argue against such variability. Licensee's risk analyses

in support of licensing actions should be supported by technical justification of operator

performance assumptions.

ix
FOREWORD

As part of the U.S. Nuclear Regulatory Commission program to address regulatory

effectiveness, the Office of Nuclear Regulatory Research (RES) is reviewing selected

regulations to determine if the requirements are achieving the desired outcomes.

SECY-97-1 80, Response to Staff Requirements Memorandum of May 28, 1997, Concerning

Briefing on IPE Insight Report," August 6, 1997, describes a plan for the RES staff to assess

the effectiveness of several major safety issue resolution efforts.

This study evaluated the effectiveness of the anticipated transient without scram (ATWS) rule.

The study compared regulatory expectations to outcomes, focusing on system modifications,

operating limitations, risk, and value-impact to establish the technical basis for its conclusion.

The study's findings indicated that the rule reduced ATWS risk at reasonable cost. The study

also indicated that although the risk from ATWS currently remains low and within the expected

range, continued implementation and monitoring of ATWS as a contributor to risk is still

warranted. In addition to licensee compliance to the ATWS rule, the following insights should

be considered within the regulatory decision-making process, as it pertains to ATWS:

1. Reactor protection system (RPS) reliability estimates are subject to large uncertainties.

Current point estimates developed using RPS probabilistic risk assessment models

show the upper and lower bounds of unreliability ranging from 1 .8E-6 to 5.7E-5. Given

the expected high RPS reliability and the relatively small operating experience, many

more years of reactor operational experience will be needed in order to generate

sufficient data to validate or reduce uncertainties with current estimates. Risk

calculations in support of licensing actions that could affect ATWS risk should address

these uncertainties.

2. Pressurized Water Reactors (PWRs) ATWS mitigation capability is highly dependent on

the moderator temperature coefficient (MTC). Negative MTC will limit the primary

pressure during an ATWS, and reduce the likelihood of primary system damage that

could result from the pressure peak. Mitigative functions are considered by the ATWS

rule to be non-viable if the ATWS peak pressure exceeds 3200 psig. Fuel designed to

achieve longer cycles and higher power ratings result in less negative MTC, and

subsequently less effective ATWS mitigation capability. These changes offsets the

benefit gained from the ATWS rule. Consistent with the underlying regulatory basis

behind the ATWS rule, consideration should, therefore, be given to potential

compensatory measures (e.g., hardware or procedural) prior to implementing fuel cycle

changes and power upgrades that could impact the ATWS risk.

3. ATWS mitigation on a boiling-water reactor (BWR) is highly dependent on operator

actions. Although improvements in the design, procedures, and training took place in

response to the ATWS rule, BWR operator response to an ATWS continues to be a

challenge for licensees. Probabilistic risk assessment/individual plant examinations for

BWRs indicate large variations in the assumptions for reliability of human actions in

response to an ATWS. Therefore, any risk analyses used to support licensing actions

should have a strong technical basis on which to base operator performance

assumptions.

xi
The information in this report will be used in a future study on the effects of power uprates and

other plant design and operational changes on safety margins.

As described below, this report is consistent with the NRC strategic performance goals in the

areas of maintaining safety; increasing public confidence; and making NRC activities more

effective, efficient, and realistic.

Maintaining safety - The study confirms that safety has been maintained or otherwise

enhanced with licensees' implementation of the resolution of the ATWS rule and associated

plant modifications. The conclusions of the report are risk-informed and performance-based

having linked ATWS risk sequences to measurable reactor performance parameters

(e.g., MTC, number of reactor scrams). Additionally, the study addressed uncertainties in

reactor protection system reliability and mitigative capability that may warrant further

consideration to ensure the expected levels of safety are maintained. The study also relied

upon operating experience and the results of licensee and NRC risk assessments to evaluate

safety implications. If the study had found that safety had not been maintained, appropriate

recommendations would have been made.

Public confidence - Public confidence should be improved by concluding that the ATWS rule

achieved expectations through NRC and industry action. In addition, a draft of the report was

publicly available through the public document room, invited comments from the stakeholders

as peer reviewers were received, and public meeting held with the Advisory Committee on

Reactor Safety. Further, the final report is publically available, openly restates and addresses

stakeholders comments (Appendix C), and had been responsive to stakeholder involvement.

Making NRC activities more effective, efficient, and realistic - The report compares the

regulatory expectations to outcomes to assess the effectiveness of the ATWS rule in achieving

its goals, and in identifying areas that may need attention. Appropriate recommendations would

have been made if warranted in this area.

Farouk Eltawila, Director

Division of Systems Analysis and Regulatory Effectiveness

Office of Nuclear Regulatory Research

xii
ABBREVIATIONS

AFW auxiliary feedwater

AMSAC ATWS mitigating system actuation

circuitry

ARI alternate rod injection

ATWS anticipated transient without scram

B&W Babcock & Wilcox Co.

BWR boiling-water reactor

CCF common-cause failure

CDF core damage frequency

CE Combustion Engineering, Inc.

CFR Code of Federal Regulations

DSS diverse scram system

EOP emergency operating procedure

FR Federal Register

GE General Electric Co.

HEP human error probability

IN information notice

IPE individual plant examination

MSIV main steam isolation valve

MTC moderator temperature coefficient

NRC

P(ATWS)

PRA

PWR

RCS

RES

RPS

RPT

RTB

RTS

RY

Nuclear Regulatory Commission,

U.S.

probability of an ATWS

probabilistic risk assessment

pressurized-water reactor

reactor coolant system

Nuclear Regulatory Research,

Office of (NRC)

reactor protection system

recirculation pump trip

reactor trip breaker

reactor trip system

reactor-year

SLC standby liquid control

SLCS standby liquid control system

TS

UET

USI

W

technical specification

unfavorable exposure time

Unresolved Safety Issue

Westinghouse

LER licensee event report

xiii
1 INTRODUCTION

As part of the Nuclear Regulatory Commission (NRC) program to assess regulatory

effectiveness, the Office of Nuclear Regulatory Research (RES) Is eviewing selected

regulations to determine if the requirements are achieving the desired outcomes.

SECY-97-180, Response to Staff Requirements Memorandum of May 28,1997, Concerning

Briefing on IPE Insight Report," August 6, 1997 [Ref. 1], describes a plan for the RES staff to

assess the effectiveness of several major safety issue resolution efforts.

The work described in this report is an assessment of the anticipated transient without scram

(ATWS) rule. An ATWS is an anticipated operational occurrence followed by failure of the

reactor trip system (RTS) portion of the protection system. The requirements and Commission

recommendations for addressing ATWS were published on June 26, 1984, in the Federal

Register (FR) (49FR26036) as Section 50.62 of Title 10 of the Code of Federal Regulations

(CFR) (10 CFR 50.62), Requirements for Reduction of Risk from Anticipated Transients

Without Scram Events for Light-Water-Cooled Nuclear Power Plants [Ref. 2]. The

Commission intended the ATWS rule requirements to provide further assurance that failure of

the reactor to scram following an anticipated operational transient would not adversely affect

the public health and safety.

2 BACKGROUND

A number of vendor, industry, and NRC staff studies during the 1970's gave conflicting results

on the reliability of the reactor protection system (RPS) system, the probability of an ATWS

(P[ATWS]), and the core damage frequency (CDF) from an ATWS. During ATWS rule

development there was considerable disagreement about the correct or appropriate value of

RTS unavailability' which is pivotal to the ATWS issue; NUREG-460,Volume 1, Anticipated

Transient with Scram for Light Water Reactors," April 1978 [Ref. 3], indicates RPS unreliability

ranged from 3.OE-06 to 1.1 E-04 depending on the type of operating experience considered

(naval or commercial nuclear), inclusion or exclusion of failure data, and whether the

differences in pressurized-water reactor (PWR) or boiling-water reactor (BWR) RPS designs

were considered in the reliability estimates. The Commission designated ATWS as Unresolved

Safety Issue (USI) A-9, Anticipated Transient Without Scram,' to determine whether ATWS, as

a potentially significant contributor to the CDF, called for additional safety requirements.

Following precursor ATWS events at a BWR (Brown's Ferry Unit 2) in 1980 and a PWR

(Salem Unit 1) in 1983, the NRC staff completed a regulatory analysis in SECY-83-293,

Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram (ATWS) Events,"

July 19,1983 [Ref. 4], which concluded that additional ATWS safety requirements were

justified. SECY-83-293 states that a pivotal aspect of the ATWS issue is the reliability of the

RPS and provides estimates of RPS unreliability on the order of 2E-4, with large uncertainty.

The ATWS rule required the installation of hardware to improve the nuclear plant's capability to

Note, for the purpose of this assessment, the reliability of the RPS Is represented numerically by K's complement -

unreliability. This Is standard practice when describing systems of high reliability. (ie. unreliability of I E-4 rather than

reliability of .9999). Industry and NRC studies often use RPS unavailablity, unreliability, and reliability interchangeably

however, regardless of the measurement the units are almost always failures per demand.

1
I

prevent an ATWS and mitigate its consequences. The uncertainty associated with the RPS

reliability was a major factor in the decision to adopt the ATWS rule. USI A-9 was resolved with

the publication of the ATWS rule.

The ATWS rule states that PWRs designed by Westinghouse Electric Corp. (. Babcock &

Wilcox Co. (B&W), and Combustion Engineering, Inc. (CE) are required to have equipment that

is diverse, reliable, and independent from the RTS to automatically initiate the auxiliary

feedwater (AFW) system and initiate a turbine trip under conditions indicative of ATWS. This

equipment is called ATWS mitigating system actuation circuitry (AMSAC). In addition to

AMSAC, B&W and CE reactors are required to have a diverse scram system (DSS) that is

reliable and independent from the RPS as compensatory measures for higher unfavorable

moderator temperature coefficients (MTCs) to prevent potentially excessive reactor coolant

system (RCS) over-pressure.

The ATWS rule also states that BWRs, which are manufactured by General Electric Co. (GE),

are required to have a diverse aftemate rod injection (ARI) system and a diverse recirculation

pump trip (RPT) that are reliable and independent from the RPS. BWRs have a standby liquid

control (SLC) system to inject borated water into the reactor vessel and the ATWS rule also

specified a minimum SLC system injection rate and boron concentration.

The ATWS rule required licensees to submit a proposed schedule for compliance with the rule

and information sufficient to demonstrate compliance. In responding to the ATWS rule, the

owners groups for each of the four U.S. reactor vendors developed generic design packages

that were then tailored to individual plants for implementation. The NRC staff evaluated

licensees' implementations on the basis of the design information they submitted. Compliance

with the ATWS rule was typically verified by subsequent inspections.

2.1 ATWS Risk

The likelihood of core damage from an ATWS depends on three factors: (1) the initiating event

frequency (anticipated transients requiring scrams), (2) the reliability of the RPS, and (3) the

reliability of ATWS mitigation systems.

During ATWS rule development, P(ATWS) was the measure of risk proposed by the industry

and adopted by the NRC staff in SECY 83-293. P(ATWS) was defined as the annual frequency

of an ATWS leading to plant conditions that exceed certain design parameters that can result in

core melt, containment failure, and the release of radioactivity. P(ATWS) can be viewed as the

expected CDF of an unmitigated ATWS. Values of P(ATWS) shown in SECY-83-293 for the

GE, W. B&W, and CE reactor groups using the simplified event trees and data shown in

Appendix A, ATWS Rule Event Trees."

The following three subsections summarize BWR and PWR ATWS sequences and event trees,

and operation action that were used in the ATWS rule development to evaluate P(ATWS).

2.1.1 Boiling-Water Reactor ATWS Sequence and Event Trees

The BWR ATWS sequence starts with an anticipated transient and the electrical or mechanical

failure of the RPS. This is followed by a RPT to reduce power and by a turbine trip. Increased

2
RCS pressure results in the safety relief valves discharging steam to the suppression pool. The

residual heat removal (RHR) system is aligned to remove heat from the suppression pool. The

steam flow heats up the suppression pool until the reactor power level can be reduced. A

suppression pool temperature of 200 F was identified as the unacceptable plant condition used

in the ATWS rule development. The suppression pool temperature limit was set by the

resolution USI A-39, Determination of Safety/Relief Valve Pool Dynamic Loads and

Temperature Limits for BWR Containments,' in NUREG-0783, Suppression pool Temperature

Limits for BWR Containments,' November 1981 [Ref. 5]. In some cases, the NRC staff has

accepted higher temperature limits based on plant specific design features. As the suppression

pool temperature increases to near the boiling point, steam condensation is less effective,

resulting in higher containment pressure; and the RHR pumps would eventually fail due to

inadequate net positive suction pressure. During an unmitigated ATWS, with continued heat

input to the suppression pool, the containment could fail while the core was still intact.

The regulatory analysis considered three BWR sequences modeled in separate event trees.

One sequence is an isolation transient', where the main steam isolation valves (MSIVs) isolate

the bypass to the condenser and divert the steam flow to the suppression pool. The second

sequence is a nonisolation transient", where the MSIVs do not Isolate the condenser, allowing

steam flow to both the condenser and the suppression pool. The isolation transient is more

severe than the nonisolation transient because the suppression pool heats up quicker. It was

assumed in SECY-83-293 that 30 percent of transients would be isolation transients and

70 percent nonisolation transients. The third sequence applies to plants with automatically

initiated SLC systems.

Low frequency events, such as MSIV closure, which would result in core damage unless the

control rods insert immediately are not included in the risk analysis.

Operator actions are required to mitigate the isolation transient and the nonisolation transient.

The operator must enter the emergency operating procedures (EOPs) to initiate SLC and lower

the reactor water level to decrease power before the suppression pool temperature reaches the

200 F limit. It was estimated in SECY-83-293 that the operator must start SLC in

approximately 2 minutes for the isolation transient and 17 minutes for nonisolation transients to

prevent the suppression pool from heating up above its 200 F limit. The human error

probabilities (HEPs) for the operator actions were considered in SECY-83-293 based on

estimates of the time the operators had to react, shorter reaction times requiring higher HEPs.

The HEPs levels considered were 0.005-1.0 for a low-stress situation, 0.2-1.0 for high-stress

situation (both levels based on NUREG/CR-1273, Handbook of Human Reliability With

Emphasis on Nuclear Power Plant Applications,' August 1983 [Ref. 6]), and 0.01 -1 proposed

for the utility group. ATWS is generally regarded as a high-stress situation because the BWR

emergency procedure guideline/EOPs for an ATWS event typically require the operating crew

to execute several difficult steps and contingency actions simultaneously; HEPs used in

SECY-83-293 were 0.05-0.5. Although the ATWS rule required installation of RPT and ARI,

and defined levels of SLC injection, mitigation remains dependent on operator action.

Appendix A contains the simplified event trees as shown in Figure A-1, 'ATWS Rule Event Tree

for GE Reactor Group,' and Figure A-2, ATWS Rule Event Tree for GE Reactor Group With

Automatic SLC Initiation." Appendix A also shows data used to calculate P(ATWS) in each

event tree and a summary of the calculation.

3
2.1.2 Pressurized-Water Reactor ATWS Sequence and Event Trees

The PWR ATWS sequence also starts with an anticipated transient and the electrical or

mechanical failure of the RPS. In a PWR, the ATWS transient results in a RCS pressure rise,

the magnitude and timing of which is dependent on the MTC, the relief capacity, and the energy

removal capacity of the steam generators. The MTC is a measure of the reduction in the core

reactivity as the water temperature increases. The energy removal capacity of the steam

generators depends largely on the secondary side inventory. Loss of main feedwater can result

in the steam generator quickly going dry. Turbine trip, which reduces steam flow and preserves

steam generator inventory, can delay steam generator dry-out and the consequent ATWS

pressure pulse.

During an ATWS, the primary coolant temperature increases, since with the turbine trip or loss

of feedwater, heat removal is diminished while the reactor continues to generate power. For a

PWR with a negative MTC, an increase in the primary coolant temperature provides negative

reactivity feedback to reduce reactor power as the primary coolant temperature increases. The

fuel temperature coefficient, also known as the Doppler coefficient, which is always negative,

also works to reduce reactor power. However, the fuel temperature coefficient only comes into

play after the primary coolant temperature has increased because of the delay caused by the

thermal time constant of the fuel; this can be an important factor in a fast moving transient such

as ATWS. Since, as a whole, the combined steady state fuel coefficient and MTC are always

negative, reactor power tends to come into balance with energy removal, but at a higher

temperature and pressure. Higher temperature expands the primary coolant, with the potential

for causing the pressurizer to go water solid.

Primary system pressure does not increase rapidly until the steam generators near dry-out.

With loss of main feedwater, steam generator dryout would generally occur in less than two

minutes. Without loss of main feedwater as initiating or consequential event, energy removal

from the steam generator could continue until condensate was depleted and the main

feedwater system tripped.

Once the steam generators near dry-out, the pressure increase is very rapid. In this part of the

transient, since the MTC acts first, it has a major impact on the peak pressure of the

fast-moving ATWS event. However, unless core power can be reduced by inserting the control

rods or injecting boron, reactor pressure would remain high and coolant would continue to be

lost through the primary system safety valves.

Two aspects of system response need be considered: (1) the integrity of primary pressure

boundary under peak ATWS pressure and 2) the capability to inject borated water to shut down

the reactor and maintain core coverage. The peak ATWS pressure is primarily a function of the

MTC and the primary system relief capacity. The capability to inject borated water could be

affected if the initial peak pressure deformed or disabled valves in the injection path or if the

primary system pressure remained higher than the shutoff head of the high-head injection

pumps for an extended time.

In SECY-83-293, the ASME Service Level C pressure of 3200 psig was assumed be to an

unacceptable plant condition during ATWS rule development. A higher ASME service level was

considered for B&W and CE plants but rejected on the basis that the RCS pressure boundary

could deform to the point of inoperability. Also steam generator tubes might fail before other

4
primary coolant system components and bypass containment. Even if the peak pressure is

limited to acceptable levels, the pressure must be reduced to successfully inject borated water

from the high-pressure injection system. The pressure is reduced by removing heat through

the steam generators, the AFW system, and steam generator relief valves. If the pressure

cannot be reduced, reactor coolant will be lost through the primary safety and relief valves and

the core uncovered. W plants generally have a larger relief capacity than the B&W and CE

plants and are more tolerant of an ATWS. For PWRs, it is likely that for an unmitigated ATWS,

the core would melt with the containment intact.

The MTC becomes more negative (less positive) later in the fuel cycle; and the MTC is more

negative (less positive) at 100 percent power than at lower power. During the first part of the

fuel cycle below 100 percent power, the MTC can be positive or insufficiently negative. If an

ATWS occurs when the MTC is either positive or insufficiently negative to limit reactor power

and the ATWS pressure increase, all subsequent mitigative functions are likely to be ineffective.

The period of the fuel cycle time when the MTC is insufficiently negative to maintain the RCS

pressure below 3200 psig during an ATWS is designated unfavorable MTC.U

Appendix A contains the simplified generic event trees as shown in Figure A-3, ATWS Rule

Event Tree For Westinghouse Reactor Group," and Figure A-4, ATWS Rule Event Tree For

B&W/CE Reactor Group." Appendix A also shows data used to calculate P(ATWS) in each

event tree and summary of the calculation.

At the time of the ATWS rulemaking, the estimated impact of periods of unfavorable MTC" was

included in the event tree as the "MTC Overpressure" factor. For Westinghouse plants, the

original ATWS expected risk approach estimated an MTC Overpressure factor of 0.01 for

turbine trip scenarios and 0.1 for non-turbine trip scenarios; and the expected outcome of the

ATWS rule with installation of AMSAC assumed that Westinghouse plants would be in the

turbine trip scenario 70 percent of the time (i.e., 70 percent of the ATWS transients would be

the lower risk scenario). As stated above, the turbine trip provides the operator additional time

to respond.

For CEIB&W plants, a higher UMTC Overpressure" factor of 0.5 was used regardless of turbine

trip status. The reason was that the MTC was expected to be insufficiently negative to limit

peak pressure below 3200 psig for up to 50 percent of the cycle. For CEIB&W plants, the

expected outcome of the ATWS rule included a factor of 10 improvement in the electrical

portion of the scram system based on installation of the DSS. These same assumptions were

used to determine the current outcomes.

Uke the BWR risk analysis, this risk model does not include low frequency anticipated

operational occurrences such as loss of primary system flow for which core damage occurs,

regardless of the MTC or other factors, unless the rods insert immediately.

2.1.3 Operator Action

In developing the ATWS rule, the NRC staff view was that (1) operator action should not be

relied upon during the first ten minutes of an accident including a manual scram and

(2) operator actions should be relied upon later in the course of an accident If the condition in

the reactor and mitigating systems is available to the operator, that sufficient time is available to

assess the condition and take action, and that the operator has been trained in the action.

5
In practice, operators are trained to identify an ATWS condition, initiate a manual scram,

manually insert the control rods, manually trip the RPS motor-generators, and begin borating.

After the Salem events and before the ATWS rule was issued, there was discussion in

NUREG-1 000, "Generic Implications of ATWS Events at the Salem Nuclear Power Plant",

April 1983, regarding credit for operator action to scram the reactor following an ATWS, since

the Salem operator scrammed the reactor 30 seconds after the RTS failed. NUREG-1000

stated that the Salem ATWS events were not as serious as they could have been as they took

place at low power, the operators were at the correct control panel to support power ascension,

and there was sufficient relief capacity and steam generator mass.

Operator recognition that an ATWS condition exists can be difficult, especially for events

involving malfunctions of the RTS which do not activate the trip indicator lights. The fast

moving ATWS event does not provide much time for the operator to recognize what has

occurred. For example, NUREG-1 000 provides peak pressure versus time curves that show

that if the operator takes no action following an ATWS, the ASME service level C stress would

be exceeded in 47, 70, and 100 seconds for CE, B&W, and W PWRs, respectively. In addition,

the pressure ramps up from a normal operating values of 2250 psia to its peaks of

approximately 3500, 4000, and 3500 psia in 47, 20, and 10 seconds for CE, B&W, and W

PWRs, respectively.

2.2 Commission Recommendations for Reducing the Risk From An ATWS

When the Commission issued the ATWS rule in 1984, it made two recommendations to reduce

the risk from an ATWS: (1) the industry develop a reliability assurance program to achieve and

maintain the desired high level of RTS reliability and (2) licensees reduce the frequency of

challenges to plant safety systems by reducing the initiating event frequency. Each of these

recommendations is discussed in more detail below.

In 49FR26036 the Commission stated that a pivotal aspect of the ATWS issue is the reliability

of the RTS and the difficulty associated with assessing the reliability of a system designed for

very high reliability. The Commission noted that despite perceived high reliability, two precursor

events had occurred.

One of the principal findings from the Salem ATWS event was of lack of attention to RTS

reliability. Accordingly, the staff recommended to the Commission that a reliability assurance

program be included in the final ATWS rule (NUREG-1000, Volume 1, Generic Implications of

the ATWS Events at the Salem Nuclear Power Plant," April 1983 [Ref. 7J). While the ATWS

rule does not require such a program the Commission recommended a RTS reliability program

with the following elements: (1) an analysis of the challenges to and failure modes of the RTS

system, considering independent and common-cause failures; (2) a numerical performance

standard for the RTS challenges and RTS unavailability to aid in the initial and continuing

evaluation of the adequacy of the system; (3) a process for evaluating operating experience to

provide feedback to assess whether the RTS is performing reliably enough; and (4) procedures

within quality assurance programs to ensure that the RTS performs satisfactorily as the

frequency of challenges to the RTS should be as low as practicable.

In 49FR26036 the Commission also observed that licensees did not have a formal program to

reduce the number of automatic scrams and urged licensees to analyze challenges to safety

6
systems to determine where improvements could be made. In a Commission briefing on

June 24, 1984, key industry representatives stated that each utility would adopt practices and

policies to reduce the number of automatic reactor trips, including identifying and correcting the

root cause of all automatic trips; that one representative had agreed to collect, analyze, and

trend industry data; and that licensees of plants operated more than 3 years, would strive to

reduce the average number of automatic trips in 1985 to three. They also stated that the

average number of automatic scrams per reactor had fallen from six in 1980 to four in 1983.

2.3 NRC Regulatory Analysis for the ATWS Rule

SECY-83-293 contains an analytical baseline value of P(ATWS) based on operating data and

assumed expected values of P(ATWS) and other performance expectations. This value credits

installed design features relevant to ATWS or planned modifications to meet regulatory

requirements to carry out initiatives unrelated to the ATWS rule. For example, after the Salem

event the baseline for the RPS reliability for W PWRs was found to be 2E-04 per demand.

However, SECY483-293 assumed a smaller failure-to-scram rate of 3E-05 per demand based

on credit for the reactor trip breaker (RTB) modifications and attention to RTB test and

maintenance as required by GL 83-28, Required Actions Based on Generic Implications of

Salem ATWS Events," July 8,1983 [Ref. 8]. SECY-83-293 shows the potential alternate

modifications for the ATWS rule requirements and calculated a value of P(ATWS) for each

altemative. The difference between the P(ATWS) for each alternative and the baseline values

of P(ATWS) established the risk reductions that were used in the value-impact analysis

(discussed in Section 3.3).

3 ASSESSMENT OF THE ATWS RULE

The goal of this assessment is to determine whether the ATWS rule and the other relevant

Commission recommendations issued with the ATWS rule were effective in achieving the

desired outcomes and whether certain areas may need attention. The assessment reviews and

uses plant specific operating experience, and risk and reliability information to make

conclusions about the adequacy of the regulatory documents; the assessment does not

address plant-specific issues as these continue to be addressed elsewhere in the regulatory

process.

3.1 Method for Assessing Regulatory Effectiveness of the ATWS Rule

For the purposes of this assessment, the regulatory documents present the regulatory

expectations (desired outcomes) as specific objectives, requirements, and guidance. The

regulatory documents are considered effective if the expectations are being achieved.

Expectations were established from objective measures stated in the ATWS rule, the

accompanying 49FR26036, and SECY-83-293 for comparison to the outcomes in the areas of

modifications, risk reduction, and value-impact. The value-impact assessment determined if

the industry's costs to implement the ATWS rule were reasonable.

The risk quantification method used to calculate the expected risk in the original ATWS analysis

was used for this assessment in order to maintain a consistent approach. In so doing, the

improvement in risk that was identified would more obviously be the result of less frequent

7
I

initiating events and improved equipment performance rather than changes in the risk model or

assumptions. This approach provides a reasonable framework for understanding the basic

elements of the ATWS risk - initiating event frequency, scram system reliability, and mitigative

reliability.

The ATWS rule generic risk expectations were calculated assuming that the ATWS rule

requirements were implemented. Thus, the original analysis assumed that AMSAC, DSS, ARI,

RPT trip, etc., were installed and operable.

For PWRs, the original generic risk analysis and this assessment assume that the following

equipment is available: emergency feedwater, primary power-operated relief valves and

safeties, and secondary power-operated relief valves and safeties - this tends toward lower

peak pressures and lower ATWS risk. On the other hand, the original generic risk analysis

assumes that the rod control system is in manual control - this tends toward higher peak

pressures and higher ATWS risk. Since the affects are offsetting, the total impact may be

minimal. Again, to determine whether the ATWS rule was effective in reducing risk from

ATWS, we looked to see if it met or exceeded the goal based on the original method of

calculation.

The original generic risk analysis event trees did not distinguish those initiating events

accompanied by loss of main feedwater. And again, for consistency, neither did this

assessment, since the goal of this assessment was to determine if the ATWS rule resulted in

improved performance, rather than "improved modeling." However, for W plants, the likelihood

of loss of feedwater events was a consideration in adoption of the values of the "MTC

Overpressure" factor and separate event trees for turbine trip and non-turbine trip scenarios.

The original generic event trees were developed after much deliberation to focus on what were

the most important sequences. The deliberations which went into determination of the type of

event trees to use considered aspects of the ATWS transient such as the status of the main

feedwater system. The current assessment was not meant to be a total reevaluation of the

approach to ATWS risk analysis, but an attempt to determine if the ATWS rule goals were met.

In order to make a valid comparison of the improvements due to less frequent initiating events

and improved equipment performance rather than changes in the risk model or assumptions,

the simplified event tree approach was used. Additionally, we believe that this approach

provides a reasonable estimate of ATWS risk.

Operator action to scram the plant was not credited in the original risk estimates of the ATWS

rule. To be consistent, the current assessment used the same approach. The goal of this

assessment was to compare the ATWS rule expectations and outcomes. Consequently, as

long as operator performance has not changed greatly since the ATWS rule was enacted or the

impact of including operator action to scram is small, this approach is valid.

First, regarding changes in operator performance to scram the plant which have

occurred since the ATWS rule was implemented, there is no basis to support major

improvements. At the time of the ATWS rule, operating procedures and training had

already been greatly improved following the Three Mile Island accident; and the Salem

ATWS event implications had been included as previously discussed. The implicit

pressure on operators to minimize the number of scrams may not have been as strong

as it is now.

8
Secondly, recent work supports the position that the impact of considering operator

action is relatively small. "Reliability Study: Westinghouse Reactor Protection System,

1984-1995,0 NUREG/CR-5500 Volume 2, shows values of RPS unavailability of 2.1 E-5

without operator action and 4.9E-6 with operator action - the operator provides reduced

risk by a factor of about 4. This improvement in unavailability assumes that the operator

initiates a manual scram 99 percent of the time when a scram signal is present and

50 percent of the time even when a scram signal is not present. In other words, these

values already assume excellent operator performance. The reliability study did not

include the potential for operator action to manually trip the RPS motor generator sets.

The motor generator breaker controls are not always accessible from the control room,

which would negate the possibility of timely operator action.

Thirdly, the initiating event frequency is assumed to be the rate of automatic scrams.

Since the time of the ATWS rule, the fraction of all scrams which are automatic has

decreased and the fraction of operator initiated scrams has increased. Manual scrams

were not included in the current ATWS initiating event frequency; their inclusion in

initiating event frequency would about double the initiating event frequency.

The outcomes were obtained from operating experience as documented by the NRC

performance indicator program and licensee event reports (LERs); completed and draft NRC

reliability studies for each reactor group; NRC survey of PWR MTCs; and vendor information

presented to the NRC. The NRC 1997-1999 performance indicator data used for counting

automatic scrams are given in Appendix B, *Plant-Specific SBO Information by Reactor Type

and Operating Status." Insights on the initiating event frequency were obtained from

NUREG/CR-5750, Rates of Initiating Events at U.S. Nuclear Power Plants: 1987-1995,0

December 1998 [Ref. 9]. NRC RPS reliability studies completed based on operating

experience since the ATWS rule are: NUREG/CR-5500, Volume 2, "Reliability Study: General

Electric Reactor Protection System 1984-1995," December 1998 Ref. 10]; NUREG/CR-5500,

Volume 3, Westinghouse Reactor Protection System 1984-1995," December 1998 Ref. 11 , the initial draft of NUREG/CR-5500, Volume 4, " Reliability Study: Babcock & Wilcox Reactor

Protection System 1984-19980, March 2000 [Ref. 12]; and the initial draft of NUREG/CR-5500

Volume 10, Reliability Study: Combustion Engineering Reactor Protection System

1984-1998, March 2000 [Ref. 13]. The RPS reliability data were obtained from NRC studies

that model the components of the RPS system as recommended by the Commission in issuing

the ATWS rule. The AFW system reliability was obtained from NUREG/CR-5500, Volume 1,

"Reliability Study' Auxiliary/Emergency Feedwater Study System, 1987-1995," August 1998

[Ref. 14]. Information on the plant and reactor group MTCs was obtained from a 1994 NRC

survey [Ref. 15], and owners' group presentations to the NRC [Ref. 16], [Ref. 17], Ref. 18].

Appendix B gives plant-specific data on the actual outcomes of the ATWS rule regarding

modifications and data from probabilistic risk assessment/individual plant examinations

(PRA/IPEs). The data were collected from NRC licensee correspondence, particularly licensee

correspondence committing to modifications in response to the ATWS rule, and from licensee

PRA/IPEs dated from November 30, 1991, to July 27, 1994, as recorded in the NRC PRA/IPE

databases. NUREG-1560, "Individual Plant Examination Program: Perspectives on Reactor

Safety and Plant Performance," October 1997 [Ref. 19], was used to obtain ATWS insights

based on licensee PRA/IPEs. When using these PRA/IPEs, it was recognized that their data

did not always reflect the current design or operating performance of safety systems.

9
iN

3.2 Comparison of Expectations and Outcomes

Table 1, Summary of ATWS Rule Expectations and Outcomes," compares the ATWS rule

expectations and outcomes in the areas of system modifications and operating limitations, risk

reduction, and value-impact. The expectations and outcomes are discussed in the sections that

follow.

Table 1 Summary of ATWS Rule Expectations and Outcomes

ATWS Rule Expectations

Area Expected Result Current Outcomes Observations

System W: AMSAC, unfavorable MTC All required modifications Expectation met

Modifications 1-10% implemented; trend to

and Operating CE: AMSAC, DSS. unfavorable increase technical

Limitations MTC <50% specification (TS) MTC limits

B&W: AMSAC, DSSunfavorable

MTc<50%

GE: ARI, RPT, and SLC

capacity; some automatic

SLC initiation

Risk Average ATWS risk < .E-05 Al plants < 1.0E-05 Expectation met but large

Expectations uncertainties in RPS reliability

W: 5.8E-06 W: 5.OE-07

cE: 2.2E-05 CE. 2.8E-07 ATWS mitigating capability

B&W: 2.2E-05 B&W: 1.3E-07 sensitive to unfavorable MTC for

GE: 1.22-05 GE. 1.0E-06 PWRs and HEP for BWRs.

Value-Impact Industry cost to implement the ATWS Industry cost to implement Expectations met

rule. $354M the ATWS rule: 166M

Significant economic benefit from

reduction of the number of scrams

, 

3.2.1 Modifications and Operating Limitations

Appendix B, Table B-1,' Operating Pressurized-Water Reactor Data," and Table B-2,

"Operating Boiling-Water Reactor Data," summarize the modifications licensees committed to

install in response to the ATWS rule. A review of licensing correspondence found that the

expected modifications from the ATWS rule were installed, typically between 1986 and 1990.

Table 2, ATWS Rule Modifications," was prepared to show the degree of defense in depth

provided by the ATWS rule modifications that were intended to prevent an ATWS; to obtain a

decrease in potential common-cause failure (CCF) in the RTS; and for the B&W and CE

reactors to ensure a reactor trip in view of relatively high Unfavorable MTCs. To assess the

degree of defense in depth, Table 2 shows the number of total RPS system failures; RPS trip

system design redundancy; the modification expectations (and for PWRs, the Unfavorable

MTC); and a summary statement of the outcome.

10
Table 2 indicates:

(1) The GE ARI with the RPS trip system redundancy appear to provide additional measure to

ensure prevention of an ATWS and or ensure a reactor trip from a CCF in the RTS.

(2) The B&W and CE RPS trip system design redundancy and the DSS appear to provide a

compensation for the higher unfavorable MTC (up to 50 percent), provide diverse means to

prevent an ATWS, and provide protection against a CCF in the RTS. In comparison to the

B&W and CE plants, W plants have no need for these compensatory measures because of

unfavorable MTC limited to 1 percent. However, 49FR26036 states the requirement for a DSS

for W plants would be published separately to provide opportunity for public comments

regarding marginal value-impact, accident prevention, and common mode failures. On

December 3, 1984 the NRC determined the DSS was not required for W plants.

Table 2 ATWS Rule Modifications

Reactor Protection System Expectations

Current Outcomes

RPS RPS Trip System Mods PWR Fuel

Failures Strategy

GE 1 all primary and backup trip system, ARI, RPT, Modifications installed

plants one-out-of-twa-twice logic SLC

W 2 all 14 RTB, shunt and undervoltage AMSAC unfavorable Modifications Installed

plants trips MTC 1-10%

B&W 0 6 plants 1/4 RTB (2 ac, 2 do), shunt and AMSAC. unfavorable Modifications installed

undervoltage trips DSS MTC<50%

SCRs provide diverse trip of

regulating rods

1 plant 1/4 RTB, shunt and undervoltage

trips,

SCRs provide diverse trip of

safety rods

CE 0 2 plants 1/4 contactors AMSAC, unfavorable Modifications hIstalled

DSS MTC<50%

9 plants 1/8 RTB, shunt undervoltage trips

3 plants 1/4RTB (2 W and 2 GE) with

automated shunt and

undervoltage

3.2.2 Comparison of ATWS Rule Risk Expectations and Outcomes

SECY-83-293 set the goal that P(ATWS) be no more than 1.0E-05IRY. SECY-83-293 did not

treat this goal as a requirement as this goal might be too costly for some plants to achieve.

SECY-83-293 established more reasonable risk expectations for each reactor group in terms of

P(ATWS), using estimates for the (1) the initiating event frequency, (2) the reliability of the

RPS, and (3) the reliability of ATWS mitigation systems. The SECY-83-293 expectations in

these three areas that were used to estimate P(ATWS) are given in Table 3, Summary of

ATWS Rule Risk Expectations and Outcomes," under the Expected column for each reactor

11
-1

group. Under the Outcomes column, Table 3 also lists the corresponding outcomes for each

reactor group based on current operating experience data gathered since the ATWS rule was

issued as summarized in Appendix A. In a few cases noted in Appendix A, there was no readily

available operating experience and the expectations were used to estimate the P(ATWS)

outcome. The Appendix A event trees provide a consistent comparison of the regulatory risk

expectations to the outcomes.

Table 3 Summary of ATWS Rule Risk Expectations and Outcomes

Risk Factor General Electric Westinghouse Babcock & Wilcox Combustion Engineering

Means

Expected Outcome Expected Outcome Expected Outcome Expected Outcome

Initiating Event 4.3 0.5 4.0 0.5 4.0 0.5 4.0 0.5

Frequency (RY)

RPS Unreliability 1 .2E-05 5.8E-06 3.0E-05 2.1 E-05 1 .2E-05 5.0E-07 1 .2E-05 1.1 E-06

Mitigation 2.3E-01 3.6E-01 4.8E-02 4.7E-02 4.6E-01 5.1 E-01 4.6E-01 5.1 E-01

Unreliability I I

P(ATWS) 1.2E-05 1.OE-06 5.8E-06 5.OE-07 2.2 E-05 1.3E-07 2.2E-05 2.8E-07

(unmitigated)

Table 3 indicates that P(ATWS) for each reactor group is < 1 .OE-O5/reactor-year (RY); the

industry has met the SECY-83-293 risk goal. Comparison of Table 3 P(ATWS) expectations to

outcomes shows that the P(ATWS) outcomes are better than expected by factors of

approximately 10, 10, 170, and 80 for the GE, W, B&W, and CE reactor groups, respectively.

To elaborate on the reductions in P(ATWS):

(1) The initiating event frequency has been reduced by a factor of 8 demonstrating that the

Commission's recommendation to reduce the number of automatic reactor scrams has

been very effective in reducing P(ATWS). The NRC performance indicator data in

Appendix A was used to develop the initiating event frequency which shows that in

1997, 1998, and 1999, respectively, 70, 67, and 59 reactors, had zero automatic

scrams. However the same data indicate that in 1997, 1998, and 1999, respectively

indicates 9, 10, and 17 reactors, respectively, had automatic scram rates of 2-4/RY so

outliers could increase plant specific ATWS risk.

(2) The data used to develop the RPS reliability indicates there have been no total system

failures of the RPS for any reactor group since issuing the ATWS rule. The RPS

unreliabilities are by far the smallest P(ATWS) factor (on the order E-04 to E-06)

indicating that RPS system failures should be very rare events. Table 3 values indicate

that the all reactor groups achieved better than expected improvements in RPS

reliability. These numbers were developed using a fault tree model of the RPS system

that may not include all failure modes, a question of completeness for all PRA

calculations. For comparison, alternate RPS reliability expectations and outcomes

12
shown in Table 4, Altemate Estimates of Reactor Protection System Reliability," were

developed using other techniques as discussed in Section 3.2.3.

Appendix B, Table B-3, Reactor Protection System Unreliability Uncertainties,"

summarizes the uncertainties that accompany the RPS unreliabilities in Table 3. The

RPS unreliabilities in Table B-3 for the B&W and CE reactors do not include the effect of

the DSS on RPS electrical unreliability. Table B-3 shows that the outcomes of upper

and lower bounds of unreliability range from 1.8-06 to 5.7E-05 and this range is smaller

than that noted Section 2.0 before the ATWS rule was issued. Consideration of the

lower values of RPS unreliability in Table B-3 would result in smaller values of

P(ATWS), while consideration of the upper values in Table B-3 would increase

P(ATWS) for each reactor type by a factor of approximately 2.5. Table B-3 also shows

the upper bound of uncertainty analysis associated with current estimates are 5.7E-05,

1.4E-05, 2.5E-05, and 0.78E-05 for the W. GE, CE and B&W reactors, respectively2.

(3) The reliability outcomes for the mitigation systems were about as expected and did not

heavily influence the risk reduction2. From this, it could be deduced that the mitigation

functions described in the ATWS rule (ARI and RPT on BWRs; AMSAC on PWRs)

achieved less risk reduction than the Commission's recommendations in 49FR26036 to

reduce the rate of reactor scrams.

As previously discussed, ATWS mitigation on a BWR is highly dependent on operator

actions. This is discussed further in Section 3.2.4 below

For the W plant mitigation systems unreliability, the value (-5E-2) is largely determined

by the unfavorable MTC. Increasing MTC can have a major adverse impact on the

unfavorable MTC. This trend is discussed further in Section 3.2.6.

3.2.3 Reactor Protection System Reliability Validity

Table 4 was developed to show how reliability estimates vary depending on the type of analysis.

SECY-83-293 baselinen estimates of RPS unreliabilities were obtained by using classical

statistics, modeling the RPS as a component, estimating the number of RPS demands from

periodic plant test and operating experience at all operating U.S. nuclear plants from initial date

of commercial operation through 1983, and failures which include Salem and Browns Ferry.

SECY-83-293 stated that counting RPS system failures and system demands was the best way

to account for CCF effects.

Table 4 also shows the corresponding updated calculation using the SECY-83-293

methodology and data plus the data since the ATWS rule was issued (1984-1995) from NRC

RPS reliability studies (references 11 and 12). These are based on the same failures (Salem

and Brown's Ferry) but a much larger number of demands.

For comparison, Table 4 shows the results of NRC RPS reliability studies. The NRC RPS

reliability studies results were obtained by modeling of the RPS components (as recommended

2 The DSS Is not relevant to the discussion here - the RPS reliability values for CE and B&W do not nclude the

contribution of the DSS as the DSS was not modeled in the RPS reliability studies.

13
by the Commission), operating data since 1984 (scram failures at Browns Ferry in 1980 and

Salem in 1983 not included), and Bayesian statistics.

Comparisons of the W and GE reliability expectation and outcomes calculated by the methods

used to establish the "baseline" values of SECY-83-293 show only modest improvements in

scram system reliability. This is expected since the calculation is based largely in the number

of demands - that number increased by a factor of about 2. These results are close to the

lower bound reliability numbers in the NRC reliability studies.

The range of values of RPS reliability illustrates the difficulty of estimating reliability values in

highly reliable systems. The significance is that the uncertainty of the values of RPS reliability

argue for maintaining defense in depth regarding ATWS.

Table 4 Alternate Estimates of Reactor Protection System Reliability

Assumptions and Data W RPS BWR RPS

Unreliability Unreflability

SECY-83-293 baseline - classical statistics 20E-05 19E-05

RPS modeled as a component; data prior to ATWS rule

W PWR - one RPS failure (Salem) in 4975 demands

BWR - one RPS failure (Brown's Ferry) In 5258 demands

SECY-83-293 expectation - estimated improvement 3E-05 1 .2E-05

Assumed improvement in the baseline RPS unreliability

SECY-83-293 baseline update to 1995 - classical statistics 9.8E-05 12E-05

RPS modeled as a component; all data to 1995

W PWR - one RIPS failure (Salem) In 10182 demands

BWR - one RPS failure (Brown's Ferry) in 8119 demands

NUREG/CR-5500 (Ref 12) - BayesIan statistics with non-infornative prior 2.1 E-05 0.58E-05

RIPS modeled to component level; component failure data from

1984 to 1995 (Salem and Brown's Ferry events not included);

demands estimated based on number of tests and unplanned

demands

3.2.4 Risk Insights From Ucensee Probabilistic Risk Assessmentindividual Plant

Examinations

NUREG-1 560 provides common ATWS risk perspectives for each reactor group gained from

the NRC staff review of 75 of the IPEs submitted to the NRC for 108 nuclear power plants.

NUREG-1 560 noted that licensee IPEs show that ATWS was relatively unimportant from a risk

perspective regardless of the reactor group. However, in some cases the ATWS contribution to

core melt was more than 10 percent of the total. NUREG-1560 qualified its conclusion about

ATWS CDF, noting the variability in the PRA/IPE modeling of ATWS events for both BWRs and

PWRs.

Comparison of the ATWS CDF for all the plants in each reactor group in Appendix B to the

P(ATWS) expectation for each reactor group found that six licensees do not meet the ATWS

risk goals. Five of those licensees assumed RPS reliability levels lower than industry average

and in the lower range of the RPS reliability uncertainties discussed in Section 3.2.2. One PWR

14
licensee assumed an inoperable power-operated relief valve; relief capacity impacts ATWS

peak RCS pressure and consequently ATWS risk. Several plants routinely operate with

blocked power-operated relief valves and their IPEs may underestimate ATWS risk.

PRAIPEs for BWRs, discussed in NUREG-1560, indicate large variations in the assumptions

about the reliability of human actions in response to an ATWS. NUREG-1560 sampled 33

plants and found the HEP for SLC initiation ranged from 0.0001 to 0.5. NUREG-1560 also

sampled 25 plants and found the HEP for automatic depressurization system inhibition ranged

from 0.00001 to 0.5. Usually, a low HEP is associated with low stress events. Similarities in

BWR design, procedures, and training would seem to indicate that more consistent HEP

assumptions should be used in the IPE analyses.

Operating experience supports the view that ATWS is a high stress event. A review of LERs in

the last 10 years found only one instance where the ATWS EOPs had been entered. This was

the focus of a special inspection documented in NUREG-1455, Transformer Failure and

Common-Mode Loss of Instrument Power at Nine Mile Point Unit 2 on August 13, 1994,"

October 1991 [Ref. 20]. This event was due to the loss of control rod position indication so the

operator was unable to verify that the reactor had shut down. Although the rods were actually

inserted and the human error was of no consequence, NUREG-1455 indicated that in early

steps of the ATWS EOP to control injection of coolant, the operators did not anticipate that

depressurizing the reactor would cause the condensate pumps to inject cold water, even

though the EOP cautioned that injecting water would induce a power excursion that would could

lead to core damage. This event illustrates that BWR EOP implementation is challenging for

the operators in an ATWS situation.

ATWS mitigation on a BWR is highly dependent on operator actions. Although improvements

in design, procedures, and training since the ATWS rule was issued should have contributed to

improved mitigative response capability, BWR operator response to an ATWS continues to be a

challenge. PRAMIPEs for BWRs indicate large variations in the assumptions for reliability of

human actions in response to an ATWS. Similarities in design, procedures, and training argue

against such variability.

3.2.5 Insights From NRC Reliability Studies and Operating Experience

NUREG/CR-5750 analyzed the data for all unexpected reactor trips and revealed that most risk

significant initiating frequencies with respect to ATWS have decreased over time. Those

events include: total loss of feedwater; loss of instrument or control air; inadvertent closure of

MSIVs; and loss of the BWR condenser heat sink. Some observations based on the NRC RPS

reliability studies (References 11, 12, 13, and 14) and operating experience are as follows:

NRC RPS reliability studies have identified potential important CCFs. Appendix B, Tables B-4

through B-7 list CCF data from the NRC RPS reliability studies. They identify the RPS

components of greatest importance to RPS reliability and risk for each reactor group.

Tables B-4 through B-7 show the components of greatest importance, as indicated by the

Fusel-Vesely importance measure, the risk ratio increase, and the probability of failure. These

tables show that the important common-mode failures, though of low probability, are the train

bistables, trip logic relays, undervoltage cards, RTB, and rod and rod mechanisms. In addition,

the PWR channel process module and the BWR solenoid-operated valve scram pilot values are

also important.

15
Review of recent operating experience in areas found to be dominant CCFs found the industry

is still addressing RTB maintenance/reliability issues that were identified from the Salem ATWS

events, and the phenomena related to control rod insertion was not considered during

development of the ATWS rule.

NRC Information Notice (IN) 99-13, lnsights From NRC Inspection of Low-and MediumVoltage Circuit Breaker Maintenance Programs," April 29, 1999 [Ref. 21], summarizes NRC

inspections of licensee circuit breaker maintenance programs at eight nuclear plant sites in

1998. The inspections followed NRC Temporary Inspection Procedure TI 2515/137, nspection

of Medium-voltage and Low-voltage Power Circuit Breakers," Revision 1, March 9, 1998

[Ref. 22]. TI 2515167 lists more than 60 INs notifying the industry of circuit breaker problems

including INs that address low voltage power circuit breakers of the type used in RPS circuit

breaker applications. Although the inspections concluded that the programs were generally

adequate, the inspections found some licensees' circuit breaker maintenance programs may

need attention. The inspectors reported these concerns to the industry. It appears the industry

is resolving these problems through Electric Power Research Institute/Nuclear Maintenance

Assistance Center workshops, an Nuclear Energy Institute circuit breaker task force, and circuit

breaker users groups all of which have helped plants to fix their circuit breaker problems.

However, information from recent circuit breaker users group meetings, indicates that not all the

plants have yet initiated effective circuit breaker maintenance programs, as advocated by NRC

and the industry groups.

The higher fuel bumup has resulted in previously unpredicted oxide growth and fuel assembly

distortion. In some cases this has resulted in slow or incomplete control rod insertion. Table 5,

Control Rod Insertion Events," summarizes recent events and NRC INs involving control rod

insertion. While the number of control rods affected is only a portion of the total available rods

actually needed to obtain reactor control and shutdown, these failures and degradations are

new phenomena and were not considered during the development of the ATWS rule. Although

these conditions do not affect present ATWS analysis assumptions directly, they cannot be

dismissed as precursor events.

3.2.6 Changes in Fuel Management May Affect Pressurized-Water Reactor ATWS Mitigating

Capability

Since the ATWS rule was developed, fuel cycle lengths have been extended. This requires

higher enrichment fuel and often results in less negative (more positive) MTCs. As previously

discussed in Section 2.1.2, the MTC strongly influences the peak RCS pressure and the ability

to mitigate an ATWS; the less negative (more positive) the MTC, the higher the pressure peak.

Pressures for economic efficiency are also prompting licensees to increase the power of the

reactor. The higher power rating leads to faster dry-out of the steam generator for PWRs and

faster heatup of the suppression pool for BWRs. The combination of less negative MTC and

higher reactor power has a greater effect on the plant response to ATWS than either one alone.

Less negative MTC and higher power result in the ATWS pressure spike occurring earlier,

allowing less time for operator actions, quicker, less time for MTC to limit the pressure increase,

and higher, increasing the likelihood of equipment damage. These fuel management changes

potentially impact both the deterministic and probabilistic analyses for ATWS and other reactor

transients.

16
Table 5 Control Rod Insertion Events

Reference Description of Event

LER 289/97-008 On 7/21/97an LER after a scram found that 8 of 61 four control rods exhibited slower

than normal scram times (4 of 61 were not within TS limits) because of a hydraulically

Three Mile Island induced effect from reduced clearances In the thermal barniers because of deposits on

the internal check valves and between the thermal barrier parts. The LER noted that

plants conditions during the event are different than during control rod trip insertion time

testing performed at hot shutdown. In LERS 95-002, 94-004, and 94-002, the licensee

also reported excessive control rod drops times.

Bulletin 96-01, 'Control Rod After three licensees reported that control rods In high-bumup fuel assemblies had

Insertion Problems' insertion times greater than expected, licensees were requested to take prescribed

(1218/96) actions to ensure the required shutdown margins are maintained during reactor trip.

IN 96-12, Control Rod Insertion Three licensees reported that control rods In high bumup fuel assemblies had insertion

Problems' times greater than expected.

(211896)

IN 94-40.Supplenent 1, *Failure of This informed licensees that after five nuclear plant licensees found that loosened pins

a Rod Control Cluster Assembly to have caused control rod(s) to jam, W recommended an inspection at the next outage.

Fully Insert Following a Reactor Trip

at Braidwood Unit 2. (12/15/94)

IN 94-72, Increased Control Rod Two licensees reported increased rod drop times, because cnd deposits caused the

Droptime From Crud Buildup' thermal barrier ball check valves to stick and reduced clearances In the thermal barrier

(10/5/94) bushing.

Table 6, UATWS Moderator Temperature Coefficient and Peak Pressure for Pressurized-Water

Reactors," summarizes the MTC and corresponding limiting peak pressures of 1979 and 1988

ATWS analyses by the PWR manufacturers. The units of reactivity in Table 6 are pcml/F

(E-05 &K/K0F).

Table 6 ATWS Moderator Temperature Coefficient and

Peak Pressure for Pressurized-Water Reactors

Parameters 1979 198 1994 NRC Technical Specification

ATWS Analysis Update of ATWS Analysis Survey

CE -TC -2.0 to 4.8 -2.6 to -5.7 0 to 43 above 70% power

Peak 4290 psla 4153 psla

Pressure

B&W MTC -10.5 18 month cycle: -11.0 0 above 95% power

24 month cycle: 4.3

Peak 3464 psla 3764psia

Pressure 18 month cycle: > 3200

24 month cycle: > 3200

W MTC -8.0 -8.0 average Unearto 0 from 70% to 100% power

_______________ range -5 to -15 One plant at + 2 at 100% power

Peak 3197 psa 3497 psla (-5pcIn/F) Pressure J

17
In 1987, the NRC staff requested the PWR owners groups to quantitatively reassess the initial

MTC analyses that support the ATWS position for PWRs and specifically address reload core

designs that have less conservative initial MTCs (a typical letter is given in [Ref. 231). Table 6

summarizes the MTCs and peak pressure obtained from the owners group responses

[Refs. 16, 17, and 18]. The responses generally concluded that no significant changes were

required in ATWS analyses. The peak pressure of 3497 psia update in Table 6 for W was

calculated based on values obtained from sensitivity analysis in NUREG-460. The sensitivity

analysis shows a pressure increase of approximately 100 psi for a change in MTC of 1 pcm/ F.

Thus, the change from the 1979 value of -8 pcml F to -5pcm/ F in Table 6 corresponds to a

300 psi increase.

The plant TSs limit the MTC value and require MTC surveillance. Table 6 also summarizes the

range of PWR TS MTC limits from a 1994 NRC survey [Ref. 161. The survey results are listed

in Appendix B. Table 6 indicates that the 1994 PWR TSs limit the MTCs to positive or zero

levels at full power. These are less negative (more positive) than the 1979 and 1988 MTC

values. Based on NUREG-460 sensitivity analysis that indicates a 1 pcm/ F less negative

(more positive) MTC increases the RCS peak pressure approximately 100 psi calculations

based on the limiting TS MTC at full power could lead to higher peak ATWS pressures and

longer unfavorable MTCs. CE and B&W reactors installed the DSS to counteract risk and peak

pressure effects of unfavorable MTC. The B&W and CE DSS the reactor trips independent of

the RPS, improving RPS electrical reliability by a factor of 10. The DSS trips the reactor at

approximately 2450 psia to prevent the RCS pressure following an ATWS from reaching the

unacceptable condition-3200 psia ASME service level C limit.

W WCAP-1 1992, "Joint Westinghouse Owners Group/Westinghouse Program: Anticipated

Transient Without Scram (ATWS) Rule Administration Process," was formally submitted to the

staff in May, 1995. WCAP-1 1992 introduced the concept of an unfavorable exposure time

(UET) which is the percentage of the fuel cycle time the pressure is in excess of 3200 psia; in

this respect UET is fundamentally the same as the unfavorable MTC percentage used in the

ATWS rule development. However the UET is calculated differently than the unfavorable MTC

percentage and in this respect they are different. WCAP-1 1992 provides a risk based approach

to justify UETs up to 38 percent for 18 and 24 month fuel cycles. The NRC staff review

[Ref. 24] did not find WCAP-1 1992 acceptable for use in licensing or other regulatory matters

based on several issues. The Westinghouse Owners Group has been working with the NRC

staff to address those issues.

Higher peak pressures and longer unfavorable MTCs or UETs lessen the effectiveness of the

mitigative functions required by the ATWS rule. If the W assessment were done using the

same method as B&W and CE with an unfavorable MTC" changing from 1 percent to

40 percent of the fuel cycle, the ATWS risk would increase by a factor of about 20 to 4.3E-6.

CE and B&W reactors installed the DSS to counteract the risk and peak pressure effects of

unfavorable MTC. The effectiveness of the ATWS rule would be compromised if unfavorable

MTCs or UETs are increased. In particular, large percentage increases in UET being

considered by W plants, without compensating DSS or mitigative capability, challenge the intent

of the ATWS rule.

18
3.3 Value-impact

SECY-83-293, Appendix C, Regulatory Analysis for Amendments Related To ATWS," provides

a generic value-impact analysis for each reactor group. The value-impact analysis ranked the

ATWS rule alternatives for each reactor group by value-impact ratio, the highest value-impact

ratio being the favored alternative. Appendix B, Table B-8, "Value Impact Data for Each

Reactor Group," summarizes the industry value-impact baseline information for each reactor

group in SECY-83-293. SECY-83-293 determined that the total impact on the industry was

$525 million (131 operating plants).

Table 7, "ATWS Rule Value-Impact Summary," is an update of similar value-impact calculations

based on current data. Table 7 shows the calculations of expectations and outcomes

normalized for 102 plants so the expected values are not the same as those in SECY-83-293.

A comparison of the estimated value-impact expectations in the original ATWS regulatory

analysis to the corresponding current outcome indicates that the costs to implement the ATWS

rule were approximately $166 million, which is over 50 percent less than the expected

$354 million. This difference is largely due to the lower monetized value of releases (in keeping

with the reduced probability of ATWS events) and fewer than expected spurious scrams caused

by ATWS mitigation equipment.

The SECY 83-293 value-impact analysis did not consider the effects of the industry

commitment to reduce the number of scrams in response the Commission's recommendation in

49FR 26036. The reduction in the number of automatic scrams accounts for a factor of

8 reduction in P(ATWS) and corresponding significant portion of the outcome value. In

addition, although not included in the this value-impact analysis, the 3.5 reduction in the

average number of automatic scrams for 102 operating reactors has significant monetary value

due to replacement power costs.

Table 7 ATWS Rule Value-impact Summary

Value-Impact Factors Expectation Outcome

Value:

102 x (P(ATWS) expected - P(A1WS) basellne)(30 years)($10 billion) $1238M

102 x (P(ATWS) outcome - P(ATWS) baseline)(30 years)($1 0 billion) $1521M

Impact

Design, installation, operation and maintenance $142M $142M

Replacement power from plant trip due to spurious actuation of ATWS $212M $ 24M

hardware costs $500,000/day, for 2 to 6 days

Total $354M $166M

In summary, compared to the expectations when the ATWS rule was issued, the value

(savings) was greater than expected and the impact (cost) was less than expected. Thus from

a value-impact perspective, the ATWS rule was effective.

19
4 CONCLUSIONS

The assessment concludes that the ATWS rule has been effective in installing modifications,

reducing ATWS risk, and implementing the rule at reasonable cost. However, uncertainties in

RPS reliability and mitigative capability warrant continued attention consistent with the NRC

performance goals to maintain the expected levels of safety and to improve effectiveness. To

elaborate:

* Hardware modifications and operating limitations required by the ATWS rule were

implemented. All PWRs installed diverse means to trip the turbine and initiate AFW.

CE and B&W plants installed a diverse scram system. Westinghouse plants generally

maintain an unfavorable MTC of one percent. BWR plants implemented diverse RPT,

alternate rod insertion circuitry, and upgraded EOPs; or installed high capacity standby

liquid control systems (SLCS).

* The mean frequency of automatic scrams (initiating events for ATWS) decreased from

approximately 4/RYs in 1983 to 0.5/RYs since 1997. This alone accounts for a

reduction of nearly one order of magnitude in the frequency of an ATWS - P(ATWS).

* RPS reliability dominates the risk from an ATWS. There have been no total failures of

the RPS system since the ATWS rule was issued in 1984. Point estimates of RPS

reliability, based on operating experience since 1984 show that the mean RPS

unreliability (one minus RPS reliability) expectations have been met for all four reactor

groups and are approximately an order of magnitude better than the RPS reliability

estimates before the ATWS rule. These generic estimates were developed using a

PRA model for the RPS and failure rates of components.

* PWR scram system reliability is related to RTP reliability. As evidenced by NRC generic

communications and industry group activities, circuit breaker problems continue to

occur. Industry programs to maintain scram system reliability continue to be useful in

limiting risk from ATWS.

* During the ATWS rulemaking the NRC staff set a goal that P(ATWS) should be no more

than 1.0E-05/RY. P(ATWS) was defined as the annual frequency of an ATWS leading

to plant conditions that exceed certain design parameters that can result in core melt,

containment failure, and the release of radioactivity and can be viewed as the expected

CDF of an unmitigated ATWS. Updating the original generic ATWS regulatory analysis,

using operating data since the ATWS rule was implemented, found that on a generic

basis, all four reactor types achieved the ATWS rule risk goal. A few individual plants

had somewhat higher risk as a result of using lower levels of RPS reliability rather than

the industry average or assumed additional equipment failures.

* Comparison of the estimated value-impact expectations in the original ATWS regulatory

analysis to the corresponding outcome indicates that the costs to implement the ATWS

rule were less than expected. This is largely due to fewer than expected spurious

scrams caused by ATWS equipment than assumed in the original analysis.

20
* The higher fuel bumup has resulted in previously unpredicted oxide growth and fuel

assembly distortion. In some cases this has resulted in slow or incomplete control rod

insertion. These failures and degradations are new phenomena and were not

considered during the development of the ATWS rule.

Although past data indicates that the risk from ATWS is in the range foreseen when the ATWS

rule was issued, several issues have the potential to erode past achievements. Attention to

these issues and regulatory actions that maintain compliance with current regulations can

assure that the risk from ATWS remains acceptable. These issues are:

* RPS reliability estimates are subject to large uncertainties. Current point estimates

developed using RPS PRA models show upper and lower bounds of unreliability ranging

from 1 .8E-6 to 5.7E-5. RPS reliability requirements are so high and ATWS events are

so rare that many more years of operating experience are needed to generate sufficient

system demands to reduce current estimates of the uncertainty. Licensee's risk

calculations in support of licensing actions that could affect ATWS risk should address

these uncertainties.

* ATWS mitigation capability on a PWR is highly dependent on the MTC. Mitigative

functions are considered by the ATWS rule regulatory basis to be non-viable if the

ATWS peak pressure exceeds 3200 psig; and a sufficiently negative MTC will limit the

ATWS peak pressure. Fuel design to achieve longer cycles and higher power ratings

may result in less negative MTCs at full power for a larger fraction of the cycle time,

during which time ATWS mitigation may be less effective. CE and B&W reactors

installed a diverse scram system to compensate for large exposure times. Further fuel

cycle changes and power upgrades that could affect the ATWS risk may require

compensatory measures (e.g., hardware or procedural), consistent with the underlying

regulatory basis behind the ATWS rule.

* ATWS mitigation on a BWR is highly dependent on operator actions. Although

improvements in design, procedures, and training since the ATWS rule was issued

should have contributed to improved mitigative response capability, BWR operator

response to an ATWS continues to be a challenge. PRA/IPE for BWRs indicate large

variations in the assumptions for reliability of human actions in response to an ATWS.

Similarities in design, procedures, and training argue against such variability.

Licensee's risk analyses in support of licensing actions should be supported by technical

justification of operator performance assumptions.

5 REFERENCES

I. U.S. Nuclear Regulatory Commission, Response to Staff Requirements Memorandum

of May 28,1997, Conceming Briefing on IPE Insight Report," SECY-97-180, August 6,

1997.

2. U.S. Code of Federal Regulations, Title 10, Part 50, Requirements for reduction of risk

from anticipated transients without scram ATWS) events for light-water-cooled nuclear

power plant,' (Federal Register, Volume 49, No. 124, page 26036, June 26, 1984).

21
3. U.S. Nuclear Regulatory Commission, "Anticipated Transient Without Scram for Light

Water Reactors," NUREG-460, Volume 1, April 1978.

4. U.S. Nuclear Regulatory Commission, "Amendments To 10 CFR 50 Related to

Anticipated Transients Without Scram (ATWS) Events," SECY-83-293, July 19,1983.

5. U.S. Nuclear Regulatory Commission, "Suppression pool Temperature Limits for BWR

Containments," NUREG-0783, November 1981.

6. Sandia National Laboratories, "Handbook of Human Reliability With Emphasis on

Nuclear Power Plant Applications," NUREGICR-1 273, August 1983.

7. U.S. Nuclear Regulatory Commission, "Generic Implications of the ATWS Events at the

Salem Nuclear Power Plant," NUREG-1000, Volume 1, April 1983.

8. U.S. Nuclear Regulatory Commission, "Required Actions Based on Generic Implications

of Salem ATWS Events," Generic Letter 83-28, July 8, 1983.

9. U.S. Nuclear Regulatory Commission, "Rates of Initiating Events at U.S. Nuclear Power

Plants: 1987-1995," NUREG/CR-5750, December 1998.

10. U.S. Nuclear Regulatory Commission, "Reliability Study: General Electric Reactor

Protection System 1984-1995," NUREGICR-5500, Volume 2, December 1998.

11. U.S. Nuclear Regulatory Commission, "Reliability Study: Westinghouse Reactor

Protection System 1984-1995," NUREGICR-5500, Volume 3, December 1998.

12. U.S. Nuclear Regulatory Commission, Reliability Study: Babcock & Wilcox Reactor

Protection System 1984-1998," initial draft of NUREG/CR-5500, Volume 4, December

1998.

13. U.S. Nuclear Regulatory Commission, "Reliability Study: Combustion Engineering Reactor Protection System 1984-1998," initial draft of NUREG/CR-5500 Volume 10,

March 2000.

14. U.S. Nuclear Regulatory Commission, "Reliability Study: Auxiliary/Emergency

Feedwater System, 1987-1995," NUREG/CR-5500 Volume 1, August 1998.

15. U.S. Nuclear Regulatory Commission, note from G. Holahan to J. Blaha, "Requested Material," May 13, 1997.

16. Babcock & Wilcox Owners Group ATWS Committee, Effects of Plant and Fuel

Changes on ATWS Basis Presentation to the NRC," February 18, 1988.

17. Combustion Engineering Owners Group, "Combustion Engineering Owners Group

Meeting With the NRC Concerning Trends in MTCs," January 11, 1988.

22
18. Westinghouse Owners Group, "Westinghouse program ATWS Rule Administration,

Combined Core Performance/Scram Systems Reliability, ACRS Subcommittee

Meeting," February 19, 1988.

19. U.S. Nuclear Regulatory Commission, Individual Plant Examination Program:

Perspectives on Reactor Safety and Plant Performance," NUREG-1560, October 1997.

20. U.S. Nuclear Regulatory Commission, "Transformer Failure and Common-Mode Loss of

Instrument Power at Nine Mile Point Unit 2 on August 13, 1994," NUREG-1455, October

1991.

21. U.S. Nuclear Regulatory Commission, Insights From NRC Inspection of Low- and

Medium-Voltage Circuit Breaker Maintenance Programs," NRC Information

Notice 99-13, April 29, 1999.

22. U.S. Nuclear Regulatory Commission, "Inspection of Medium-voltage and Low-voltage

Power Circuit Breakers," Revision 1, NRC Temporary Inspection Procedure TI 2515/137

March 9, 1998.

23. U.S. Nuclear Regulatory Commission, letter from A. Thadani, Office of Nuclear

Regulatory Regulation, to Roger Newton, Chairman of Westinghouse Owners Group,

"ATWS Moderator Temperature Coefficient," June 12, 1987.

23
APPENDICES
APPENDIX A

ANTICIPATED TRANSIENT WITHOUT SCRAM

RULE EVENT TREES
DATA FOR BOILING-WATER REACTOR EVENT TREE

SECY-83-293 calculated P(ATWS) for a BWR using the results from the event tree in

Figure A-1 (using the data in Table A-1. 1 for the isolation transient, assuming 30 percent of

transients were isolation transients) and the results from the event tree in Figure A-1 using the

data in Table A-1.2 for the nonisolation transient, (assuming 70 percent of transients were

nonisolation transients). RPS electrical unreliability values include the channel and trip system.

RPS mechanical unreliability values include the hydraulic control unit and the rods. The event

tree values for the RPS unreliability expectations reflect the addition of the ARI system under

the ATWS rule. SECY-83-293 used the results from the event tree in Figure A-2 and the data

in Table A-2.1 to evaluate P(ATWS) for a BWR with automatic SLC initiation.

Automatic Electrical

Trip RPS

Mechanical Initiate Maintain

RPS EPGs Water

Level

Establish

Longterm

Cooling

Unacceptable

Plant

Conditions

Figure A-1 ATWS Rule Event Tree for General Electric Reactor Group

A-1
Table A-1.1 Data for Boiling-Water Reactor ATWS Isolation Transient

Figure A-1 Event Figure A-1 Likelihood Estimates

Expected (SECY-83-293)E Outcome (reference)

AT 4.3 events/RY (30% of 0.5 events/RY (note 1)

transients)

Electrical RPS Reliability <2E-06 3.7E-06 (Ref. 1)

Mechanical RPS Reliability 1 .OE-05 2.1 E-06 (Ref. 1)

Initiate EPGs 0.5 0.5 (assumed, no update available)

Maintain Water Level 0.1 0.1 (assumed, no update available)

Establish Long-term Cooling 0.05 0.05 (assumed, no update

available)

Table A-1.2 Data for Boiling-Water Reactor ATWS Nonisolation Transient

Figure A-1 Event Figure A-1 Likelihood Estimates

Expected (SECY-83-293) Outcome (reference)

AT 4.3 events/RY (70% of 0.5 events/RY (Note 1)

transients)

Electrical RPS Reliability <2E-06 3.7E-06 (Ref. 1)

Mechanical RPS Reliability 1.OE-05 2.1 E-06 (Ref. 1)

Initiate EPGs 0.5 0.5 (assumed, no update available)

Maintain Water Level 0.05 0.05 (assumed, no update

available)

Establish Long-term Cooling 0.05 0.05 (assumed, no update

available)

lIi

l l

.11

A-2
Automatic Electrical

Trip RPS

Mechanical

RPS

Establish

SLCS Longterm

Cooling

Unacceptable

Plant

Conditions

Figure A-2 ATWS Rule Event Tree for General Electric Reactor Group

Automatic Standby Liquid Control System

Table A-2.1 Data for Boiling-Water Reactor

ATWS With Automatic Standby Liquid Control System Initiation

Figure A-2 Event Figure A-2 Likelihood Estimates

Expected (SECY-83-293) Outcome (reference)

AT 4.3 events/RY 0.5 eventslRY(Note 1)

Electrical RPS Reliability <2E-06 3.7E-06 (Ref. 1)

Mechanical RPS Reliability 1.OE-05 2.1E-06 (Ref. 1)

SLCS 0.01 1 .OE-05

Establish Long-term Cooling 0.05 0.05 (assumed, no update

I________________________ __________available)

A-3
DATA FOR PRESSURIZED-WATER REACTOR EVENT TREE

SECY-83-293 evaluated P(ATWS) for the Westinghouse PWRs using the results from the

event tree in Figure A-3 using the data in Table A-3.1 for the turbine trip transient (assuming it

occurred in 70 percent of transients) summed with the results from event tree in Figure A-3

using the data in Table A-3.2 for the nonturbine trip transient (assuming it occurred in 30

percent of transients). RPS electrical unreliability values include the train channel, and the trip

breakers. RPS mechanical unreliability values include the rods. The event tree values for the

RPS unreliability expectations reflect the addition of the shunt trip for the Westinghouse reactor

group. SECY-83-293 used the results from the event tree in Figure A-4 and the data in Table

A-4.1 to evaluate P(ATWS) for the CEIB&W PWRs.

I1

Automatic Electrical Mechanical

Trip RPS RPS

MTC Over

Pressure AFW BPI Unacceptable

Plant

Conditions

Figure A-3 ATWS Rule Event Tree for Westinghouse Reactor Group

A-4
Table A-3.1. Data for Westinghouse Pressurized-Water Reactor

ATWS Turbine Trip Transient

Figure A-2 Figure A-3 Likelihood Estimates

Event Expected (SECY-83-293) Outcome (reference)

AT 4.0 events/RY (70 percent of transients) 0.5 events/RY (note 1)

RPS Electrical 2E-05 2E-05 (Ref. 2 and note 2)

RPS Mechanical 1 E-05 1 .2E-06 (Ref. 2 and note 2)

MTC 0.01 0.01 (Ref. 5)

Overpressure

ATWS Reliability 0.001 0.00045 (Ref. 8)

HPI 0.01 0.01 (assumed, no current data

available)

Table A-3.2 Data for Westinghouse Pressurized-Water Reactor

ATWS Non-Turbine Trip Transient

Figure A-3 Figure A-3 Likelihood Estimates

Event Expected (SECY-83-293) Outcome (reference)

AT 4.0 events/RY (30 percent of transients) 0.5 events/RY (note 1)

RPS Electrical 2E-05 2E-05 (Ref. 2 and note 2)

RPS Mechanical 1 E-05 1 .2E-06 (Ref. 2 and note 2)

MTC 0.01 0.01 (Ref. 5)

Overpressure

ATWS Reliability 0.001 1 E-05

HPI 0.01 0.01

A-5
Automatic Electrical Mechanical

Trip RPS RPS

MTC Over

Pressure AFW HPI Unacceptable

Plant

Conditions

; 

Figure A-4 ATWS Rule Event Tree for Combustion Engineering/

Babcock & Wilcox Reactor Group

Table A-4.1 Data for Combustion Engineering/

Babcock & Wilcox ATWS Transient

Figure A-4 Event Figure A-4 Likelihood Estimates

Expected (SECY-83-293) Outcome (reference)

AT 4.0 events /RY 0.5 events/RY (note 1)

RPS Electrical* 2E-05 CE: 1.OE-05 (Ref. 3, note 3)

B&W: 4.OE-06 (Ref. 4, note 4)

RPS Mechanical 1 E-05 CE: 5.8E-08 (Ref. 3, note 3)

B&W: 1.0E-07 (Ref. 4, note 4)

MTC Overpressure 0.5 0.5 (Ref. 6,7)

ATWS Unreliability 0.001 0.00045 (Ref. 8)

HPI 0.01 0.01

* Values used in risk calculation are a factor of 10 less to account for DSS.

A-6
NOTES AND REFERENCES FOR APPENDIX A

Note 1: The mean of the 1997,1998,1999 BWR and PWR initiating event frequencies from

Appendix B, Table B-2, was 0.50RY and 0.52/RY. Used 0.50 overall to simplify.

Note 2: These values are approximately the same as the RPS unreliabilities for the

Westinghouse Analog 7300 and Eagle 21 systems, which were shown to be similar in

Reference 2.

Note 3: This reflects the average plant RPS unreliabilities for the four representative CE RPS

reliability groups in Reference 3.

Note 4: This reflects the average plant RPS unreliabilities for the two representative B&W RPS

reliability groups in Reference 4.

1. U.S. Nuclear Regulatory Commission, "Reliability Study: General Electric Reactor

Protection System 1984-1995," NUREG/CR-5500, Volume 2, December 1998.

2. U.S. Nuclear Regulatory Commission, "Westinghouse Reactor Protection System

1984-1995," NUREG/CR-5500, Volume 3, December 1998.

3. U.S. Nuclear Regulatory Commission, "Reliability Study: Combustion Engineering Reactor

Protection System 1984-1998," initial draft of NUREGICR-5500, Volume 10, March 2000.

4. U.S. Nuclear Regulatory Commission, " Reliability Study: Babcock & Wilcox Reactor

Protection System 1984-1998," the initial draft of NUREGICR-5500, Volume 4, March 2000.

5. Westinghouse Owners Group, "Westinghouse program ATWS Rule Administration,

Combined Core Performance/Scram Systems Reliability, ACRS Subcommittee Meeting,"

February 19,1988.

6. Babcock & Wilcox Owners Group ATWS Committee, "Effects of Plant and Fuel Changes on

ATWS Basis Presentation to the NRC," February 18,1988.

7. Combustion Engineering Owners Group, "Combustion Engineering Owners Group Meeting

With the NRC Conceming Trends in MTCs," January 11, 198.

8. U.S. Nuclear Regulatory Commission, "Reliability Study: Auxiliary Feedwater System

NUREG/CR-5500, Volume 1,

A-7
APPENDIX B

PLANT-SPECIFIC AND GENERAL

ATWS INFORMATION BY REACTOR GROUP
Plant Specific ATWS and General Information by Reactor Type

Table B-1 Operating Pressurized-Water Reactor Data

Plant Plant CDF ATWS Percent ATWS Modification Miscellaneous Data

CDF CDF of Summary Plant CDF

1997,1998,1999 Summary of NRC Survey of

ATWS Initiating Event 1994-1996 PWR MTC

Frequency (per RY) Technical Specification Limits

Arkansas Nuclear 4.67E-05 9.93E-07 2.13 DSS 0,0,0 +9E-05 (0 above 95% RTP) One AMSAC

Unit 1

Arkansas Nuclear 3.40E-05 1.02E-06 3.00 DSS 0,0,0 +5E-05 (0 above 70% RTP) One DEFAS

Unit 2

Beaver Valley 2.14E-04 4.30E-05 20.1 AMSAC 2,1,0 0

Unit 1

Beaver Valley 1.92E-04 8.06E-06 4.20 AMSAC 2,0,0 0

UnIt 2

Braidwood 2.74E-05 3.70E-07 1.35 AMSAC 0,0,0 +7E-05 below 70% RTP (linear to

Unit 1 0 from 70 to 100%)

Braidwood 2.74E-05 3.70E-07 1.35 AMSAC 0,1,2 +7E-05 below 70% RTP (linear to

Unit 2 0 from 70 to 100%)

Byron 3.09E-05 4.20E-07 1.36 AMSAC Unit 1: 0,0,1 +7E-05 below 70% RTP (linear to

Units 1&2 Unit 2: 1,0,0 0 from 70 to 100%)

Callaway 5.85E-05 4.80E-07 0.821 AMSAC 0,0,1 I

B-1
Plant Specific ATWS and General Information By Reactor Group

Table B-1 Operating Pressurized-Water Reactor Data (Cont.)

Plant Plant CDF ATWS Percent ATWS Modification Miscellaneous Data

CDF CDF of Summary

Plant CDF

1997, 1998,1999 Summary of NRC Survey of

ATWS Initiating Event 1994-1996 PWR MTC

Frequency (per RY) Technical Specification Limits

Calvert Cliffs 2.40E-04 2.40E-05 10.0 DTT 1,0,1 +7E-05 below 70% RTP (linear to

Unit 1 DSS +3E-05 from 70 to 100%)

DAFAS

Calvert Cliffs 2.40E-04 2.40E-05 10.0 DTT 0,0,0 +7E-05 below 70% RTP (linear Unit 2 DSS to +3E-05 from 70 to 100%)

DAFAS

Catawba 5.80E-05 1.OOE-06 1.72 AMSAC 0,0,0 +7E-05 below 70% RTP (linear Unit 1 to 0 from 70 to 100%)

Catawba 5.80E-05 1.00E-06 1.72 AMSAC 1,0,1 +7E-05 below 70% RTP (linear Unit 2 to 0 from 70 to 100%)

Comanche 5.72E-05 5.OOE-06 8.74 AMSAC 1,0,0 +5E-05 below 70% RTP (linear Peak Unit 1 to 0 from 70 to 100%)

Comanche 5.72E-05 5.00E-06 8.74 AMSAC 0,1,0 +5E-05 below 70% RTP (linear Peak Unit 2 to 0 from 70 to 100%)

Crystal River 1.53E-05 1.OOE-10 0.000655 DSS 0,1,0 +9E-05 (0 above 95% RTP) Unit 3 AMSAC

Davis-Besse 6.60E-05 3.54E-07 0.536 DSS 1,2,0 +9E-05 (0 above 95% RTP)

I__ ___AMSAC _ _ _ _ _ _ _ 

B-2

+_1.. . -1
Plant Specific ATWS and General Information By Reactor Group

Table B-I1 Operating Pressurized-Water Reactor Data (Cont.)

Plant Plant CDF ATWS Percent ATWS Modification Miscellaneous Data

CDF CDF of Summary

Plant CDF

1997,1998,1999 Summary of NRC Survey of

ATWS Initiating Event 1994-1996 PWR MTC

Frequency (per RY) Technical Specification Limits

DC Cook 6.26E-05 2.85E-06 4.55 AMSAC 0,0,0 SE-05 below 70% RTP

Unit 1 (linear to 0 from 70 to 100%)

DC Cook 6.26E-05 2.85E-06 4.55 AMSAC 0,0,0 +5E-05 below 70% RTP

Unit 2 (linear to 0 from 70 to 100%)

Diablo Canyon 8.80E-05 7.OOE-07 0.795 AMSAC 0,0,1 +5E-05 below 70% RTP

Unit 1 (linear to 0 from 70 to 100%)

Diablo Canyon 8.80E-05 7.OOE-07 0.795 AMSAC 2,0,0 +5E-05 below 70% RTP

Unit 2 (linear to 0 from 70 to 100%)

Fariey 1.30E-04 7.30E-08 0.0562 AMSAC 0,1,1 +7E-05 below 70% RTP

Unit 1 (linear to 0 from 70 to 100%)

Farley 1.30E-04 7.30E-08 0.0562 AMSAC 0,0,0 +7E-05 below 70% RTP 

Unit 2 (linear to 0 from 70 to 100%)

Fort Calhoun 1.36E-05 2.86E-07 2.10 DTT 0,010 +5E-05 below 80% RTP (+2E-05

DSS above 80%)

DAFAS

Ginna 8.74E.05 1.60E-07 0.183 AMSAC 0,0,2 +5E-05 below 70% RTP (0 above 70%)

B-3
Plant Specific ATWS and General Information By Reactor Group

Table B-1 Operating Pressurized-Water Reactor Data (Cont.)

Plant Plant CDF ATWS Percent ATWS Modification Miscellaneous Data

CDF CDF of Summary

Plant CDF

1997,1998,1999 Summary of NRC Survey of

ATWS Initiating Event 1994-1996 PWR MTC

Frequency (per RY) Technical Specification Umits

Harris 7.OOE-05 5.OOE-06 7.14 AMSAC 3,0,2 +5E-05 below 70% RTP

(linear to from 70 to 100%)

Indian Point 3.1 3E-05 1.81 E-06 5.78 AMVSAC 3,0,1 0

Unit 2

Indian Point 4.40E-05 8.70E-06 19.80 AMSAC 2,1,2 0

Unit 3 i

Kewaunee 6.65E-05 6.85E-08 0.103 AMSAC w/o 0,1,0 0

C-20

permissive

McGuire 4.OOE-05 1.50E-06 3.75 AMSAC 1,0,0 +7E-05 below 70% RTP

Unit 1 (linear to 0 from 70 to 100%)

McGuire 4.00E-05 1.50E-06 3.75 AMSAC 2,1,1 +7E-05 below 70% RTP

Unit 2 (linear to 0 from 70 to 100%)

Millstone 3.42E-05 1.83E-08 0.0535 DTT 0,0,0 +7E-05 (4 above 70% RTP)

Unit 2 DSS

DAFAS

Millstone 5.61 E-05 3.40E-06 6.06 AMSAC 0,1,0

Unit 3

B-4
Plant Specific ATWS and General Information By Reactor Group

Table B-1 Operating Pressurized-Water Reactor Data (Cant.)

Plant Plant CDF ATWS Percent ATWS Modification Miscellaneous Data

CDF CDF of Summary

Plant CDF

1997,1998,1999 Summary of NRC Survey of

ATWS Initiating Event 1994-1996 PWR MTC

Frequency (per RY) Technical Specification Limits

North Anna 7.16E-05 4.20E-07 0.60 AMSAC Unit 1: 0,0,0 +6E-05 (0 above 70% RTP)

Units 1 &2 Unit 2: 0,1,0

Oconee 2.30E-05 1.00E-07 0.435 DSS 0,0,2 +9E-05 (0 above 95% RTP)

Unit 1 AMSAC

Oconee 2.30E-05 1.001E-07 0.435 DSS 1,2,4 +9E-05 (0 above 95% RTP)

Unit 2 AMSAC

Oconee 2.30E-05 1 .OOE-07 0.435 DSS 1,1,0 +9E-05 (0 above 95% RTP)

Unit 3 AMSAC

Palisades 5.07E-05 4.30E-06 8.48 DTT 0,0,0 +5E-05

DSS

DAFAS

Palo Verde 9.00E-05 3.08E-06 3.42 DTT Unit 1: 1,1,1 Linear from +5E-05 at 0 to 0 at

Units 1 & 2 DSS Unit 2: 0,0,1 100% RTP

DAFAS

Palo Verde 9.00E-05 3.08E-06 3.42 DTT 1,0,0 Linear from +5E-05 at 0 to 0 at

Unit 3 DSS 100% RTP

DAFAS

Point Beach 1.15E-04 2.72E-07 0.237 AMSAC w/o 0,0,0 +5E-05 below 70% RTP (linear to

Unit 1 C-20 0 from 70 to 100% RTP)

permissive .-

B-5
Plant Specific ATWS and General Information By Reactor Group

Table B-1 Operating Pressurized-Water Reactor Data (Cont)

Plant Plant CDF ATWS Percent ATWS Modification Miscellaneous Data

CDF CDF of Summary

Plant CDF

1997,1998,1999 Summary of NRC Survey of

ATWS Initiating Event 1994-1996 PWR MTC ______ ___________ Frequency (per RY) Technical Specification Limits

Point Beach 1.15E-04 2.72E-07 0.237 AMSAC w/o 0,0,0 +5E-05 below 70% RTP (linear to

Unit 2 C-20 0 from 70 to 100% RTP)

permissive

Prairie Island 5.05E-05 3.20E-07 0.634 AMSAC 1,2,1 +7E-05 ITC less than +5E-05

Unit 1 (ITC 0 above 70% RTP)

Prairie Island 5.05E-05 3.20E-07 0.634 AMSAC 0,1,0 +7E-05 ITC less than +5E-05

Unit 2 (ITC 0 above 70% RTP)

Robinson 3.20E-04 5.70E-06 1.78 AMSAC 1,2,0 +5E-05 below 50% RTP

Unit 2 0 above 50% RTP

Salem 5.20E-05 1.40E-06 2.69 AMSAC 0,0,2 0

Unit 1

Salem 5.50E-05 1.30E-06 2.436 AMSAC 0,0,0 0

Unit 2

San Onofre 3.OOE-05 2.70E-06 9.00 DTT Unit 1: 0,0,0 +5E-05 (above 70% RTP)

Units 2 & 3 DSS Unit 2: 0,0,0

DEFAS

St. Lucie 2.30E-05 4.13E-07 1.80 DTT 1,0,1 +7E-05 below 70% RTP (+2E-05

Unit 1 DSS above 70% RTP)

DAFAS

B-6
Plant Specific ATWS and General Information By Reactor Group

Table B-1 Operating Pressurized-Water Reactor Data (Cont.)

Plant Plant CDF ATWS Percent ATWS Modification Miscellaneous Data

CDF CDF of Summary

Plant CDF

1997,1998,1999 Summary of NRC Survey of

ATWS Initiating Event 1994-1996 PWR MTC

Frequency (per RY) Technical Specification Limits

St. Lucle 2.62E-05 1.76E-06 6.72 DTT 0,0,0 +7E-05 below 70% RTP (+2E-05

Unit 2 DSS above 70% RTP)

DAFAS

Seabrook 6.60E-05 6.63E-06 10.1 AMSAC 1,1,0 0

Sequoyah 1.70E-04 7.1OE-06 4.18 AMSAC 0,2,0 0

Unit 1

Sequoyah 1.70E-04 7.1 OE-06 4.18 AMSAC 0,2,0 0

Unit 2

South Texas Unit 1 4.30E-5 3.OOE-07 0.698 AMSAC 1,0,3 +7E-05 below 70% RTP

(linear to 0 from 70 to 100%)

South Texas Unit 2 4.30E-5 3.OOE-07 0.698 AMSAC 1,1,1 +7E-05 below 70% RTP

(linear to 0 from 70 to 100%) -

V.C. Summer 2.00E-04 2.03E-06 1.02 AMSAC 1,0,1 +7E-05 below 70% RTP

(linear to 0 from 70 to 100%)

Surry Unit 1 1.25E-04 3.20E-07 0.256 AMSAC 0,2,0 +3E-05 below 50% RTP

(linear to 0 from 50 to 100%)

Surry Unit 2 1.25E-04 3.20E-07 0.256 AMSAC 0,0,1 +3E-05 below 50% RTP

(linear to 0 from 50 t o 100%)

B-7
Plant Specific ATWS and General Information By Reactor Group

Table B-1 Operating Pressurized-Water Reactor Data (Cont)

Plant Plant CDF ATWS Percent ATWS Modification Miscellaneous Data

CDF CDF of Summary

Plant CDF

1997, 1998, 1999 Summary of NRC Survey of

ATWS Initiating Event 1994-1996 PWR MTC

Frequency (per RY) Technical Specification Umits

Three Mile 4.49E-05 1.OOE-1 0 0.00022 DSS 1,0,0 +9E-05 below 95% RTP

Island Unit 1 AMSAC

Turkey Point 3.73E-04 4.40E-06 1.18 AMSAC 1,1,0 +5E-05 below 70% RTP

Unit 3 (linear to O from 70 to 100% RTP)

Turkey Point 3.73E-04 4.40E-06 1.18 AMSAC 1,0,0 +5E-05 below 70% RTP

Unit 4 _ (linear to O from 70 to 100% RTP)

Vogtle 4.90E-05 1.13E-07 0.23 AMSAC 0,0,0 +7E-05 below 70% RTP

Unit 1 (linear to O from 70 to 100% RTP)

Vogtle 4.90E-05 1.13E-07 0.23 AMSAC 0,2,0 +7E-05 below 70% RTP

Unit 2 (linear to O from 70 to 100% RTP)

Waterford 1.80E-05 1.30E-07 0.722 DTT 0,0,1 Not surveyed

Unit 3 DSS

DAFAS

Watts Bar 8.OOE-05 3.80E-06 4.75 AMSAC 3,1,0 0

Unit 1

Wolf Creek 4.20E-05 3.1 OE-08 0.0738 AMSAC 1,0,1 +6E-05 below 70% RTP __________________ I________ ____ ___ ___ ___ ____ _ _ ___ ___ ___ ___ ___ ___ ___ (linear to 0 from 70 to 100% R TP)

B-8
Plant Specific ATWS and General Information by Reactor Type

Table B-2 Operating Boiling-Water Reactor Data

Plant Plant CDF ATWS CDF Percent ATWS Modification 1997,1998, 1999

CDF of Plant CDF Summary ATWS Initiating Event

Frequency (per RY)

Browns Ferry 4.80E-05 1.30E-06 2.71 ARI Not available

Unit 1 RPT

Increased SLCS boron

concentration

Browns Ferry 4.80E-05 1.30E-06 2.71 ARI 2.1.2

Unit 2 RPT

Increased SLCS boron

concentration

Browns Ferry 4.80E-05 1.30E-06 2.71 ARI 0,0,0

Unit 3 RPT

Increased SLCS boron

concentration

Brunswick Unit 1 2.70E-05 7.OOE-07 2.59 ARI 0,0,0

RPT

Increased SLCS addition rate

(2-pump operation)

Brunswick Unit 2 2.70E-05 7.00E-07 2.59 ARI 0,0,2

APT

Increased SLCS addition rate

(2-pump operation)

Clinton 2.66E-05 1.40E-07 0.526 ARI 0,0,0

RPT

Increased SLCS boron

concentration

B-9
Plant Specific ATWS and General Information by Reactor Type

Table B-2 Operating Boiling-Water Reactor Data (Cont.)

Plant Plant CDF ATWS CDF Percent ATWS Modification 1997, 1998, 1999

CDF of Plant CDF Summary ATWS Initiating Event

Frequency (per RY)

Cooper 7.97E-05 3.90E-06 4.89 ARI 0,0,0

RPT

Increased SLCS addition rate

(2-pump operation) and suction

piping mods

Dresden 1.85E-05 5.34E-07 2.89 ARI 1,3,0

Unit 2 RPT

Increased SLCS addition rate

(2-pump operation) and boron

concentration

Dresden 1.85E-05 5.34E-07 2.89 ARI 0,1,1

Unit 3 RPT

Increased SLCS addition rate

(2-pump operation) and boron

concentration

Duane Arnold 7.84E-06 1.90E-06 24.2 ARI 0,0,0

RPT

Increased SLCS addition rate

(2-pump operation) and boron

concentration

Fermi 5.70E-06 1.80E-06 31.6 ARI 0,1,0

APT

Increased SLCS boron

concentration

B-10
Plant Specific ATWS and General Information by Reactor Type

Table B-2 Operating Boiling-Water Reactor Data (Cont.)

Plant Plant CDF ATWS CDF Percent ATWS Modification 1997,1998,1999

CDF of Plant CDF Summary ATWS Initiating Event

Frequency (per RY)

FitzPatrick 1.92E-06 1.20E-08 0.625 ARI 0,1,2

RPT

Increased SLCS boron

concentration

Grand Gulf 1.72E-05 5.56E-08 0.323 ARI 0,0,0

RPT

Increased SLCS boron

concentration

Hatch 2.23E-05 3.84E-07 1.72 ARI 0,0,1

Unit 1 RPT

Increased SLCS boron

concentration

Hatch 2.36E-05 4.78E-07 2.03 ARI 1,0,2

Unit 2 RPT

Increased SLCS boron

concentration

Hope Creek 4.63E-05 7.45E-07 1.61 ARI 0,1,0

RPT

Increased SLCS addition rate

(2-pump operation) and boron

concentration

B-11
Plant Specific ATWS and General Information by Reactor Type

Table B-2 Operating Boillng-Water Reactor Data (Cont.)

Plant Plant CDF ATWS CDF Percent ATWS Modification 1997,1998,1999

CDF of Plant CDF Summary ATWS Initiating Event _________________ ________________ ___________ Frequency (per RY)

LaSalle 4.74E-05 1.87E-07 0.395 ARI 0,0,1

Unit 1 RPT

Increased SLCS boron

concentration

LaSalle 4.74E-05 1.87E-07 0.395 ARI 0,0,1

Unit 2 RPT

Increased SLCS boron

concentration

Limerick 4.30E-06 9.30E-07 21.6 ARI 0,0,2

Unit 1 RPT

Auto start SLC

Limerick 4.30E-06 9.30E-07 21.6 ARI 0,0,1

Unit 2 RPT

Auto start SLC

Monticello 2.60E-05 2.50E-06 9.62 ARI 0,0,1

RPT

Increased SLCS boron

concentration

Nine Mile Point 5.50E-06 5.40E-07 9.82 ARI 0,0,2

Unit 1 RPT

Increased SLCS addition rate

(2-pump operation) and boron

concentration

B-12
Plant Specific ATWS and General Information by Reactor Type

Table B-2 Operating Boling-Water Reactor Data (Cont.)

Plant Plant CDF ATWS CDF Percent ATWS Modification 1997,1998,1999

CDF of Plant CDF Summary ATWS Initiating Event

Frequency (per RY)

Nine Mile Point 3.10E-05 1.10E-06 3.55 ARI 0,0,2

Unit 2 RPT

Increased SLCS boron

concentration

Auto start SLC

Oyster Creek 3.90E-06 2.40E-07 6.15 ARI 0,0,0

RPT

increased SLCS boron

concentration

Peach Bottom 5.53E-06 1.44E-06 26.0 ARI 1,0,1

Unit 2 RPT

increased SLCS boron

concentration

Peach Bottom 5.53E-06 1.44E-06 26.0 ARI 0,0,0

Unit 3 RPT

Increased SLCS boron

concentration

Perry 1.30E-05 4.74E-06 36.5 ARI 3,1,0

RPT

Pilgrim 5.80E-05 4.1 OE-6 7.07 ARI 1,0,1

RPT

Increased SLCS boron

concentration

B-13
Plant Specific ATWS and General nformation by Reactor Type

Table B-2 Operating Boiling-Water Reactor Data (Cont.)

Plant Plant CDF ATWS CDF Percent ATWS Modification 1997, 1998, 1999

CDF of Plant CDF Summary ATWS Initiating Event

Frequency (per RY)

Quad Cities Unit 1 1.20E-06 7.61 E-08 6.34 ARI 0,2,1

RPT

Increased SLCS addition rate

(2-pump operation) and boron

concentration

Quad Cities Unit 2 1.20E-06 7.61 E-08 6.34 ARI 0,1,0

RPT

Increased SLCS addition rate

(2-pump operation) and boron

concentration

River Bend 1.55E-05 1.OOE-10 0.000645 ARI 1,0,1

RPT

Increased SLCS boron

concentration

Susquehanna Unit 1 1.70E-05 3.20E-07 1.88 ARI 0,1,1

RPT

Increased SLCS addition rate

(2-pump operation) and suction

piping mods

Susquehanna Unit 2 1.70E-05 3.20E-07 1.88 ARI 0,2,1

RPT

Increased SLCS addition rate

(2-pump operation) and suction

piping mods

B-14
Plant Specific ATWS and General Information by Reactor Type

Table B-2 Operating Bolilng-Water Reactor Data (Cont.)

Plant Plant CDF ATWS CDF Percent ATWS Modification 1997, 1998, 1999

CDF of Plant CDF Summary ATWS Initiating Event

Frequency (per RY)

Vermont Yankee 4.30E-06 8.85E-07 20.6 ARI 2,1,0

RPT

Increased SLCS boron

concentration

Washington Nuclear 1.75E-05 6.25E-07 3.57 ARI 0,1.0

Plant Unit 2 RPT

Increased SLCS addition rate

(2-pump operation) and boron

concentration

B-1 5
Table B-3 Reactor Protection System Unreliability Uncertainties

General Electric Westinghouse Babcock & Combustion

Wilcox Engineering

Before Mean 1 .9E-04 2.0E-04 0 0

(data through 1983)

Lower 5%-Upper 95% 0.0098-9.OE-04 0.1-9.5E-04 0-2.6E-03 0-2.5E-03

Expected Mean 1.2E-05 3.OE-05 1.2E05 1.2E-05

(improvement based on

ATWS regulatory actions)

Outcome Mean 5.8E-0 2.1 E-05 4.OE-08 1 E-05

(data since 1984) Lower 5%-Upper 96% 1 .8-14E406 0.58-5.7E-05 2.2-7.8E-0B 0.S5-2.5E405

Table B-4 General Electric Reactor Protection System Common-cause Failure Insights

(NUREGICR-5500, Volume 3)

RPS Components of Greatest Importance to RPS Reliability and Risk Greatest CCF

Failure Probabity

Importance Measures Important to RPS Important to CDF

Reliability

FV Importance Risk Increase Ratio

Channel Bistabie 6.5E-02 1.7E+5 3.8E-07

Solenoid-operated Valve Scram Pilot Value 2.9E-01 1.7E+5 1.7E-0

Rod 4.8E-02 1 .7E+5 2.5E-07

HCU AOV scram inlet & outiet valves 1.2-03 1.7E+5 6.9E-09

HCU Accumulator 1.9E-02 1.7E+5 1.1 E-07

Train Relays 4.7E-02 1.7E+5 2.8E-07

Channel Bistable 5.3E-01 1 .7E+5 3.1 E-06

Table B-5 Westinghouse Reactor Protection System Common-cause Failure Insights

(NUREGICR-5500, Volume 2)

RPS Components of Greatest Importance to RPS Reilabiity and Risk Greatest CCF

Probabiit

Importance Measures Important to RPS Important to CDF

Reliability

FV Importance Risk Increase Ratio

RCCAICRDM 5.6202 4.6E+4 1 .2E-06

RTB 7.4E-02 4.6E+4 1.6E-06

SSPS Universal Card 9.7E-02 4.6E+4 2.1 E-06

Undervoltage Trip Driver Card 4.8E-01 4.6E+4 1 E-05

Channel Bistables 2.2-12E-02 .27-4.3E+4 1.2-4.2E-05

Channel Pmcessing Module (CCPCDTCMM) 1.7-8.02-02 .27-4.3E+4 CCP: 1.5E-05

CDT: 2.5E-04

Train Bistables

3.5E-03 4.3E+4 8.2E-06

B-1 6
Table B-6 Babcock and Wilcox Reactor Protection System Common-cause Failure Insights

(NUREG/CR-5500, Volume 4)

RPS components of greatest importance to RPS reliability and Risk CCF Failure

Probability

Importance Measures Important to RPS Important to CDF

reliability

FV Importance Risk Increase Ratio

Rod 2.3E-02 2.3E+5 1 E-07

Reactor Trip Breaker (Mechanically) 5.2E-01 2.3E+5 2.3E-06

Trip Logic-Trip Relay 9.OE-02 2.3E+5 4.OE-07

Channel Bistables 2.7E-01 2.3E+5 1.2E-06

Trip Logic-Logic Relay 6.OE-02 2.3E+5 2.7E-06

Table B-7 Combustion Reactor Protection System Common-cause Failure Insights

(NUREG/CR-5500, Volume 10)

RPS Components of Greatest Importance to RPS Reliability and Risk CCF Failure

Importance Measures Important to RPS Important to CDF Probability

Reliability

FV Importance Risk Increase Ratio

Reactor Trip Breaker (Mechanically) 3.1 E-01 9.OE+4 3.4E-06

Rod/Assembly 5.OE-03 9.OE+4 5.5E-08

Trip Logic-Trip Relay 5.8E-01 9.OE+4 6.4E-06

Channel Bistables 8.OE-02 8.9E+4 9.OE-07

Trip Logic-Logic Relay 1.9E-02 8.9E+4 2.1 E-07

B-17
Table B-8 Value Impact Data for Each Reactor Group

P(ATWS) Value Impact V-I

$ Million $ Million ratio

Modifications for the ATWS Rule (frequency

of an

ATWS

event/year)

GE

0 Baseline 5.3E-05

1. Increase SLCS to 86 GPM and ARI 1.2E-05 12.3 3.5 3.5

2. Increase SLCs capacity and automatic initiation

(new plants) 2.6E-06 2.8 5.0 0.56

w

0 Baseline 3.7E-05

1 a. Diverse AFW automatic actuation and turbine

trip (AMSAC) 5.8E-06 9.4 2.8 3.3

CEIB&W

0 Baseline 8E-05

1. DSS and diverse turbine trip and AFW

Initiation 2.2E-05 17.4 5.5 3.2

!I

B-1 8
APPENDIX C

RESOLUTION OF COMMENTS
Appendix C

Resolution of Comments

On October 18, 2000, a letter entitled Draft Report, Regulatory Effectiveness of the Anticipated

Transient Without Scram Rule," was sent to David Modeen, Director of Engineering, Nuclear Energy

Institute, from Farouk Eltawila, Acting Director, Division of Systems Analysis and Regulatory

Effectiveness, Office of Nuclear Regulatory Research, for review prior to finalizing the report.

The letter requested comments regarding (1) the reasonableness of the approach, (2) the

appropriateness of the conclusions, (3) and other regulations, accompanying regulatory guides, and

inspection documents that should be assessed to make U.S. Nuclear Regulatory Commission

activities more effective, efficient and realistic. Letters were also sent to other members of the

industry, the NRC staff, and made publically available.

In response, letters with comments were received from the Union of Concerned Scientists, General

Electric, Combustion Engineering, and the Westinghouse House Owners Group. These letters are

entered into ADAMS, Package Accession Number ML010220103, and are publicly available.

Comments were received from the NRC staff of the Reactor Systems Branch (SRXB) and the

Probabilistic Safety Assessment Branch (SPSB), Division of Systems Safety and Analysis (DSSA),

Office of Nuclear Reactor Regulation (NRR), and Region II. The NRC staff also met with the

Advisory Committee on Reactor Safety (ACRS) on February 2, 2001. The ACRS wrote a letter to

the Executive Director of Operations (EDO), March 8, 2001 (ML010730375), with their comments

and a response from the EDO to the ACRS was signed April 9, 2001 (ML01 0950412).

The resolution to each of the comments is addressed below. The comments are presented by

organization, in the order of the date received, followed by a restatement of their comments

verbatim, and the resolution of each comment. The resolution is stated specifically in terms of 'no

change required," with an explanation, if appropriate, or a revision to the report that is shown in

quotations and italics. Conforming changes were made throughout the report but are not listed in

the resolution of the comment.

C-1
UNION OF CONCERNED SCIENTISTS (UCS) provided the following comments in a letter to the

NRC dated October 30, 2000.

UCS COMMENT 1: On page x of the Executive Summary, the fourth bullet states that Hall four

reactor types achieved the risk goal of P(ATWS)<l .OE-5 per reactor year." However, data in the

report suggests that this conclusion may not be accurate. For example:

a) The last bullet on page x continuing onto page xi points out that some BWR risk analyses may

underestimate the risk of ATWS" because Probabilistic risk assessmentfindividual plant

examinations for BWRs indicate large variations in the assumptions for reliability of human actions

in response to an ATWS. Similarities in design, procedures, and training argue against such

variability." According to Table I on page 7, the outcome for BWRs after implementation of the

ATWS rule modifications was P(ATWS) of 1 .OE-6, or one order of magnitude below the stated risk

goal of 1.0E-5. Recommendation: Perform a perturbation study to determine if the least

conservative assumption for human action reliability could cause the BWR P(ATWS) to increase

above 1.0E-5. If not, revise the report to include the results of the study to support the conclusion.

Otherwise, revise the conclusion accordingly.

RESOLUTION: Section 3 emphasizes that this is a generic assessment, not plant specific - "the

assessment does not address plant-specific issues as these continue to be addressed elsewhere

in the regulatory process."

Changes were made in several sections of the report to emphasize that this is a generic

assessment, not plant specific.

This assessment is intended to determine if the ATWS rule achieved its expected results on a

generic basis, not plant specific. The generic risk calculation in this assessment used

reasonable human performance assumptions, not the low HEPs assumed in some of the

licensee risk assessments. With those values, the generic P(ATWS) met the ATWS rule goal.

The risk quantification method used to calculate the expected risk in the original ATWS analysis

was used for this assessment in order to maintain a consistent approach. In so doing, the

improvement in risk that was identified would more obviously be the result of less frequent

initiating events and improved equipment performance rather than changes in the risk model or

assumptions. Since most of the relevant operator training and procedure improvements were

made prior to the ATWS rule, the same assumptions regarding operator performance were used

in this assessment as in the original ATWS risk calculations. This approach provides a

reasonable framework for understanding the basic elements of the ATWS risk - initiating event

frequency, scram system reliability, and mitigative reliability.

b) The next-to-last bullet on page x states that ATWS mitigation capability on a PWR is highly

dependent on the moderator temperature coefficient (MTC). Mitigative functions are considered

non-viable N the ATWS peak pressure exceeds 3200 psig; and a sufficiently negative MTC will limit

the ATWS peak pressure. On page 4, the report states During the first part of the fuel cycle below

100 percent power, the MTC can be positive. If an ATWS occurs when the MTC is either positive or

insufficiently negative to limit reactor power and the ATWS pressure increase, all subsequent

mitigative functions are likely to be ineffective. The percentage of the fuel cycle time when the MTC

is insufficient to maintain the RCS pressure below 3200 psig during an ATWS is designated the

'unfavorable exposure time.' At the time of the ATWS rulemaking, the UET was assumed to be

C-2
1 percent for W [Westinghouse] and 50 percent for B&W/CE reactors...". Table 1 on page 7 reports

the UET greater than expected for a few plants." Table 6 on page 14 reports that a peak pressure

of 3200 psia is exceeded 1 to 10 percent of the time for Westinghouse plants. Recommendation:

Perform a perturbation study to determine if the plants exceeding the expected unfavorable

exposure time causes the P(ATWS) to increase above 1.OE-5. If not, revise the report to include

the results of the study to support the conclusion. Otherwise, revise the conclusion accordingly.

RESOLUTION: Based on a comment from Westinghouse, the phrase "unfavorable exposure

time" was replaced by unfavorable MTC" which has a similar meaning. See response to WOG

General Comment 1.

Section 3.2.6 revised to read:

If the W assessment were done using the same method as B&W and CE with an

"unfavorable MTC"changing from 1 percent to 40 percent of the fuel cycle, the ATWS risk would

increase by a factor of about 20 to 4.3E-6."

(c) The next-to-last bullet on page x states that ATWS mitigation capability on a PWR is highly

dependent on the moderator temperature coefficient (MTC). Mitigative functions are considered

non-viable if the ATWS peak pressure exceeds 3200 psig; and a sufficiently negative MTC will limit

the ATWS peak pressure." On page 11, Section 3.2.4 of the report states Comparison of the

ATWS CDF for all the plants in each reactor group in Appendix B to the P(ATWS) expectation for

each reactor group found that six licensees do not meet the ATWS risk goals." This statement

contradicts, or at least seriously undermines, the conclusion that all reactor types achieved the

ATWS goal. Recommendation: Revise the overall conclusion to indicate that some reactors did

not achieve the ATWS risk goal.

RESOLUTION: Conclusions sections revised to read:

During the ATWS....on a generic basis, all four reactor types achieved the ATWS rule risk goal.

A few individual plants had somewhat higher risk as a result of using lower levels of RPS

reliability rather than the industry average or assumed additional equipment failures."

d)The next-to-last paragraph on page 11 reports Several [PWR] plants routinely operate with

blocked power-operated relief valves and their IPEs may underestimate ATWS risk."

Recommendation: Perform a perturbation study to determine if routine operation with blocked

power-operated relief valves causes the P(ATWS) to increase above 1.OE-5. If not, revise the

report to include the results of the study to support the conclusion. Otherwise, revise the conclusion

accordingly.

RESOLUTION: The report conclusion was revised. See resolution to UCS Comment 1 (c).

UCS COMMENT 2: The report indicates that the effectiveness of the ATWS rule for PWRs is highly

dependent on the MTC issue. The last sentence on page 6 states *Information on the plant and

reactor group MTCs was obtained from a 1994 NRC survey (Ref. 15), and owners' group

presentations to the NRC (Ref. 16), (Ref. 17), (Ref. 18).m The latter three references all date back

to 1988. Thus, information on this important parameter is at least six year old. Recommendation:

Determine the current plant and reactor group MTCs. Revise the report as appropriate. As a

minimum, the report should be revised to reference contemporary sources for the MTCs.

C-3
RESOLUTION: No change required. The readily available information serves to make the point

and support the conclusions without the cost of an additional survey.

UCS COMMENT 3: The third paragraph on page 4 states in SECY-83-293 it was assumed that a

peak pressure of above 3200 psig was unacceptable for ASME Service Level C. ... Also steam

generator tubes might fail before other primary coolant system components and bypass

containment." Table 6 on page 14 reports peak pressures of 3962 psia for CE plants, 3600-4000

psia for B&W plants, and >3200 psia up to 10 percent of the time for Westinghouse plants. Page 1

defines an ATWS as an anticipated operational occurrence followed by failure of the reactor trip

system (RTS) portion of the reactor protection system (RPS)." On October 11, 2000, a

subcommittee of the Advisory Committee on Reactor Safeguards conducted a meeting regarding

the differing professional opinion (DPO) initiated by Dr. Joram Hopenfeld of the NRC staff. The

meeting transcript is available on the Internet at htto:lwww.nrc.aov/ACRS/rrsl/Trans Let/index

toI/ACRS sub tran/adhocO01 011. In addition to the DPO, the NRC also created Generic Safety

Issue 163 (GSI-1 63) about Dr. Hopenfeld's concerns. Basically, Dr. Hopenfeld is concerned that

allowing nuclear power plants to operate with cracked steam generator tubes increases the potential

for unacceptable consequences from design bases events. During his presentation to the ACRS

subcommittee, Dr. Hopenfeld referred to a May 2000 memo issued by Westinghouse to plant

owners informing them that the support plates inside steam generators were designed for a

differential pressure of 1,500 psid. If an ATWS can result in a reactor coolant system pressure in

excess of 3,200 psia, it seems certain that a differential pressure of at least 2,000 psid will be

experienced across the steam generator tube walls. If the operational occurrence that triggered the

ATWS is one that causes secondary side pressure to decrease, then the resulting differential

pressure across the steam generator tube walls may approach the reactor coolant system pressure.

The ATWS rule might not be adequate by accepting peak pressures of the reactor coolant system in

excess of 3,200 psia if those pressures cause degraded steam generator tubes to rupture.

Recommendation: Evaluate the impact of primary side pressures exceeding 3,200 psia on steam

generator tube integrity. If tube integrity is assured even under those extreme conditions, revise the

report to include the results of the evaluation. Otherwise, revise the report to reflect that the ATWS

rule as currently implemented compromises steam generator tube integrity.

RESOLUTION: No change required. It was not within the scope of the ATWS rule making or

this report to determine what components will fail as a result of exceeding 3200 psia. The steam

general tube issues of concern are being addressed as Generic Safety Issue 188, "Steam

Generator Tube Leaks/Ruptures Concurrent with Containment Bypass."

On February 1, 2001, Dr. Dana Powers, Chairman of the Advisory Committee for Reactor

Safeguards (ACRS) Ad Hoc Subcommittee on a Differing Professional Opinion provided the

Executive Director for Operations (EDO) a copy of its summary report on matters pertaining to a

differing professional opinion on steam generator tube integrity. In its draft report,

"Voltage-Based Alternative Repair Criteria," the Ad Hoc Subcommittee discussed its conclusions

and recommendations. This draft report was issued in final form as NUREG-1740 dated

February 2001. On March 5, 2001, the EDO directed that the Office of Nuclear Reactor

Regulation (NRR) and the Office of Nuclear Regulatory Research (RES) review the ACRS Ad

Hoc Subcommittee report and develop a joint action plan to address the conclusions and

recommendations (ADAMS Accession Number ML01 0670217).

UCS COMMENT 4: The NRC is to be commended for evaluating the effectiveness of the ATWS

rule. My first impression was that this evaluation was long overdue, but after having reviewed the

C-4
draft report, I realize that the delay was necessary in order to compile sufficient operating

experience to permit a meaningful evaluation.

RESOLUTION: No change required.

UCS COMMENT 5: It is a great idea to provide the ADAMS accession number (ML003753154) for

the draft report both on the report's cover page and in the transmittal letter. This practice is

currently not the norm, but it makes document retrieval much easier. Recommendation:

Encourage the rest of the NRC staff to adopt this extremely useful practice.

RESOLUTION: No change required. The suggestion has been made as requested.

UCS COMMENT 6: The paragraph immediately preceding Table 6 on page 14 reports that the

MTC values summarized in the table came from References 17, 18, and 19. It appears that the

information really came from References 16, 17, and 18 instead. Recommendation: Check the

proper references and revise the report if necessary.

RESOLUTION: The report was revised as indicated in the comment.

C-5
GENERAL ELECTRIC provided the following comments in a letter to the NRC dated October 30,

2000.

GE COMMENT 1: GE has received the subject report and completed a review of the document. In

this report, the NRC states that ATWS risk is comprised of three elements: a) frequency of scrams,

b) reliability of the reactor protection system (RPS), and c) reliability of ATWS mitigation systems.

These items are discussed separately below:

a. The NRC recognizes that the scram frequency has come down by a factor of ten since the

ATWS rule was issued and this by itself has greatly reduced the risk associated with ATWS. GE

concurs with this conclusion.

RESOLUTION: No change required.

b. When the ATWS rule was issued, the NRC estimated that RPS reliability was about IE-5 per

demand, while GE estimated it was an order of magnitude better, about 1 E-6 per demand. At this

very low failure rate, it takes many years of data collection to demonstrate the failure rate is

accurate. With several ensuing years of additional operation without a major RPS failure in the

industry, the NRC has reduced their failure rate estimate, and though not as low as the GE estimate,

there is a smaller difference between the GE and NRC estimates.

RESOLUTION: Section 2.0 indicates the wide range of RPS unreliability (3.OE-06 to 1.1 E-04)

before the ATWS rule. Section 3.2.2 is revised to "Table B-3 shows that the outcomes of upper

and lower bounds of unreliability range from 1 .8E-06 to 5.7E-05 and this range is smaller than

that noted Section 2.0 before the ATWS rule was issued." Section 3.2.3 and the report

conclusions makes the remaining GE point by discussing that the values of RPS unreliability

illustrates the difficulty of estimating reliability values in highly reliable systems.

c. Reliability of mitigation systems is dominated by short-term operator action reliability for ATWS.

The report states, ... examinations for BWRs indicate large variations in the assumptions for

reliability of human actions in response to an ATWS. Similarities in design, procedures, and training

argue against such variability. Consequently, some BWR risk analysis may underestimate the risk

of ATWS.0 Operator action reliability has long been open to uncertainty and disagreement. In the

absence of concrete specifications, different assessments are likely to have different assumed

values. However, it is clear that improvements made in design, procedures, and training since the

ATWS rule was issued have all contributed to improving operator action reliability. GE does not

necessarily agree that BWR risk analysis may underestimate the risk of ATWS."

RESOLUTION: The report conclusion was revised to delete BWR risk analysis may

underestimate the risk of ATWSO and add "Risk analyses in support of licensing actions would

need to be justified on a more consistent basis. The original generic BWR risk analysis event

trees used to develop the ATWS rule expectations were based on the completed implementation

of the ATWS rule. For BWRs, that included ARI, RPT, and improved SLC. Most improvements

in operating procedures and training had been implemented earlier as a result of the TMI and

responses to the Brown's Ferry partial ATWS. To maintain consistency with the original

approach, the same operator response assumptions were used in this assessment.

C-6
GE COMMENT 2: The Background section of the report states that the Commission designated

ATWS as Unresolved Safety Issue (USI) A-9. It might be beneficial to the reader to state how and

when the USI was resolved, otherwise, one might assume that USI A-9 was still open.

RESOLUTION: Section 2.0 was revised after the discussion of USI A-9 to add "USI A-9 was

resolved with the publication of the ATWS rule."

GE COMMENT 3: Section 2.1.1 of the draft NRC report discusses the basis and application of the

200 F suppression pool temperature limit.

(a) In NEDO-30832-A, "Elimination of Limit on BWR Suppression Pool Temperature for SRV

Discharge with Quenchers," May 1995, the NRC accepted elimination of the 200 F local pool

temperature limit for T-type and X-type SRV discharge quenchers. This is based on the fact that as

suppression pool temperature approaches saturation, condensation loads decrease. Therefore, the

statement, "As the suppression pool temperature increases, the potential for unstable condensation

of the discharge to the pool may overload the containment structure, " is not factual.

RESOLUTION: Section 2.1.1 revised to read:

win some cases, the NRC staff has accepted higher temperature limits based on plant specific

design features. As the suppression pool temperature increases to near the boiling point, steam

condensation is less effective, resulting in higher containment pressure; and the RHR pumps

would eventually fail due to inadequate net positive suction pressure. During an unmitigated

ATWS, with continued heat input to the suppression pool, the containment could fail while the

core was still intact."

(b) The statement, During an ATWS at a BWR, the containment would probably fail prior to core

damage, is true, but should be clarified to note that failure only is postulated after a long series of

other failures of ATWS mitigation features such as ARI, RPT, boron injection, manual insertion of

control rods, and additional failures in containment mitigation features such as pool cooling and

containment pressure relief

RESOLUTION: See resolution to GE Comment 3(a) above.

(c) As discussed above, the 200 F local pool temperature limit has been removed for plants which

meet the requirements of NEDO-30832-A. Therefore, plant specific limits are used as the limiting

suppression pool temperature for ATWS. It is suggested that "the 200 F limit' be replaced with

"the plant-specific pool temperature limit for ATWS evaluations."

RESOLUTION: See resolution to GE Comment 3(a) above.

C-7
REGION 2 provided comments in a note dated November 30, 2000.

REGION 2 COMMENT: The report was well organized, and discussed the major cost and impact

issues associated with the NRC's ATWS rule. Risk information was included, but was primarily

based on the work performed in 1983 to support the original paper recommending the rule.

Information from the licensee's IPE submittals was also included. Both of these sources of risk

information are already dated, and many licensees are using more recent information in their current

risk models. Section 3.2 could be expanded to also include a new subsection to show the new risk

estimates, event trees and associated fault trees for ATWS that have been developed for the NRCs

Standardized Plant Analysis Risk (SPAR) Models for RES. Revision 3 SPAR models are already

available for many plants, and the ATWS risk is available in the solution to these models. The event

trees reflect the current owners group guidelines for coping with ATWS, and would be more

accurate than the trees used in the original 1983 work. The old event trees should stay in their own

section in the report, as they were used as the basis for comparing the current risk to the projected

risk improvement. The updated event trees and fault trees would provide a new baseline to judge

future proposed improvements to the rule.

With the exception of the recommended improvement noted above, Region 2 has no comments.

The report is well written, and conveys the necessary information in a concise manner.

RESOLUTION: No change required. Given that the goal of this assessment was to determine if

real changes in plant and equipment performance were achieved with the ATWS rule, rather

than model improvements and changed assumptions, the simplified event trees provide a

consistent and reasonable approach for this assessment.

As a matter of interest, we looked at the SPAR models (which are not yet approved) for ATWS

and they approximate those used for the ATWS rule development except for the operator action

to scram the reactor. The HEP for operator action to scram the reactor was not significantly

different that the information already in the ATWS report from the RPS reliability studies.

C-8
NRC, Office of Nuclear Reactor Regulation, Division of Systems Safety Analysis, Reactor

Systems Branch (NRRIDSSAISRXB) provided the following comments in a note NRC dated

December 8, 2000.

NRRIDSSAISRXB COMMENT 1: The staff of SRXB considers the approach taken by the Office of

Research (RES) to assess the "effectiveness" of the ATWS Rule to be very reasonable and logical.

The Background" provides the reader with an excellent history of the ATWS issue, which is

important for understanding the complex regulatory history.

RESOLUTION: No response required.

NRRIDSSAISRXB COMMENT 2: The "Assessmenr' Section is accurate and complete.

Subsection 3.2.6 deals with the subject of fuel management. DSSA believes that a change in fuel

management has a significant effect on ATWS. Fuel management is at the heart of the ATWS

issue. When a vendor or a licensee designs a core for a particular cycle, fuel enrichment, burnable

absorber, and boron concentration are parameters that play a major role in determining the length of

the fuel cycle and such Technical Specifications (TS) as those associated with the shutdown margin.

The interplay between the cycle length and the shutdown margin is a crucial one. For longer fuel

cycles (say 18 months instead of 12 months) with higher enriched fuel, increased boron

concentration or fuel with integral burnable absorber are necessary to maintain the shutdown margin

to a TS value at beginning of cycle (BOC). However, fabrication of fuel with integral burnable

absorber is very costly, and it is far more expensive than increasing the boron concentration.

Should a vendor or licensee substitute soluble boron for the burnable absorber, the increased boron

concentration will make the moderator coefficient become less negative, and under certain

circumstances even positive.

RESOLUTION: No change required.

NRR/DSSAISRXB COMMENT 3: The "Conclusion3 Section is very well written in terms of clarity

and decisiveness, and provides the reader with unbiased results. The staff is in agreement with the

RES conclusion that the ATWS Rule has been effective in implementing and modifying system

hardware at a reasonable cost while reducing the occurrence of an ATWS event.

RESOLUTION: No change required

NRRIDSSASRXB COMMENT 4: The staff has no recommendations for additional reviews of

regulatory effectiveness.

RESOLUTION: No change required

NRRIDSSAISRXB COMMENT 5: Finally, DSSA notes that recent submittals related to ATWS

depart from the traditional bases supporting the rule making. Specifically, licensees or vendors are

proposing to rely more on operator performance during unfavorable exposure time (UET) as the

primary means for ATWS mitigation. The RES effectiveness study does not speak to the issues or

concerns regarding these approaches as it relates to the technical bases of the present rule. It is

our view that the study should be modified to discuss the impacts of the new industry strategies.

C-9
RESOLUTION: The impacts of new industry strategies regarding the UET are briefly discussed

in 3.2.6; further review of new industry strategies regard ATWS are beyond the scope of the

report as explained in DSSAPSAB Comment 4.

The technical basis regarding MTC is addressed in Section 2.1.2 and 3.2.6. The report was

revised regrading the technical bases of the present rule regarding operator performance as

follows:

"2.1.3 Operator Action

In developing the ATWS rule, the NRC staff view was that (1) operator action should not be

relied upon during the first ten minutes of an accident including a manual scram and (2) operator

actions should be relied upon later in the course of an accident if the condition in the reactor and

mitigating systems is available to the operator, that sufficient time is available to assess the

condition and take action, and that the operator has been trained in the action.

In practice, operators are trained to identify an A TWS condition, initiate a manual scram,

manually insert the control rods, manually trip the RPS motor-generators, and begin borating.

After the Salem events and before the ATWS rule was issued, there was discussion in

NUREG-1000, "Generic Implications of A7WS Events at the Salem Nuclear Power Plant", April

1983, regarding credit for operator action to scram the reactor following an ATWS, since the

Salem operator scrammed the reactor30 seconds afterthe RTS failed. NUREG-1000 stated

that the Salem ATWS events were not as serious as they could have been as they took place at

low power, the operators were at the correct control panel to support power ascension, and there

was sufficient relief capacity and steam generator mass.

Operator recognition that an ATWS condition exists can be difficult, especially for events

involving malfunctions of the reactor trip system which do not activate the trip indicator lights.

The fast moving ATWS event does not provide much time for the operator to recognize what has

occurred. For example, NUREG- 1000 provides peak pressure versus time curves that show that

if the operator takes no action following an ATWS, the ASME service level C stress would be

exceeded in 47, 70, and 100 seconds for CE, B& W, and W PWRs, respectively. In addition, the

pressure ramps up from a normal operating values of 2250 psia to its peaks of approximately

3500, 4000, and 3500 psia in 47,20, and 10 seconds for CE, B&W, and PWRs, respectively."

NRRIRES Coordinator provided comments on December 8, 2000, based on the Reactor Systems

Branch (SPXB) comments.

COMMENT 1. Sub-section 3.2.6 deals with the subject of fuel management, and states that

changes in fuel management "may" affect the ATWS mitigating capability. Fuel management is at

the heart of the ATWS issue, and we recommend that "may be replaced with a stronger word, such

as will orcan.0

RESOLUTION: No change required. Since some of the material relates to proposed industry

methods we do not wish to use wording that would pre-disposition staff reviews.

COMMENT 2. In regard to the risk associated with the ATWS event, the study should identify

concemsl issues associated with current Vendors suggestions to substitute man-made efforts (such

as plant operational procedures) for natural controls such as negative moderator coefficient.

C-10
RESOLUTION: See resolution to DSSASPXB Comment 5.

NRC/Office of Nuclear Reactor Regulation/Division of Systems Safety and

Analysis/Probabilistic Safety Assessment Branch (NRRIDSSAIPSAB) provided the following

comments in a note dated December 11, 2000.

NRRIDSSAIPSAB GENERAL COMMENT: Overall, the draft report Regulatory Effectiveness of

the Anticipated Transient Without Scram Rulem did a good job in demonstrating that the Anticipated

Transient Without Scram (ATWS) rule was effective in making hardware modifications and imposing

operating limits, reducing ATWS risk, and implementing the ATWS rule at reasonable cost. In

particular, the sections on boiling water reactor (BWR) and pressurized water reactor (PWR) ATWS

sequence and event trees provided a thorough discussion of how the ATWS rule evaluated the

expected core damage frequency (CDF) of an unmitigated ATWS. The value-impact analysis also

provided a good discussion of how the ATWS rule resulted in savings to the nuclear industry due to

a reduction in the number of automatic scrams.

RESOLUTION: No change required

NRR/DSSA/PSAB COMMENT 1: A few terms are used in the executive summary that are not

defined until the main report. It would be helpful to the reader if these terms were defined in the

executive summary. In particular, the term Unfavorable Exposure Time" (UET) is first used on

page ix, but not defined until page 4, section 2.1.2 of the main report. Similarly, P(ATWS) is

first used on page x, but not defined until page 2 of the main report.

RESOLUTION: The UET concept was deleted on from the report (See response to the

resolution to WOG General Comment 1). P(ATWS) was defined in the executive summary

consistent with its definition in the report.

NRR/DSSAPSAB COMMENT 2: Some terms and phrases need to be defined and/or

explained when they are first used. Specifically, ASME Service Level C on page 4 needs to

be explained (i.e., what is ASME Service Level C and how does that impact the ATWS

analysis?) and the acronym SBO on page 6 needs to be defined.

RESOLUTION: The referenced sentence was revised to read in SECY-83-293, the ASME

Service Level C pressure of 3200 psig was assumed be to an unacceptable plant condition

during ATWS rule development."

NRR/DSSA1PSAB COMMENT 3: Were any other reports issued regarding the ATWS rule? It

may be useful to contrast the results of this study to the results of other studies.

RESOLUTION: There were no other ATWS related reports with the same scope. Section 3.1

discusses the use of several ATWS related reports (references 9 through 19) that were used for

this assessment.

NRR/DSSAIPSAB COMMENT 4: The second numbered item on page 13 of Section 3.2.5 is a

very important finding that needs to be highlighted more. New phenomena, resulting from

using higher fuel bum-up, that may affect the operability and reliability of the reactor protection

system (RPS) (i.e., control rod insertion) needs to be investigated further and considered by the

staff in reviewing related licensee submittals and/or in discussions with the owners groups

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(e.g.,WOG). This observation needs to be presented in the executive summary and the

conclusions, Section 4, to ensure it is properly raised as potentially eroding the effectiveness of

the ATWS rule.

RESOLUTION: The conclusions of the report were revised to add:

"The higher fuel bumup has resulted in previously unpredicted oxide growth and fuel assembly

distortion. In some cases this has resulted in slow or incomplete control rod insertion. These

failures and degradations are new phenomena and were not considered during the development

of the ATWS rule."

NRRtDSSAPSAB COMMENT 5: It should also be pointed out that the individual issues

identified in the executive summary and conclusions, Section 4, that could erode the

effectiveness of the ATWS rule are not independent of each other, but rather, have synergistic

effects. For example, the use of higher burn-up fuel to achieve longer cycles, may be creating

new phenomena that would impact the reliability of the RPS, which typically drives the low

values and contributions of ATWS sequences due to its postulated extremely high reliability

though it has a large uncertainty associated with it (this new phenomena would tend to increase

the uncertainty or at least skew it towards higher unreliability values), and would also increase

the unfavorable exposure time (UET) and possibly increase the peak pressure reached during

an ATWS event. The report needs to conclude that the identified issues, especially those

associated with the use of higher bumup fuel, needs to be carefully watched, considered, and

investigated by the staff.

RESOLUTION: See response to NRR/DSSAPSAB 4 and WOG General Comment 4.

NRRIDSSAIPSAB COMMENT 6: On page 15, the first paragraph states that the WOG met

with the NRC staff on August 23, 2000, to obtain NRC concurrence that the WOG approach to

ATWS for licensing issues, such as MTC, is acceptable ..." At this meeting, it was made clear

to the WOG that the NRC staff at the meeting would not be concurring or approving their

approach, but rather, the meeting allowed the WOG to present how they were addressing

issues identified during a 1998 meeting with the staff and allowed the WOG and staff to discuss

(and refine) these issues further. Discussions with the WOG regarding their proposed

approach and its implications will continue. The RES staff involved in the ATWS rule's

effectiveness study should remain engaged and provide input in these discussions and with the

other staff addressing these issues (e.g., NRR SRXB and SPSB staff).

RESOLUTION: The report was revised to delete "On August 23, 2000....by the staff."

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Westinghouse Owners Group (WOG) provided the following general and detailed comments

in a letter to the NRC dated December 20, 2000.

CLARIFICATION: WOG provided comments that address traditional ATWS rule issues that

are the subject of the report. The WOG also provided comments presenting material that

potentially supports a new licensing basis for ATWS and this material is not within the scope of

the report. Some of the new material is supported by WCAP-1 1992 that was submitted to the

staff in May 1995, and found not to be acceptable for use in licensing or other regulator

matters, and that the WOG and the NRC continue to work to address the concerns as

mentioned in Section 3.2.6 of the report. WOG comments not within the scope of this report

will state see the Clarification."

WOG GENERAL COMMENT 1: From a WestinghouselWOG perspective, the term

Unfavorable Exposure Time' or UET represents the duration of a given fuel cycle, for a

specific plant configuration, in which the total core reactivity feedback is insufficient to preclude

exceeding a peak RCS pressure of 3200 psig following an ATWS event. UET was

defined by the Westinghouse Owners Group (WOG) for use in a more detailed ATWS PRA

model developed as part of the "Westinghouse / WOG ATWS Rule Administration Process,"

WCAP-1 1992. The concept of UET is also being applied in a revised Risk-Informed ATWS

PRA model supporting an ongoing WOG ATWS PRA program.

To determine the UET values used in this ATWS PRA model, the reactivity feedback required

to just yield a peak RCS pressure of 3200 psig is first determined by specific ATWS transient

analyses. The reactivity feedback conditions for a given reload core model are then compared

to the transient reactivity feedback models to determine the value of UET for a given plant

configuration. For the Westinghouse / WOG ATWS PRA model, a total of 12 UET values are

typically determined and used.

The term UET as defined and applied above was not a term directly used in the basis of the

Final A1WS Rule as documented in SECY-83-293. In SECY-83-293, the terms "favorable

MTC" and unfavorable MTC are applied in the discussion of the simplified ATWS PRA model.

These terms are not the same as UET defined above. Hence, while Westinghouse and the

WOG concur with and support the industry's use and correct application of this term in ATWS

related discussions and documentation, it is requested that the use of the term UET in this Draft

Report be replaced with "unfavorable MTC" to be consistent with the reference to

SECY-83-293.

RESOLUTION: The report was revised to delete use of the term UET and use unfavorable or

favorable MTC, except in Section 3.2.6 which acknowledges WCAP-1 1992 and the UET

concept. See the Clarification above.

WOG GENERAL COMMENT 2: For Westinghouse PWRs, the discussions of ATWS events

as they relate to RCS pressure, unfavorable MTC, and ATWS mitigation should be clearly

defined and limited to only anticipated transients that result in a direct or consequential loss of

main feedwater. The only ATWS events that potentially lead to high RCS pressures that

challenge the ASME Service Level C Stress Limit (i.e., 3200 psig) are those events that result

in a direct (i.e., Loss or Normal Feedwater ATWS) or consequential (i.e., Loss of Load ATWS)

loss of main feedwater. The AMSAC system required by the Final ATWS Rule and installed at

Westinghouse PWRs is designed to actuate a turbine trip and initiate the AFWS based only on

detection of a condition representative of a loss of main feedwater. Other non-loss of feedwater

C-13
related anticipated transients considered in the generic ATWS analyses supporting the basis of

the Final ATWS Rule do not lead to RCS overpressure concerns and, hence, have no ATWS

related mitigation Systems.

RESOLUTION: Not just Westinghouse PWRs, but all PWRs are subject to high pressures

during an ATWS only if they result in direct or consequential loss of main feedwater, since steam

generator relief capacity is generally capable of relieving 100 percent steam flow. However, the

following considerations apply:

First of all, given there is no reactor trip, no turbine trip, and no loss of feedwater, but there is

some condition requiring a reactor scram (hence, an ATWS), then the operator may be unlikely

to recognize that an ATWS condition exists quickly enough to take action. There may be no

pressure spike, but core damage from departure from nucleate boiling or other cause.

Secondly, for the condition without loss of feedwater, reactor power would remain high since

there would be no negative reactivity feedback to reduce reactor power. Then, if the turbine

were tripped, condensate would be depleted through the steam generator safeties and the main

feedwater system would eventually trip on loss of suction pressure. Thus, even with no

immediate loss of feedwater, condensate is eventually depleted out the steam generator

safeties; and then the steam generators would dry-out, only on a longer time scale than two

minutes.

Thirdly, the turbine will almost always trip on loss of main feedwater; and conversely, the main

feedwater system will usually not trip on turbine trip. This logic was part of the UMTC

Overpressure" factor where turbine trip is the lower risk scenario.

The point is that given an ATWS with main feedwater operating doesn't assure that the outcome

will be benign. Complex operator actions are still required to achieve safe shutdown.

Section 3.1 was revised to add:

The ATWS rule original risk expectations were calculated assuming that the ATWS rule

requirements were implemented. Thus, the original analysis assumed that AMSAC, DSS, ARI,

RPT trip, etc., were installed and operable.

"The original generic risk analysis event trees did not distinguish those initiating events

accompanied by loss of main feedwater. And again, for consistency, neither did this

assessment, since the goal of this assessment was to determine if the ATWS rule resulted in

improved performance, rather than improved modeling." However, for W plants, the likelihood

of loss of feedwater events was a consideration in adoption of the values of the -MTC

Overpressure" factor and separate event trees for turbine trip and non-turbine tp scenarios."

WOG GENERAL COMMENT 3: Drawing conclusions based on the continued use of the

SECY-83-293 PRA approach is inappropriate and may result in misleading conclusions and

recommendations. The SECY-83-293 model is too simplistic to assess the risk from ATWS

events and does not account for a number of important considerations. One such

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consideration is the probability that the plant is operating in a configuration that corresponds to

the UET (i.e., unfavorable MTC) continually cited in this report. The conditions for this

referenced UET are no rod insertion, all auxiliary feedwater available, and no PORVs blocked,

but this could be a low probability configuration.

RESOLUTION: Section 3.1 was revised to read:

"For PWRs, the original generic risk analysis and this assessment assume that the following

equipment is available: emergency feedwater, primary power-operated relief valves and

safeties, and secondary power-operated relief valves and safeties - this tends toward lower peak

pressures and lower A TWS risk. On the other hand, the original generic risk analysis assumes

that the rod control system is in manual control - this tends toward higher peak pressures and

higher ATWS risk Since the affects are offsetting, the total impact may be minimal. Again, to

determine whether the ATWS rule was effective in reducing risk from ATWS, we looked to see if

it met or exceeded the goal based on the original method of calculation."

"The original generic event trees were developed after much deliberation to focus on what were

the most important sequences. The deliberations which went into determination of the type of

event trees to use considered aspects of the ATWS transient such as the status of the main

feedwater system. The current assessment was not meant to be a total reevaluation of the

approach to ATWS risk analysis, but an attempt to determine f the ATWS rule goals were met.

In order to make a valid comparison of the improvements due to less frequent initiating events

and improved equipment performance rather than changes in the risk model or assumptions, the

simplified event tree approach was used. Additionally, we believe that this approach provides a

reasonable estimate of ATWS risk."

WOG GENERAL COMMENT 4: The draft report frequently specifies and refers to a 1 % UET

(i.e., unfavorable MTC) as being applicable to Westinghouse PWRs. This is incorrect. A 1%

unfavorable MTC was modeled in one of the SECY-83-293 simplified PRA models (Figure 11 in

SECY-83-293) that represented what Is called turbine trip transients in the PRA evaluation. As

discussed in SECY-83-293, Section 5.5, Item 2 for Westinghouse plants, a 1% unfavorable

MTC was assumed to exist for this turbine trip event since the pressure transient for this type

ATWS is relatively mild. As earlier discussed in the same section of SECY-83-293, for

non-turbine trip events that challenge the ASME Service Level C limit, a 10% unfavorable MTC

was assumed (Figure 12 in SECY-83-293).

Neither of these unfavorable MTC values is considered to be applicable for Westinghouse

PWRs. The Westinghouse generic ATWS analyses performed in response to NUREG-0460

and which form the deterministic analysis basis for the Final ATWS Rule are documented In

Westinghouse letter NS-TMA-2182, ATWS Submittal," dated December 30, 1979. In these

analyses, a full power MTC of -8 pcmP/F was used in the analyses and a sensitivity to a change

in MTC to -7 pcrn/F was included. In 1979, these MTCs represented the values of MTC that

Westinghouse PWRs would be more negative than for 95% and 99% of the cycle, respectively,

as specified by NUREG-0460. The peak RCS pressures using these MTC values (without

changes in any other assumptions) are shown in NS-TMA-2182 to be less than 3200 psig.

RESOLUTION: With regard to the first paragraph of the comment, see resolution to WOG

General Comment 1 above which clarifies the use of the 1, 5, and 10 percent unfavorable MTCs.

C-15
Also, peak pressures are in excess of 3200 psia for the MTCs noted in the comment. Section

3.2.6 of the report was revised as follows:

Section 2.1.2 revised to read:

"At the time of the ATWS rulemaking, the estimated impact of periods of "Unfavorable MTC" was

included in the event tree as the "MTC Overpressure" factor. For Westinghouse plants, the

original ATWS expected risk approach estimated an UMTC Overpressure" factor of 0.01 for

turbine trip scenarios and 0.1 for non-turbine trip scenarios; and the expected outcome of the

ATWS rule with installation of AMSAC assumed that Westinghouse plants would be in the

turbine trip scenario 70 percent of the time (i.e., 70 percent of the ATWS transients would be the

lower risk scenario). As stated above, the turbine trip provides the operator additional time to

respond.

For CEIB& W plants, a higher "MTC Overpressure" factor of 0.5 was used regardless of turbine

trip status. The reason was that the MTC was expected to be insufficiently negative to limit peak

pressure below 3200 psig for up to 50 percent of the cycle. For CE/B& W plants, the expected

outcome of the ATWS rule included a factor of 10 improvement in the electrical portion of the

scram system based on installation of the DSS. These same assumptions were used to

determine the current outcomes.

Like the BWR risk analysis, this risk model does not include low frequency "anticipated

operational occurrences" such as loss of primary system flow for which core damage occurs,

regardless of the MTC or other factors, unless the rods insert immediately."

Section 3.2.6 was revised to read:

"Since the ATWS rule was developed, fuel cycle lengths have been extended. This requires

higher enrichment fuel and often results in less negative (more positive) MTCs. As previously

discussed in Section 2.1.2....

Pressures for economic efficiency are also prompting licensees to increase the power of the

reactor. The higherpower rating leads to faster diy-out of the steam generator for PWRs and

faster heatup of the suppression pool for BWRs. The combination of less negative MTC and

higher reactor power has a greater effect on the plant response to ATWS than either one alone.

Less negative MTC and higher power result in the ATWS pressure spike occurring earlier,

allowing less time for operator actions, quicker, less time for MTC to limit the pressure increase,

and higher, increasing the likelihood of equipment damage. These fuel management changes

potentially impact both the deterministic and probabilistic analyses for ATWS and other reactor

transients.

In 1987.... no significant changes were required in ATWS analyses. The peak pressure of

3497 psia update in Table 6 for W was calculated based on values obtained from sensitivity

analysis in NUREG-460. The sensitivity analysis shows a pressure increase of approximately

100 psi for a change in MTC of 1 pcm/ F. Thus, the change from the 1979 value of -8 pcm/ F

to -5pcm/ F in Table 6 corresponds to a 300 psi increase.

C-16
Table 6 ATWS MTC and Peak Pressure for PWRs

Parameters 1979 1988 1994 NRC Technical Specification

ATWS Analysis Update of ATWS Analysis Survey

CE MTC -2.0 to -6.8 -2.6 to -5.7 0 to +3 above 70% power

Peak 4290 psia 4153 psia

Pressure

B&W MTC -10.5 18monthcycle: -11.0 O above 95% power

24 month cycle: -4.3

Peak 3464 psia 3764psia

Pressure 18 month cycle: > 3200

24 month cycle: > 3200

W MTC -8.0 -8.0 average inear to 0 from 70% to 100% power

range -5 to -15 One plant at 2 at 100% power

Peak l97 psia 3497 psla (-5pcr/F) Pressure III

Table 6 also summarizes ...1979 and 1988 MTC values. Based on NUREG-460 sensitivity

analysis that indicates a 1 pcm/ F less negative (more positive) MTC increases the RCS peak

pressure approximately 100 psi calculations based on the limiting TS MTC at full power could

lead to higher peak ATWS pressures and longer unfavorable MTCs. CE and B& W reactors

installed the DSS to counteract risk andpeak pressure effects of unfavorable MTC. The B&W

and CE DSS the reactor trips independent of the RPS, improving RPS electrical reliability by a

factor of 10. The DSS trips the reactor at approximately 2450 psia to prevent the RCS pressure

following an ATWS from reaching the unacceptable condition-3200 psia ASME service level C

limit."

Westinghouse WCAP-1 1992.... staff in May, 1995. WCAP-11992 introduced the concept of an

unfavorable exposure time (UET) which is the percentage of the fuel cycle time the pressure is in

excess of 3200 psia; in this respect UET is fundamentally the same as the unfavorable MTC

percentage used in the ATWS rule development. However the UETis calculated differently than

the unfavorable MTC percentage and in this respect they are different. ..

Higher peak pressures...ATWS rule. if the W risk assessment in SECY 83-293 were done

using the same method as B& Wand CE with an "unfavorable MTC" changing from 1 percent to

40 percent of the fuel cycle, the A1WS risk would increase by a factor of about 20 to 4.3E-6&...

WOG GENERAL COMMENT 5: Operator actions to trip the plant are an important part of

protecting the reactor and need to be considered in this report. The reactor protection system is a

highly reliable system that is backed up by operator actions to trip the reactor. Plant operators are

highly trained in this action and there is a high probability of success. Such operator actions need to

be considered when drawing conclusions related the reactor protection system reliability and the risk

from ATWS events. In fact, in the one ATWS event for a Westinghouse PWR (i.e., Salem), the

reactor was tripped manually within 30 seconds after the reactor trip breakers did not open when

demanded by the solid-state protection system.

C-17
RESOLUTION: See resolution to NRR/DSSS/PSAB Comment 4. Section 3.1 was revised to

read:

"Operator action to scram the plant was not credited in the original risk estimates of the ATWS

nle. To be consistent, the current assessment used the same approach. The goal of this

assessment was to compare the AT//S rule expectations and outcomes. Consequently, as long

as operator performance has not changed greatly since the ATWS rule was enacted or the

impact of including operator action to scram is small, this approach is valid.

First, regarding changes in operator performance to scram the plant which have occurred since

the ATWS rule was implemented, there is no basis to support major improvements. At the

time of the ATWS rule, operating procedures and training had already been greatly improved

following the TMI accident; and the Salem ATWS event implications had been included as

previously discussed. The implicit pressure on operators to minimize the number of scrams

may not have been as strong as it is now.

Secondly, recent work supports the position that the impact of considering operator action is

relatively small "Reliability Study: Westinghouse Reactor Protection System, 1984 1995m

NUREG/CR-5500 Volume 2, shows values of RPS unavailability of 2. 1E-5 without operator

action and 4.9E-6 with operator action - the operator provides reduced risk by a factor of

about 4. This improvement in unavailability assumes that the operator initiates a manual

scram 99 percent of the time when a scram signal is present and 50 percent of the time even

when a scram signal is not present. In other words, these values already assume excellent

operatorperformance. The reliability study did not include the potential for operator action to

manually trip the RPS motor generator sets. The motor generator breaker controls are not

always accessible from the control room, which would negate the possibility of timely operator

action.

Thirdly, the initiating event frequency is assumed to be the rate of automatic scrams. Since the

time of the ATWS rule, the fraction of all scrams which are automatic has decreased and the

fraction of operator initiated scrams has increased. Manual scrams were not included in the

current A1WS initiating event frequency; their inclusion in initiating event frequency would

about double the initiating event frequency."

WOG GENERAL COMMENT 6: In several locations within this draft report, it is stated that B&W

and CE reactors are required by the ATWS Rule to have DSS as compensatory measures for

higher UETs and Westinghouse reactors do not have this requirement due to a 1% UET. For this

comment, the discussion above for General Comments I and 4 first apply. Furthermore, the

concept that B&W and CE require DSS due to higher UETs, i.e., higher unfavorable MTC, is

misleading. In the 1979 baseline ATWS analyses, all NSSS vendor were subject to the use of the

same MTC specifications as given in NUREG-0460 (see General Comment No.4 above). The fact

of the matter is that the pressure relief capacity of Westinghouse PWRs is significantly higher, and

thus, significantly more favorable than that for the B&W and CE PWRs. The results of the generic

deterministic ATWS analyses for Westinghouse PWRs showed a peak RCS pressure of less than

3200 psig when using the base case MTC of -8 pcmIOF as specified by NUREG-0460. The peak

RCS pressures for the B&W and CE PWRs were significantly higher. As indicated in Table 6 of this

draft report, the 1979 baseline peak RCS pressures for B&W and CE PWRs were 3464 psia and

4290 psia, respectively. It is recognized in SECY-83-293 that with the 1979 baseline MTCs for the

B&W and CE plants, a peak pressure in excess of 3200 psig results 50% of the time. Hence, DSS

was required for these plants since a change to more favorable MTC conditions was not a truly

C-18
viable option from a plant Operation versus reactor physics perspective. It should also be noted that

on December 3, 1984, the Commission decided not to issue the proposed rule on ATWS that would

have required a diverse scrarn system for Westinghouse plants.

RESOLUTION: See resolution to WOG General Comments 1 to 4.

The resolution to WOG General Comment 4 clarifies how the CE and B&W DSS provides

compensation and that the Westinghouse peak pressures is in excess of 3200 psia for MTC of -

8pcml°F.

To the contrary, the relief capacity of the CE -3800 plants is larger than the W plants.

Section 3.2.1 and Table 2 were revised to acknowledge that December 3, 1984, Commission

decision not to issue a DSS rule for W plants.

WOG GENERAL COMMENT 7: Portions of this report and in particular, Section 3.2.5, include

discussion on partial rod insertion events. The original concern identified and associated with

ATWS is the potential for a common-mode failure of the reactor trip system following anticipated

transients that rely on a reactor trip for event mitigation. For such a condition, the real concern is

the potential for no automatic or timely rod insertion. In comparison with this potential scenario, the

occurrence and consequences of the partial rod insertion events specified in TableS are far from

being comparable to a common-mode failure of the entire reactor trip system. All plants' Technical

Specifications have reactor trip scram time requirements and associated testing and surveillance

requirements. The purposes of these tests are to specifically confirm reactor trip system and rod

insertion characteristics to those assumed in plants' licensing basis safety analyses. The detection

of slower than required rod insertion times as a result of these tests indicates a problem. Such

problems are appropriately identified and addressed. In addition, only a portion of the total available

rods is actually needed to obtain reactor control and shutdown. Hence, such instances of slow or

partial rod insertion as listed in Table 5 should not be considered and classified as a precursor to

ATWS events.

RESOLUTION: Slow or partial insertion of the rods are within the scope of the ATWS rule. An

ATWS is an anticipated operational occurrence followed by failure of the reactor trip portion of

the protection system. The reactor trip system function includes the reactivity control systems.

49FR26036 that issued the ATWS rule specifically states the reactor trip system consists of the

control rods and the control rod mechanisms.

Section 3.2.5 was revised to read: While the number of controls rods affected is only a portion

of the total available rods actually needed to obtain reactor control and shutdown, these failures

and degradations are new phenomena and were not considered during the development of the

AIWS rule. Although these conditions do not affect present ATWS analysis assumptions

directly, they cannot be dismissed as precursor events."

Also see NRR/DSSAIPSAB Comment 4 which takes a different view than the WOG.

WOG GENERAL COMMENT 8: The report summarized its conclusions In the following manner:

OThe ATWS Rule was effective in reducing ATWS risk and that the cost of implementing the rule

was reasonable. However, uncertainties in reactor protection system reliability and mitigative

capability may warrant further attention to ensure the expected levels of safety are maintained.nThis

assessment was based on several of factors for a PWR. One factor is that the RPS system has

C-19
been shown to be extremely reliable. This contributes to the inability to quantify the PRA numbers

for failure and was used in the justification for further attention. The inability to quantify the numbers

for RPS reliability due to the extremely reliable nature of the system is not justification for additional

rulemaking. Since the system is so reliable that the numbers can not be accurately quantified, this

is justification that rulemaking is not required. Another factor cited is that the cost of modification

implementation was less than expected due to reduction in the number of trips per year, which was

a recommendation of the rule. It is not appropriate to include industry efforts to improve reliability

and availability in the cost benefit of the rulemaking unless the modifications imposed by the rule

can be shown to have resulted in that improved reliability. Efforts to reduce the number of trips were

in place prior to the rulemaking. The actual cost of the modifications and their continuing costs for

maintenance are not effected by the efforts of the utility to improve availability times. The imposed

modifications clearly have not increased the availability times of any plant. Therefore, there is no

Oreductionin cost as a result of the rulemaking.

The executive summary of the draft itself appropriately frames the current industry situation in that

reliability requirements are so high and ATWS events are so rare that many years of operating

experience are needed to generate sufficient system demands to reduce current estimates of the

uncertainty. The frequency and number of industry events support that useful predictions can be

made regarding future occurrences. Increasing the size of the sample or the number of

observations may reduce the sampling error, but not any biases. In the absence of a large number

of observations, related knowledge of the subject and scientific judgment must be relied upon in

framing a course of action. Statistical significance or the absence of it by itself is not a rational

basis for additional action.

RESOLUTION: No change is required. The report does not call for additional rulemaking. The

original ATWS rule value-impact expectation estimated the costs for the ATWS rule

modifications and the annual replacement power due to reactor trips from spurious actuation of

protective features that were required by the ATWS rule. A like comparison found the actual

impact was less than expected due fewer spurious reactor trips from spurious actuation of

protective features that were required by the ATWS rule. The industry effort reduce the number

of trips was discussed but not included value-impact comparison of the expectations and

outcomes even though the Commission instrumental in formalizing the industry scram reduction

initiative. Also see resolution to the CE comment

The report concluded that some issues (based on operating experience) have the potential to

erode past achievements and that attention to these issues and regulatory actions that maintain

compliance with current regulations can assure the risk from ATWS remains acceptable; specific

actions were not mentioned as stated in the comment.

Three issues were identified including RPS reliability uncertainty, PWR moderator temperature

coefficient levels and BWR human error probability assumptions for operator action. Regarding

RPS reliability, the report indicates the ATWS rule evolved as a result of the statistical

uncertainty in RPS reliability due to low amounts of data and (three) ATWS events that

statistically should not have occurred. The report assessed the full range of statistics and

accompanying uncertainties provided by the inclusion and exclusion of failure data and the

application of ATWS rule and present day methodologies.

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Specific Comments

WOG COMMENT 1: Section 2, Third Paragraph. See General Comment No. 6 as it relates to

the last sentence of this paragraph.

RESOLUTION: See resolution to WOG General Comment 6.

WOG COMMENT 2: Section 2.1, First Paragraph. This paragraph defines three factors that impact

the likelihood of damage from an ATWS event. Left out of these factors is the ability of the operator

to take actions to trip the reactor if the RPS fails (see WOG General Comment 5). The operators

are trained to trip the reactor from the control room within the time available. The operators can also

take an action to trip the reactor by interrupting power from the motor-generators to the CRDMs (not

all plants can do this from the control room, so all plants do not credit this action in their PRA

model). It is important to include these operator actions since they are highly reliable and act as a

diverse mechanism for reactor trip.

RESOLUTION: See resolution to WOG General Comment 3 and 5.

WOG COMMENT 3: Section 2.1, Second Paragraph. The statement P(ATWS) was defined as the

annual frequency of an ATWS leading to plant conditions that exceed certain design parameters

that can result in core melt, containment failure, and the release of radioactivity is misleading. An

ATWS event that results in core damage does not necessarily continue to containment failure and a

release of radioactivity. Additional component failures are required, independent of the ATWS

event, for containment failure to occur. This should be clarified in the report.

RESOLUTION: P(ATWS) as defined in the report exactly as defined by the industry and

adopted by the NRC during ATWS rulemaking. See resolution to WOG Comment 4

WOG COMMENT 4: Section 2.1.2, Paragraphs 1 and 2. The last sentence In each paragraph is

For PWRs, it is likely that for an unmitigated ATWS, the core would melt prior to containment

failure. This doesn't seem to fit at the end of the first paragraph. In addition, please recognize that

exceeding the ASME Service Level C stress limit, i.e., 3200 psig, has conservatively been equated

to resulting in core damage for the purposes of the ATWS PRA. It is not clear what mechanism is

implicitly implied that would lead to containment failure. If the sentence is maintained, it is

recommended that the sentence be revised to read"... Prior to any potential for containment failure."

RESOLUTION: The report was revised to read For PWRs, it is likely that, for an unmitigated

ATWS, the core would melt with the containment intact."

WOG COMMENT 5: Section 2.1.2, Paragraph 2. The discussion here describing the ATWS

transient conditions only pertains to the condition with a loss of main feedwater. See General

Comment 2 above. Also, the water temperature increase described in the analysis can not

occur without the fuel temperature increase. Therefore, it is inappropriate to discount the fuel

temperature increase and thus the effects of Doppler. It is also inaccurate to state that the water

must heat prior to the fuel.

RESOLUTION: See resolution of WOG General Comment 2 on main feedwater. Regarding the

doppler coefficient, the fuel thermal time constant can delay the impact of the doppler coefficient

in a fast moving transient.

C-21
Section 2.1.2 revised to read:

... However, the fuel temperature coefficient only comes into play after the primary coolant

temperature has increased because of the delay caused by the thermal time constant of the fuel;

this can be an important factor in a fast moving transient such as ATWS. ...

WOG COMMENT 6: Section 2.1.2, Paragraph 4. A higher ASME service level was only

considered for B&W and CE plants due the higher peak RCS pressures for these plants.

RESOLUTION: The report was revised to read ....for B&W and CE plants a

WOG COMMENT 7: Section 2.1.2, Paragraph 5. The discussion in this paragraph regarding

positive MTC below 100% power is only applicable to plants that have requested and received

licensing amendments to permit operation under these conditions. Not all plants have licensed

these conditions. The third sentence in this paragraph states that ...all subsequent mitigative

functions are likely to be ineffective.' This is not totally accurate. Manual rod insertion is a

viable mitigative function that would he effective. Regarding the definition of unfavorable

exposure time" given in this paragraph, please see General Comment No. 1 above.

RESOLUTION: The discussion regarding the MTC below 100 percent is supported by plant

specific data in Appendix B, Table B-1, that shows most plants have a positive MTC below 100

percent power. See resolution to WOG General Comments 3 and 5 regarding operator action

and WOG General Comment regarding the UET.

WOG COMMENT 8: Section 2.2, Paragraph 2. The last sentence should read ... despite

perceived high reliability, ....

RESOLUTION: The sentence was revised as indicated.

WOG COMMENT 9: Section 2.2, Paragraph 4. The discussion in the last part of this paragraph

appears open ended. It states an objective for 1985 and states values for 1980 and 1983. Was the

1985 objective met? Page 17 states an average of 0.5 trips/reactor year since 1997.

RESOLUTION: No change required. The report clearly and closely states what the industry

stated in the Commission briefing on June 24, 1984. Deleting proportions of what the industry

said could imply this solely an NRC initiative. The industry did not meet the 1985 objective as

the average number of scrams increased to 5.28

WOG COMMENT 10: Section 3.1, Last paragraph. A review of the information in Appendix B

indicates the need for the last sentence in this paragraph. Comments received from WOG utility

representatives question the basis for some of the plant-specific numbers in Table B-1. It is

recommended that a reference basis be included in the table for the plant- specific values.

Otherwise, the validity of the conclusions drawn from the use of these values may be challenged.

RESOLUTION: No change required. The references for all also for the data used in the report

information are provided in Section 3.1 and the reference section of the report a bibliography for

each reference. In addition, the first sentence of the paragraph being discussed above explicitly

indicates where the data came from as follows: Appendix B gives plant-specific data on the

actual outcomes of the ATWS rule regarding modifications and data from probabilistic risk

assessment/individual plant examinations (PRA/IPEs). The data were collected from...."

C-22
WOG COMMENT 11: Section 3.2, Table I. The UET<1%" in the 'Expected Result block for

Westinghouse plants is not correct. See General Comment No. 4 above.

RESOLUTION: See resolution to WOG General Comment 1.

WOG COMMENT 12: Section 3.2.1, Second paragraph and Table 2. The second paragraph

states that Table 2 was prepared to show the degree of defense-in-depth provided by the

ATWS rule modifications that were intended to prevent an ATWS;...... It should also be noted

on this table that the operator could trip the plant via the reactor trip switch in the control room.

This provides a diverse means of reactor trip that is effective if the reactor trip signal failed due

to failure of the analog channels or components in the logic cabinets. It should also be noted

that at some plants the operators could also trip the reactor by interrupting power from the

motor-generators to the CRDMs. This action is also taken from the control room and is effective if

the reactor trip signal failed due to failure of the analog channels, logic cabinet components, or

reactor trip breakers. The UET<1 %' in the PWR Fuel Strategy block of Table 2 for Westinghouse

plants is not correct. See General Comment No. 4 above. Also note that a December 3, 1984,

letter from the Commission to Westinghouse clearly states that DSS is not required for

Westinghouse PWRs. Hence, the Current Outcomes' block of Table 2 for Westinghouse plants

should state DSS not required'.

RESOLUTION: See resolution to WOG General Comment 1, 3, 4, and 5. Table 2 was revised

delete to note DSS not proposed to date." The supporting text was revised accordingly to

include acknowledgment of the December 3, 1984, letter.

WOG COMMENT 13: Section 3.2.1, Third paragraph, Item (2). It is not obvious from the

information provided in Table 2 how it is can be concluded that DSS appears to provide a

compensation for higher UETs (up to 50%) for B&W and CE plants. Also, the discussion on 1%

UET for Westinghouse plants is incorrect.

RESOLUTION: See response to WOG General Comment 4.

WOG COMMENT 14: Section 3.2.2, Table 3. The Outcome' RPS reliability for Westinghouse

plants is given as 2.1 E-05. This does not account for the backup operator actions as discussed

in Comments 2 and 12. Including the operator action to trip the plant from the control room via

the reactor trip switch reduces this value to 5.5E-06 (NUREG/CR-5500, Vol.2, Reliability Study:

Westinghouse Reactor Protection System, 1984-1995', December 1998). This also does not

account for tripping the reactor via power interruption from the motor-generators to the CRDMs.

It is misleading to provide the reliability as 2.1 E-05 without stating this does not include credit for the

operator actions to trip the plant. Also, the paragraph following Table 3, second sentence, should

read Comparison of Table 3

RESOLUTION: See resolution to resolution to WOG General Comment 5. The second

sentence in the paragraph following Table 3 was revised as suggested.

WOG COMMENT 15: Section 3.2.2, First paragraph under Item #2. The statement 'These

numbers were developed using a fault tree model of the RPS system that may not include all

failure modes, a question of completeness for all PRA calculations' can be misleading. This

may be true for a complex system with little or no operating history, but there is a significant

C-23
amount of operating history on the RPS and operation of the RPS is well understood. It is

suggested that the phrase that may not include all failure modes' is either dropped from the

report, or a discussion of why this statement is applicable to the RPS evaluation is included in

the report. This discussion should address the quality of the fault trees and data used to

determine the RPS reliability, and also how the fault trees and data were developed. This

statement should make it clear that these fault trees were developed for an NRC program

assessing the reliability of the Westinghouse RPS by NRC contractors, and that their

development was done by personnel familiar with the RPS design and operation, and

considered the past operating experience of the RPS.

RESOLUTION: The phrase was removed.

WOG COMMENT 16: Section 3.2.2. Item #3. This indicates that the unreliability of the

mitigation systems for Westinghouse plants is controlled by the UET and that as the UET

increases the mitigation unreliability increases proportionally. This is not true. The 5% UET

referred to in the NRC report is associated with a plant operating configuration that will result in,

given that an ATWS event occurs, 100% (all) auxiliary feedwater available, relief from all

pressurizer safety valves and PORVs (no PORVs are blocked), and no control rod insertion.

The statement in the report does not account for the probability, which may be very low, of the

plant being in an operating configuration that results in these conditions. If a plant is operating with

the rod control system in automatic, no PORVs blocked, and all auxiliary feedwater

available, then the UET is 0, and significant increases to the UET for the first set of plant

conditions may have no impact on the LIFT for the second set of plant conditions. The

unreliability of the mitigation systems (as defined in this report) and the UET is dependent on the

plant operating configuration. The probability of being in each plant configuration needs to be taken

into account.

It is also noted that the values for mitigation unreliability provided in Table 3 refer only to the

situation when there is no credit for any control insertion, both PORVs are available, and all

auxiliary feedwater is available. In addition to this configuration, there are eleven other

configurations or conditions a plant can be operating in depending on the availability of the

PORVs, auxiliary feedwater systems, and the ability for a limited amount of control rod insertion.

There are UETs associated with each set of conditions. Referencing a UET for only one set of

conditions does not provide the complete picture. Also see General Comment No. 1.

RESOLUTION: See resolution to WOG General Comments 1 and 4, and the Clarification.

WOG COMMENT 17: Section 3.2.3, Paragraph 5. The statement 'The range of values of RPS

reliability illustrates the difficulty of estimating reliability values in highly reliable systems is true if

you want to base the system reliability on total system failures. Fault tree methods that break the

system down to the basic component level at which reliability data is collected is another approach

that can be used to assess the reliability of highly reliable Systems. The best reliability assessment

from this table is that based on NUREG/CR-5500 which uses the fault tree approach. This

approach is a well-established and acceptable method of determining the reliability of highly reliable

systems.

RESOLUTION: No change required. Section 3.2.3 including Table 4 explicitly provides the

NUREG/CR-5500 results along with other the methods each used in the ATWS rulemaking to be

consistent with the report methodology. We cannot revise the report to reflect Westinghouse

preferences or to label a methodology as the best.

C-24
WOG COMMENT 16: Section 3.2.3, Table 4. In the assessment of W RPS Unreliability'

corresponding to SECY-83-293 baseline update to 1995" it is not appropriate to include the one

RPS failure (Salem). Following the Salem RPS failure the reactor trip breakers were redesigned

such that an automatic trip occurs by either an under-voltage trip or shunt coil trip. With the

redesigned breaker, the automatic signal would have tripped the reactor at Salem.

RESOLUTION: No change required. There were actually two failures at Salem 3-days apart.

See resolution to Comment 17 above. Statistically, there are two perspectives and both are

presented. As stated in Section 3.2.5, the RTBs are an Important common-mode failure.

WOG COMMENT 19: Section 3.2.5, Item 1. With regard to the importance of common cause

failures, it should be noted in this section that these importances are based on an RPS model

that does not credit the operator action to trip the reactor. To get a more realistic assessment of the

importance of common cause failures it is necessary to include the operator action to trip the reactor

from the control room.

RESOLUTION: No change required. See resolution to WOG General Comment 3 and 5.

WOG COMMENT 20: Section 3.2.5, Item 2. This is a very important point. A highly reliable

operator action is available to back up a highly reliable system and this needs to be considered

when drawing conclusions about the risk of ATWS events at Westinghouse plants. Also, see

General Comment No. 7 regarding control rod insertion events.

RESOLUTION: See resolution to WOG General Comment 7.

WOG COMMENT 21: Section 3.2.6, Paragraph 3. References 17, 18, and 19 should be

References 16,17, and 18.

RESOLUTION: Revised accordingly.

WOG COMMENT 22: Section 3.2.6, Table 6. The 1979 Baseline information for Westinghouse

plants presented in Table 6 is misstated. As discussed in General Comment No. 4 above, in 1979

the 99% MTC value for Westinghouse plants was -7 pcm/0F (-0.7 1 E-4 ,K/ 0F) and the

corresponding peak RCS pressure was below 3200 psig. Hence, the 1979 Baseline data block in

Table 6 for Peak Pressure should read >3200 psig < 1% of the time." Also, no Westinghouse

plants have a positive MTC at 100% power and not all Westinghouse plants are licensed to operate

with positive MTC at part-power conditions. This should he reflected in the 1994 MTC survey block

in Table 6.

RESOLUTION: See resolution to WOG General Comment 4. The 1994 Survey block reflects

the Tech Spec limit, not the operating point.

WOG COMMENT 23: Section 3.2.6, Paragraph 4. Reference 16 should be Reference 15. Also,

the statement that HFP MTC values are less negative at HFP conditions than they were in 1979 is

correct. These less negative MTC values at HFP are the result of changes in licensed

allowable operation at part-power conditions associated with the implementation of positive MTC

Technical Specifications. The need for positive MTC at part-power conditions is to support longer

operating cycles. Not only are longer cycles an economic benefit, the reduction in the number of

startups and shutdowns serve to reduce the potential number of reactor trip demands. In 1979 the

C-25
TS MTC at full power was 0 pcm/0F, the same as it is today. Hence, the peak ATWS pressure at

the HFP MTC limit remains unchanged.

RESOLUTION: The reference numbers were revised accordingly. See resolution to WOG

General Comment 4 and the Clarification.

WOG COMMENT 24: Section 3.2.6, Paragraph 5. The statement that 'WCAP-1 1992 described a

risk based approach to justify increasing UETs from 1 percent to 37 percent' is incorrect. The intent

of WCAP-1 1992 was to present a PRA methodology for demonstrating continued compliance with

the risk-based PRA approach used in the basis of the Final ATWS Rule and to show that the value

of P(ATWS) of 1 .OE-05 for Westinghouse plants was still being met when considering changes in

plant conditions (e.g., MTC, AFW, PORV availability) important to ATWS. As indicated in Comment

No. 16 above, there are UETs associated with different sets of conditions. Referencing a UET for

only one set of conditions does not provide the complete picture.

RESOLUTION: See Resolution to WOG General Comment 4.

WOG COMMENT 25: Section 3.2.6, Paragraph 6. The statement Since the UET is the

percentage of the fuel cycle during which an ATWS is unmitigated, increasing the UET from

1 percent to 37 percent results in increasing P(ATWS) by a factor of 37 as calculated by the

techniques of SECY-83-293" may be correct if the techniques of SECY-83-293 are correct. But

SECY-83-293 is no longer an appropriate approach for this calculation. It does not account for

the various configurations in which the plant may be operating and does not credit partial rod

insertion. Increasing the UET by a factor of 37 may have no impact on ATWS risk if the plant is

operating in a configuration consistent with a 0 UET. The 37% UET referred to here is

associated with a plant operating configuration that may not be the most probable operating

configuration. As discussed above in General Comment No. 6, CE and B&W plants were

required to install DSS and AMSAC due to limited pressure relief capacity relative to

Westinghouse plants. An option of adding relief valves was also considered to reduce the peak

pressure in CE and B&W plants but was eliminated due to the high cost associated with adding

valves and the resulting low valve-impact ratio. Finally, the last two sentences in this paragraph

appear to imply that the implementation of DSS would resolve any ATWS concerns for

Westinghouse plants seeking significantly higher UETs. If this is true, this option should be

more clearly described in the corresponding text for this section and in the Executive Summary

and Conclusions sections of the report. -

RESOLUTION: See Clarification and resolution to WOG General Comment 4.

WOG COMMENT 26: Section 3.3, Paragraph 3. The first sentence should read .... effects of

the industry ..... in 49FR26036." The last sentence should reflect 7140 trips. Also this sentence

uses 20 years in determining the total number of trips but Table 7 reflects 30 years.

RESOLUTION: Sentence edited as suggested. See Resolution to CE comment.

WOG COMMENT 27: Section 3.3, Paragraph 4. The last sentence should read ....the

ATWS rule was cost effective."

RESOLUTION: No change required. This is a conclusion that can be found in conclusions,

Section 4.0 of the report as stated in this comment.

C-26
WOG COMMENT 28: Section 3.3 and Section 4, Second Bullet. It is not reasonable to claim

the reactor trip frequency reduction is due entirely to the Commission's recommendation to

reduce the number of scrams. A good percentage of this reduction would most likely have

occurred anyway as utilities gained experience in plant operation, from changes to test and

maintenance strategies, and due to economic pressures to keep the plants online.

RESOLUTION: No change required. There is no such claim. The Section 2.2 acknowledges

this was an industry initiative and that the Commission involvement served to obtain an industry

commitment to formal the program.

WOG COMMENT 29: Section 4, First and Eighth Bullets. See earlier discussions on the

definition of UET and on the reference to 1% UET in lieu of DSS for Westinghouse plants.

These apply here also.

RESOLUTION: See resolution to WOG General Comment 1.

WOG COMMENT 30: Table B-I. As previously discussed in Comment No. 10, comments received

from WOG utility representatives question the basis for some of the plant-specific numbers in Table

B-1. It is recommended that a reference basis be included in the table for the

plant-specific values to allow a cross-reference with the data source. Otherwise, the validity of

the conclusions drawn from the use of these values may be questionable. Since so much of the

discussion in this report involves the plant type by NSSS vendor, it is suggested that a column be

added to identify the NSSS vendor / plant type.

The units for the MTC should be provided in the last column title block. Many of the values

given in this column are incorrect or misrepresented. For example, the TS MTC for the Byron,

Braidwood, and V.C. Summer units is not just +7E-05. The TS MTC for these plants are the

same as given for the Catawba units. The Robinson Unit 2 TS MTC (per TS 3.1.3 in

Amendment #176) is +5E-05 below 50% RTh and 0 above 50% RTh. The South Texas units

show a 0 MTC in Table B-1 but the TS MTC is the same as given for the Comanche Peak units.

RESOLUTION: Table B-1 was updated.

C-27
COMBUSTION ENGINEERING (CE) provided the following comments in a letter to the NRC dated

December 29, 2000.

CE COMMENT: After review of the reference letter, the CEOG finds the basis for the ATWS rule

reasonable. However, it is difficult to attribute to the ATWS rule the economic benefits achieved by

extensive industry efforts to reduce plant trips and challenges to safety systems. The CEOG

applauds staff efforts to improve the cost effectiveness of NRC activities and decisions, and

welcomes opportunities to promote regulatory efficiency and realism.

RESOLUTION: The Section 3.3 of the report is revised as follows: "In addition, although not

included in the this value impact analysis, the 3.5 reduction in the average number of automatic

scrams for 102 operating reactors has significant monetary value due to replacement power

costs. "

C-28
NRC FORM U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-9) (Assigned by NRC. Ad Vol, Sp., Rev.

3201.3202 BIBLIOGRAPHIC DATA SHEET *dtdd r2ureSfy.)

(Se hnaions on me rwse)

2. TITLE AND SUBTITLE NUREG-1780

Regulatory Effectiveness of the Anticipated Transient Without Scram Rule 3. DATE REPORT PUBUISHED

MONTH | YEAR

09 2003

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

William S. Raughley George F. Lanik

7. PERIOD COVERED (lickniv Date)

1984 - -1999

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (NVNRCp oviSdeDsbm OflorRegUn.S NudearRgu oComn assnr mWling adde conaor prid name and maift address.)

Division of Systems Analysis and Regulatory Effectiveness

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

S. SPONSORING ORGANIZATION - NAME ANDADDRESSboNvCe to.Saaet^roncrpvNRCOF~don, Officeor Region, US.NudarR rnissy

Same as above.

10. SUPPLEMENTARY NOTES

11. ABSTRACT (20 words or Aess)

The Nuclear Regulatory Commission's Office of Nuclear Regulatory Research is reviewing selected regulations to determine if

the requirements are achieving the desired outcomes. This nitiative Is part of an evolving program to make NRC activities and

decisions more effective, efficient, and realistic. The goal of this evaluation is to determine whether the anticipated transient

without scram (ATWS) rule and the recommendations Issued with it were effective. The effectiveness of the ATWS rule was

determined by comparing regulatory expectations to outcomes. The report conclwded that the ATWS rule was effective in

reducing ATWS risk and that the cost of implementing the rule was reasonable. However, uncertainties in reactor protection

system reliability and mitigative capability may warrant further attention to ensure the expected levels of safety are maintained.

12. KEY WORDS/DESCRIPTORS (Listwrd aphrases tsat Jiseh kig m t ,e*,ot) 13 AVAILASLITY STATMENT

anticipated transient without scram unlimited

reactor protection system 14. SECURITY CLASSIFICAT1ON

common-cause failure (This Page)

unclassified

(This Repor)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-19 This form was alselrorwally produred by Eft Federal Forms, Ine-
N

LPrinted

on recycleds

paper

Federal Recycling Program
t 4 . -- _ _ _ _

NUREG-1780 REGULATORY EFFECTIVENESS OF THE ANTICIPATED TRANSIENT

WITHOUT SCRAM RULE

SEPTEMBER 2003

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS