October 20, 2016 

MEMORANDUM TO: William M. Dean, Director 

 Office of Nuclear Reactor Regulation 

FROM: John W. Lubinski, Director /RA MRoss-Lee for/

 Division of Engineering 

 Office of Nuclear Reactor Regulation 

 

 Joseph G. Giitter, Director /RA/

 Division of Risk Assessment 

 Office of Nuclear Reactor Regulation 

SUBJECT: DEGRADATION OF BAFFLE-FORMER BOLTS IN 

PRESSURIZED-WATER REACTORS—DOCUMENTATION OF 

INTEGRATED RISK-INFORMED DECISIONMAKING PROCESS 

IN ACCORDANCE WITH NRR OFFICE INSTRUCTION LIC-504 

In accordance with Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-504, 

Revision 4, “Integrated Risk-Informed Decision-Making Process for Emergent Issues,” effective 

June 2, 2014, the staff of the U.S. Nuclear Regulatory Commission (NRC) has performed a riskinformed evaluation of the safety significance of recently identified reactor vessel baffle-former 

bolt (BFB) degradation. As discussed in the enclosed evaluation, the staff has identified the 

facilities of greatest concern, assessed the need for immediate shutdown of those facilities, and 

prepared available options. This memorandum presents the staff’s recommendation. Note that 

the risk-informed assessment of the BFB degradation issue considered risk in conjunction with 

other factors (e.g., defense-in-depth and safety margins) in order to inform the NRC staff’s 

recommendation. 

In the reactor vessel internals of most pressurized-water reactors (PWRs), BFBs connect the 

vertical baffle plates to the horizontal former plates forming the overall baffle-former assembly, a 

component of the reactor vessel internals. The main function of the baffle plates is to direct and 

concentrate the flow of coolant through the core. Detachment or deflection of the baffle plates 

as a result of degraded BFBs could challenge core coolable geometry and control rod insertion. 

During the spring 2016 refueling outage inspections, licensees of two PWRs identified 

unexpected levels of degradation of the BFBs and concluded that the plants were in an 

unanalyzed condition. The NRC staff evaluated the potential risk and associated consequences 

of continued plant operation with extensive BFB degradation and determined the regulatory 

actions that would be needed. 

The two plants with high levels of BFB degradation are Westinghouse reactors with a 4-loop 

design, operating in a downflow configuration with Type 347 stainless steel bolts. The NRC 

staff determined that plants with these characteristics are most susceptible to high levels of BFB 

degradation and identified a total of seven reactors in this group. Although the two plants that 

found extensive bolt degradation have taken corrective actions, including replacement of 
- 2 - 

 

degraded bolts, and have subsequently restarted plant operations, five reactors in the 

susceptible group have not inspected the bolts to date. The staff, therefore, evaluated the risk 

represented by continued operation of these plants, where a similar level of bolt degradation 

may exist, and determined whether immediate shutdown and inspection of the bolts are 

necessary. The attached Table 1 describes the LIC-504 process initiation for this issue and 

identifies the evaluation team. 

The staff’s evaluation identified the following options for the susceptible group: 

• Option 1: Require immediate shutdown and inspection of the BFBs. 

• Option 2: Allow continued operation until the next refueling outage, at which time the 

plants have committed to examine all accessible BFBs. 

• Option 3: Issue a generic communication to gather additional information to support a 

future regulatory decision. 

• Option 4: Maintain the status quo, under which plants would inspect the BFBs 

consistent with the current recommended schedule in Materials Reliability Program 

(MRP)-227-A. 

The NRC staff recommends Option 2 because the level of risk represented by operation for only 

one cycle does not warrant immediate regulatory action per the guidance in LIC-504, i.e., an 

associated core damage frequency less than 1x10-3 per reactor year and a large early release 

frequency less than 1x10-4 per reactor year. While this is also true for Option 1, immediate 

shutdown would place an unnecessary burden on the plants. The staff eliminated Options 3 

and 4 because they would extend the time frame for inspections or other corrective actions and 

increase the uncertainties related to risk. 

MRP Letter 2016-022, dated July 27, 2016, transmitted interim guidance from the Electric 

Power Research Institute (EPRI) MRP recommending that all the plants in the most 

susceptible group (Westinghouse 4-loop, downflow, with Type 347 stainless steel BFBs) 

conduct ultrasonic testing (UT) of all BFBs at the next scheduled refueling outage. For 

these plants, the next refueling outage is no later than fall 2017. This guidance is classified as 

“needed” under the industry materials protocol, Nuclear Energy Institute (NEI) 03-08, “Guideline 

for Management of Materials Issues,” issued January 2010. The NRC staff plans to review the 

interim guidance and document its review in a publically available safety assessment. The staff 

does not expect that any of the plants in the susceptible group will deviate from the 

recommended EPRI MRP interim guidance. Therefore, the staff does not expect to issue 

regulatory actions to ensure the implementation of Option 2. However, NEI 03-08 calls for the 

NRC to be notified of deviations from “needed” guidance. Therefore, if any licensee in the most 

susceptible group intends to deviate from the interim guidance, the NRC would be notified and 

could take regulatory action to ensure that the plant performs UT examinations at the next 

refueling outage. 

The NRC staff notes that the enclosure references Westinghouse letter NSAL 16-1, Revision 1, 

dated August 1, 2016. This letter was issued to Westinghouse PWR owners to provide a 

10 CFR Part 21 evaluation and recommendations in response to BFB degradation. The NRR 
- 3 - 

 

staff has not reviewed engineering analyses supporting the evaluation or endorsed its 

conclusions. It is discussed in this enclosure only to provide context to the staff’s own 

engineering judgement in evaluating potential risk and regulatory options. 

The enclosure contains details of the staff’s evaluation. 

Enclosure: 

Review of Degraded Reactor Core Baffle-Former Bolts in Westinghouse 4-Loop Pressurized 

Water Reactors in Accordance With Office of Nuclear Reactor Regulation Office Instruction 

LIC-504, “Integrated Risk-Informed Decision-Making Process for Emergent Issues” 

CONTACTS: Jeffrey C. Poehler, NRR/DE/EVIB Steven Laur, NRR/DRA/APLA 

 301-415-8353 301-415-1465 
- 3 - 

 

 

staff has not reviewed engineering analyses supporting the evaluation or endorsed its 

conclusions. It is discussed in this enclosure only to provide context to the staff’s own 

engineering judgement in evaluating potential risk and regulatory options. 

The enclosure contains details of the staff’s evaluation. 

Enclosure: 

Review of Degraded Reactor Core Baffle-Former Bolts in Westinghouse 4-Loop Pressurized 

Water Reactors in Accordance With Office of Nuclear Reactor Regulation Office Instruction 

LIC-504, “Integrated Risk-Informed Decision-Making Process for Emergent Issues” 

CONTACT: Jeffrey C. Poehler, NRR/DE/EVIB Steven Laur, NRR/DRA/APLA 

 301-415-8353 301-415-1465 

Distribution 

PUBLIC LPL1-1 R/F RidsNrr RidsNrr/Dorl 

RidsNrrDra RidsNrrDeEvib RidsNrrDraApla RidsNrrDe 

RidsRgn1MailCenter RidsRgn2MailCenter RidsRgn3MailCenter RidsRgn4MailCenter 

MRoss-Lee JPoehler SLaur CFong 

PClifford SLyons JHickey EBenner 

GWilson RidsNrrPMIndianPoint RidsNrrPMDCCook 

RidsNrrPMSalem RidsNrrDiabloCanyon MScott, R1 

GDentel, R1 MGray, R1 NFloyd, R1 KMangan, R1 

BHaagensen, R1 AZiedonis, R1 FBower, R1 PFinney, R1 

JKulp, R1 EBurket, R1 TGody, R2 SWalker, R2 

AButcavage, R2 KO’Brien, R3 MHolberg, R3 EFernandez, R3 

DHills, R3 TVegel, R4 GWerner, R4 

RidsAcrsAcnw_MailCTR ITseng, EMCB 

ADAMS Accession Number: ML16225A341 *concurred by email 

OFFICE DORL/LPL1-1/LA NRR/DE/EVIB NRR/DRA/APLA NRR/DRA/APLA 

NAME KGoldstein* JPoehler* SLaur* CFong 

DATE 8/16/2016 8/15/2016 8/17/2016 10/04/2016 

OFFICE NRR/DE/EVIB/BC QTE NRR/DORL/D RI/DRS/BC 

NAME JMcHale JDougherty (EBenner for) ABoland PKrohn* 

DATE 8/25/2016 8/19/2016 10/06/2016 10/05/2016 

OFFICE OGC NRR/DE/EVIB/BC NRR/DRA/D NRR/DE 

NAME BHarris DRudland JGiitter (MRoss-Lee for) JLubinski 

DATE 10/13/2016 10/18/2016 10/19/2016 10/20/2016 

 OFFICIAL RECORD COPY
 

Table 1: LIC-504 Process Initiation 

Date LIC-504 Initiated: 5/12/2016 Date of Report: [ ] draft [X] final 

Summary Description of Issue: An unexpectedly large number of potentially defective 

baffle-former bolts have been found in two pressurized-water reactors (PWRs). All 

Westinghouse, Babcock & Wilcox, and two Combustion Engineering-designed PWRs have 

baffle-former bolts. Plants that have identified significant baffle-former bolt degradation are 

limited to older, 4-loop-designed Westinghouse plants with downflow configuration and 

Type 347 stainless steel bolts. The LIC-504 process will evaluate whether this issue warrants 

immediate shutdown and inspection of plants where baffle-former bolts are susceptible to 

extensive degradation, or whether other options can be pursued, such as allowing the 

susceptible plants to be inspected at the next scheduled refueling outage. 

Decision Authority Name/Title Organization Telephone 

John Lubinski/Director NRR/DE 301-415-3298 

Evaluation Team: Name/Title Organization Telephone 

Management Lead Dave Rudland/Branch 

Chief 

NRR/DE/EVIB 301-415-1896 

Project Manager Doug Pickett/Sr. Project 

Manager 

NRR/DORL/LPL1-1 301-415-1364 

Technical Lead Jeff Poehler/Sr. 

Materials Engineer 

NRR/DE/EVIB 301-415-8353 

Team Members Steve Laur/Sr. 

Reliability and Risk 

Analyst 

NRR/DRA/APLA/RILIT 301-415-1465 

 Sara Lyons/Reliability 

and Risk Analyst 

NRR/DRA/APLA/RILIT 301-415-2861 

 Paul Clifford/Sr. 

Technical Advisor for 

Reactor Fuel 

NRR/DSS 301-415-4043 

 Ian Tseng/Mechanical 

Engineer 

NRR/DE/EMCB 301-415-7964 

 James Hickey/Sr. 

Technical Advisor for 

Reactor Systems 

NRR/DSS 301-415-2180 

 Bob Hardies/Sr. 

Technical Advisor for 

Materials Science and 

Engineering Mechanics

NRR/DE 301-415-5802 
 

Enclosure 

Review of Degraded Reactor Core Baffle-Former Bolts in Westinghouse 

4-Loop Pressurized-Water Reactors in Accordance with Office of Nuclear 

Reactor Regulation Office Instruction LIC-504, “Integrated Risk-Informed 

Decision-Making Process for Emergent Issues” 

ISSUE SUMMARY 

The U.S. Nuclear Regulatory Commission (NRC) has recently identified degradation in a 

significant number of core baffle-former bolts (BFBs) in two Westinghouse 4-loop units (Indian 

Point, Unit 2, and Salem, Unit 1) (Refs. 1, 2) during refueling outage inspections. The 

mechanism that causes this degradation is believed to be irradiation-assisted stress-corrosion 

cracking (IASCC), with potential contributions from fatigue and stress relaxation. Visual 

examinations of the degradation found missing or protruding bolt heads. Failure of the bolts 

may result in loose parts in the reactor coolant system (RCS). Degradation of a sufficient 

number of these bolts could result in baffle deformation or displacement, which could physically 

affect fuel cladding or increase coolant bypass flow. 

The licensees discovered BFB degradation while performing ultrasonic testing (UT) 

examinations of BFBs in accordance with NRC-approved Electric Power Research Institute 

(EPRI) Materials Reliability Program (MRP) Topical Report MRP-227-A, “Materials Reliability 

Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines” (Ref. 3). 

Inspections of core support structures are conducted at a 10-year interval in accordance with 

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME 

Code), Section XI, inservice inspections of core support structures. The determination of which 

reactor vessel internal components are considered core support structures is licensee specific. 

Therefore, some licensees may consider the baffle-former assembly and the BFBs to be core 

support structures while others may not. The ASME Code-required visual examination, known 

as a VT-3 visual examination, is intended to determine the general mechanical and structural 

condition and does not require a close examination of the BFBs. 

To date, no displacement of baffle plates has been observed, even in plates with a large 

majority of degraded bolts. A bounding estimate of the risk of this condition indicates that core 

damage frequency is less than 1x10-3 per year (see the risk assessment in the next section of 

this report). Therefore, the NRC staff concludes that immediate regulatory action is not 

warranted in response to this issue. 

Although degradation of BFBs was first identified in the late 1980s, recent operating experience 

(OE) indicates that degradation can be more extensive than previously predicted and is an 

unanalyzed condition. Therefore, the NRC assessed the generic implications of BFB 

degradation for pressurized-water reactors (PWRs). 

The purpose of this activity under LIC-504, “Integrated Risk-Informed Decision-Making Process 

for Emergent Issues,” effective June 2, 2014, is to identify those plants most susceptible to BFB 

degradation, assess any immediate safety concerns associated with continued plant operation 

of the most susceptible plants, develop options for addressing the BFB degradation issue, 

evaluate each option using the risk-informed approach, and make a recommendation to NRC 

management. This report also documents the management decision and the basis for that 

decision. 
- 2 - 

ISSUE CHARACTERIZATION 

This issue involves degradation of BFBs that secure the baffle plates to the former plates which, 

in turn, are bolted to the core barrel by barrel-former bolts (Figure 3 of Attachment 1). Operating 

experience indicates that designs using Type 347 stainless steel and a “downflow” configuration 

are most susceptible to this degradation. During the spring 2016 refueling outage inspections, 

licensees for two PWR plants determined that 27 and 22 percent of their BFBs were potentially 

degraded and that their plants were in an unanalyzed condition (Refs. 1 and 2, respectively). 

Evaluations of plant operations with significant BFB degradation, combined with a loss-ofcoolant accident (LOCA) or seismic event, are necessary to identify the likelihood and potential 

consequences of baffle plate deformation or displacement. 

This section of the report is organized into the following parts: 

(1) Design and Function of Baffle-Former Assembly 

(2) Operating Experience—Observed Degradation 

(3) Bolting Pattern Analyses 

(4) Evaluation of Events that Could Result in Failure of Baffles with Degraded BFBs 

(5) Potential Consequences of BFB Cracking 

(6) Evaluation of the As-Found Condition—Cracked BFBs 

– Applicable Regulations 

– Defense in Depth 

– Safety Margins 

– Risk Assessment 

– Performance Monitoring 

1. Design and Function of Baffle-Former Assembly 

Attachment 1 provides a detailed discussion of the baffle-former assembly. 

This assembly is made up of vertical baffle plates that are bolted to horizontal former plates 

which, in turn, are bolted to the core barrel by the barrel-former bolts. The degraded bolts are 

those that attach the vertical baffle plates to the horizontal former plates and are referred to as 

“baffle-former bolts,” or BFBs in this document. Bolts made from one of several different 

stainless steel alloys have been employed in different core designs. 

The core design can be such that reactor coolant flow between the core barrel and the baffle is 

either upwards or downwards. These designs are referred to as either “upflow” or “downflow” 

configurations, respectively. Some plants have converted from the downflow to the upflow 

configuration at some point in their operating history. 

2. Operating Experience—Observed Degradation 

Two plants found significant numbers of degraded BFBs during spring 2016. These plants are 

Indian Point, Unit 2, which found 27 percent of BFBs potentially degraded, and Salem, Unit 1, 

which found 22 percent of BFBs potentially degraded. Clustering of failed bolts (large groups of 

contiguous or adjacent failed bolts) occurred at both plants but was more severe at Salem 1, 

where three of eight octants showed degradation in most of the bolts, while other octants had 

only a few degraded bolts. Visual examinations of baffle-edge bolts in these plants found no 

evidence of degradation. 
- 3 - 

Voluntary UT examinations of BFBs were performed at four Westinghouse-designed reactors in 

the late 1990s. Two of these plants were 2-loop designs with Type 347 BFBs, while two were 

3-loop designs with Type 316 BFBs. The 2-loop plants found 5–10 percent of BFBs potentially 

degraded, while the 3-loop plants found no degraded bolts. Bolts were replaced at all four 

plants. All the potentially defective bolts were replaced at one 2-loop plant, while the defective 

bolts plus some additional bolts were replaced at the other 2-loop plant. The licensee for the 

3-loop plants proactively replaced a subset of the bolts. 

Before spring 2016, UT examinations, in accordance with MRP-227-A (Ref. 3) at plants with 

29–35 effective full-power years (EFPYs) of operation, found 10 percent or less of the BFBs 

potentially degraded1 in 2-loop and less than 1 percent of the BFBs potentially degraded in 

3-loop designs. The total number of degraded bolts in the 2-loop and 3-loop plants does not 

represent a safety issue because it is bounded by minimum bolting pattern analyses for those 

plants. 

In 2010, one Westinghouse 4-loop plant, D.C. Cook Unit 2, conducted a visual examination and 

identified a cluster of broken bolts on one large baffle plate. The licensee replaced the broken 

bolts and some additional bolts, finding a total of about 40 defective bolts on the plate. The 

licensee did not perform UT examinations of the BFBs. 

Based on a review of operating experience, the potential for significant bolt degradation is 

currently believed to be limited to Westinghouse 4-loop designs with a downflow configuration 

and Type 347 stainless steel bolts. Westinghouse Technical Bulletin TB-12-5, issued 

March 2012 (Ref. 4), following the operating experience at D.C. Cook Unit 2, identified seven 

operating reactors that were considered most susceptible to BFB degradation—Indian Point 

Units 2 and 3, Salem Units 1 and 2, D.C. Cook Units 1 and 2, and Diablo Canyon Unit 1. The 

NRC staff confirmed the bolt material as Type 347 stainless steel for all these plants. 

Westinghouse Nuclear Safety Advisory Letter (NSAL) 16-12, Revision 1 “Baffle-Former Bolts,” 

dated August 1, 2016 (Ref. 5), issued in response to the experience at Indian Point, Unit 2, and 

Salem, Unit 1, identifies the same seven plants as being most susceptible to BFB cracking. 

NSAL 16-1 classifies the seven 4-loop downflow plants with Type 347 bolts as “Tier 1a.” EPRI 

Materials Reliability Program (MRP) Letter 2016-022, dated July 27, 2016 (Ref. 6), contains 

MRP interim guidance that recommended that all plants identified as Tier 1a plants in 

Westinghouse NSAL-16-1 conduct UT examinations of all BFBs at the next scheduled refueling 

outage. This guidance is classified as “needed,” as defined in the protocol of NEI 03-08, 

“Guideline for Management of Materials Issues,” Revision 2, issued January 2010 (Ref. 11). 

The identification of the most susceptible group of plants to BFB cracking in the NSAL and the 

MRP letter is consistent with the staff’s assessment based on its OE review. 

Two 4-loop, downflow plants with Type 316 stainless steel bolts, Sequoyah Units 1 and 2, are 

classified as Tier 1b in NSAL 16-1. MRP Letter 2016-022 (Ref. 6) recommends that the Tier 1b 

plants complete a visual VT-3 examination of the full population of BFBs at the next scheduled 

refueling outage. MRP Letter 2016-022 further recommends that, if degradation is detected, the 

plants complete actions consistent with the Tier 1a plants; if no degradation is detected during 

 

1 This number is cumulative, counting BFBs found during previous inspections in the late 1990s.

2 NSAL-16-1 was issued to Westinghouse PWR owners to provide a 10 CFR Part 21 evaluation 

and recommendations in response to recent BFB degradation. The NRR staff has not reviewed the 

engineering analyses supporting the evaluation or endorsed its conclusions or methods. It is discussed in 

this enclosure only to provide context to the staff’s own engineering judgement in evaluating potential risk 

and regulatory options. 
- 4 - 

the visual VT-3 examination, plants should complete a UT examination consistent with the 

guidance for Tier 1a during the second refueling outage after issuance of the interim guidance. 

As discussed above, significant BFB degradation has been observed only in 

Westinghouse-designed PWRs. All but two Combustion Engineering (CE) units employ a 

welded, rather than bolted, baffle assembly (referred to as the core shroud in CE designs). UT 

inspections have not yet been conducted at the CE units with bolted core shrouds, but these 

units are considered less susceptible to cracking, because annealed Type 316 bolts were used. 

Babcock & Wilcox-designed PWRs use Type 304 stainless steel BFBs, have an upflow 

configuration, and have had very few degraded bolts. Based on operating experience, it 

appears that 4-loop Westinghouse plants with the “downflow” baffle flow configuration and 

Type 347 stainless steel BFBs are the most susceptible. Table 1 lists the Westinghouse 4-loop 

downflow plants with Type 347 BFBs. 

Several plants were originally designed as downflow but were modified to convert to upflow. 

Because of the lower differential pressure across the baffle plates, upflow plants are considered 

somewhat less susceptible to BFB cracking. Westinghouse NSAL-16-1 identifies three 4-loop 

converted upflow plants in the United States—Diablo Canyon Unit 2, and McGuire Unit 1 and 

Unit 2. The recommendation for these plants in NSAL-16-1 is that, if the 4-loop converted 

upflow plants have operated for more than 20 calendar years in a downflow configuration, they 

should evaluate the need to conduct a UT volumetric inspection of BFBs on an accelerated 

schedule, considering the plant-specific condition and design parameters compared to the 

4-loop downflow plants with Type 347 bolts. NSAL-16-1 did not explain the reason for choosing 

20 years as the threshold above which converted upflow plants should consider accelerated 

inspection. However, the NRC staff notes that most Westinghouse PWRs switched to a 

low-leakage core design well before 20 calendar years; therefore, the 20-year threshold for 

consideration of accelerated inspections may be reasonable for converted upflow plants, since 

the combination of lower stresses from the upflow conversion and reduced neutron flux from a 

low-leakage core would result in fewer bolts initiating IASCC. The NRC staff will continue to 

gather information on the potential need to expand the group of plants considered most 

susceptible to degradation. 

 
- 5 - 

Table 1 Westinghouse 4-Loop Downflow Plants with Type 347 BFBs 

Plant EFPYs Licensed 

MWt 

Flow 

Configuration 

Bolt 

Material 

Year 

MRP-227-A 

Baseline 

Performed or 

Scheduled 

Notes

D.C. Cook 1 28 3,304 Downflow Type 347 SS Fall 2017 1, 

D.C. Cook 2 27 3,468 Downflow Type 347 SS Fall 2016 1, 2 

Diablo Canyon 1 27 3,411 Downflow Type 347 SS Spring 2017 1, 2 

Indian Point 2 31.1 3,216 Downflow Type 347 SS Spring 2016 1 

Indian Point 3 27 3,216 Downflow Type 347 SS Spring 2017 1, 2 

Salem 1 27 3,459 Downflow Type 347 SS Spring 2016 

1 

Salem 2 24.6 3,459 Downflow Type 347 SS Spring 2017 1, 2 

1. In the susceptible group of plants in Westinghouse Technical Bulletin TB-12-05 (Ref. 4) and NSAL 16-01 (Ref. 5) 

2. Expected revised inspection date based on MRP interim guidance 

3. Bolting Pattern Analysis 

WCAP-15270, “Determination of Acceptable Baffle-Barrel-Bolting for Four-Loop Westinghouse 

Downflow and Converted Upflow Domestic Plants,” dated November 30, 2001 (Ref. 7), 

documents a generic minimum bolting pattern analysis for Westinghouse 4-loop plants. This 

analysis used the NRC-approved minimum bolting pattern methodology of WCAP-15029-P-A, 

“Westinghouse Methodology for Evaluating Acceptability of Baffle-Former-Barrel Bolting 

Distributions under Faulted Load Conditions,” dated December 31, 1998 (Ref. 8). The 

nonproprietary version of this report is WCAP-15030-NP-A (Ref. 9). WCAP-15270 covers 

downflow and converted upflow 4-loop designs and analyzes a large number of hypothetical 

cases of BFB degradation. The WCAP analysis demonstrated that a very large percentage of 

BFBs could fail and still meet the acceptance criteria, but none of the cases analyzed included 

clustering of failed bolts as extensive as was seen at Indian Point, Unit 2, or Salem, Unit 1. 

4. Evaluation of Events that Could Result in Failure of Baffles with Degraded BFBs 

The NRC staff evaluated whether normal operations, LOCAs, or seismic events are capable of 

causing deformation or detachment of baffle plates. These events are evaluated against their 

potential contribution to core damage frequency (CDF) separately below. CDF is typically 

defined as the likelihood of an accident involving the loss of adequate cooling to reactor fuel 

elements up to and including major damage to a reactor core with the consequent release of 

fission products but not necessarily involving a breach of the reactor vessel. 

Normal Operation 

During normal operation, it is unlikely that baffle plates would detach or significantly deflect, 

even with a significant number of potentially degraded bolts, since plates were observed to 
- 6 - 

remain in place in the severely affected plants even with plates in which most of the bolts were 

degraded. It is most likely that only one baffle plate would detach, since not all plates would 

have the same number of degraded bolts. However, if the plate were detached or had large 

corner gaps, localized fuel cladding damage might occur that would then be detected by reactor 

coolant activity monitoring. The most likely effects of BFB degradation during normal operation 

are the presence of loose parts in the form of bolt heads and locking bars, and possible gaps 

between the corners of adjacent baffle plates. The loose parts are relatively small and may not 

be identifiable to control room operators through the loose parts monitoring systems. Based on 

the previous statements, the contribution of degraded BFBs to CDF during normal operation can 

be considered negligible. 

Large-Break LOCA 

All Westinghouse plants are licensed for leak-before-break (LBB) for the main reactor coolant 

loop piping. Therefore, the largest size line that must be considered for any LOCA analysis of 

the baffle plates are in the range of 10-14 inches, which are the largest RCS branch lines 

(pressurizer surge line, residual heat removal line, or accumulator line). Line breaks in this size 

range have been analyzed in various generic and plant-specific acceptable bolting pattern 

analyses, with acceptable results, but not for cases of clustering of degraded BFBs. The NRC 

staff evaluated the potential for one or more baffle plates detaching from the formers and 

affecting peripheral fuel assemblies following a LOCA combined with a significant number of 

degraded BFBs. Based on the observed patterns of bolt degradation at the two plants, it is 

unlikely that a significant number of baffle plates would detach, but it is possible that one or 

more plates could detach and affect the fuel. The impact of the plates on the peripheral fuel 

assemblies from the differential pressure of a LOCA could cause localized damage to the fuel 

assemblies (grid crush) and possible difficulties inserting control rods associated with the 

peripheral fuel assemblies. However, based on a sampling of information in the updated final 

safety analysis reports (UFSARs) for the plants of interest, the plants of most concern (4-loop 

downflow plants) do not have control rods in peripheral locations, with the exception of Salem 

Unit 2. During an inspection, the NRC staff reviewed a calculation performed by the licensee for 

Salem Unit 2, demonstrating that the four control rods in peripheral locations are not needed for 

safe shutdown. It is expected that the overall flow volume through the core would not be 

affected if baffle plates detached, although local cooling to some peripheral fuel assemblies 

could be degraded. Since most plants have switched to low-leakage core designs, peripheral 

assemblies tend to have higher burnup and are lower power assemblies, which should mitigate 

cooling concerns. Westinghouse did not perform a coolable geometry analysis considering 

degraded BFBs for a large break LOCA in support of NSAL-16-1 (Ref. 5). However, the large 

margin available to the cladding temperature limit associated with a large break LOCA was used 

to support the evaluation in NSAL-16-1 of a medium break LOCA, as detailed in the next 

section. 

Detachment of one or more baffle-former plates would also change the flow pattern, in that 

some of the bypass flow between the baffle plates and the core barrel would leak into the main 

flow through the core region, but this should not significantly affect the overall flow volume 

through the core. 

 
- 7 - 

Medium-Break LOCAs 

It is expected that the forces on the bolts during a medium-break LOCA3 would be smaller than 

the forces resulting from a 14- or 10-inch line break. The staff bases this opinion on its review 

of information on bolt stresses and fuel grid loads in WCAP-15270 (Ref. 7) for various different 

sized line breaks. Thus, unless clustering of degraded bolts exists, coolable geometry would 

likely be maintained during a medium break LOCA if the percent of degraded bolts is similar to 

that observed in Indian Point Unit 2, or Salem Unit 1. 

NSAL-16-1 discussed the potential for loss of coolable geometry if a LOCA occurs, assuming a 

clustering of degraded bolts. A 4-inch line break was the largest break considered, which would 

bound the medium break LOCA as defined by the NRC. The evaluation in NSAL-16-1 assumed 

all the BFBs were failed on one baffle plate, consistent with recent OE showing clustering of 

failed bolts. The NSAL indicated that there is a potential for grid deformation on the core 

periphery as well as the potential for limited grid deformation in the inboard fuel assemblies. 

However, the NSAL concluded that a core-coolable geometry is maintained for assemblies on 

the core periphery with grid deformation, based on taking credit for the low power generation in 

the peripheral assembles and the observation that any flow redistribution that may occur would 

tend to benefit the inboard assemblies. The NSAL indicated that the overall core flow area is 

unchanged, with only a small difference in the cooling geometry. The NSAL, therefore, 

concluded that the increase in core bypass and the amount and location of the grid deformation 

would have little impact on the 4-inch LOCA transient. The NSAL further concluded that, based 

on this and considering the available margin to the cladding temperature limit and the margins in 

the small-break LOCA evaluation model, core cooling is maintained for the grid deformation 

caused by the potentially degraded BFB. The large margin to the cladding temperature limit 

associated with large break LOCA is used in NSAL-16-1 to support the determination that the 

predicted fuel grid deformation resulting from a 4-inch line break would not approach the 

cladding temperature limit, and thus a coolable geometry would be maintained. 

The NSAL found that BFB degradation would also not compromise long-term core cooling 

because (1) grid deformation, should it occur, could reduce the rod-to-rod spacing (however, 

significant margins in the licensing basis analyses exist), (2) slight changes in core geometry, 

such as those resulting from fuel assembly grid deformation, would not significantly affect 

long-term heat removal, as coolant flow velocities are low, fuel assembly crossflow would not be 

fully impeded, and the overall core flow area would be preserved, and (3) BFB degradation 

would not increase the core decay heat. 

Small-Break LOCA 

The loads for a small-break LOCA4 would be even smaller for a medium-break LOCA. Even 

with significant clustered defective BFBs, it is unlikely that bolt loads or fuel grid impact loads 

would be high enough to cause baffle plate displacement or localized cladding damage. The 

LOCA evaluation in NSAL-16-1 described in the previous section bounds the small-break 

LOCA. Therefore, the conclusions of the medium-break LOCA evaluation are bounding for 

 

3 This is defined in NUREG-1829 as having a flowrate greater than 1,500 gallons per minute (gpm) and equal 

to or smaller than 5,000 gpm, correlating to an effective break size greater than 1.5 inches and equal to or 

smaller than 3 inches for a PWR. 4 In NUREG-1829, “Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation 

Process,” issued April 2008 (Ref. 10), the NRC staff defines this as having a flow rate greater than 100 gpm 

but equal to or smaller than 1,500 gpm, correlating to an effective break size of greater than 1/2 inch and 

equal to or smaller than 1.5 inches. 
- 8 - 

small-break LOCAs, and little or no fuel grid damage is expected for a small-break LOCA. 

Based on the above, the risk posed by small-break LOCAs can be considered negligible. 

Seismic 

The potential for an earthquake to result in BFB failures was evaluated from two 

perspectives: traditional engineering and risk.

Nuclear plants are designed to withstand earthquakes using accepted engineering 

principles. Earthquakes equal to or exceeding the SSE are considered to be the only significant 

seismic events with respect to BFB failures. The generic Westinghouse 4-loop analysis report, 

WCAP-15270 (Ref. 7), indicates that stresses in the BFBs are much less for an SSE than for a 

LOCA, and that fuel grid impact loads for an SSE alone have considerable margin before 

exceeding the allowable loads. NSAL-16-1 stated that the evaluation considered seismic 

conditions and determined that, for BFB degradation, seismic effects are bounded by LOCA 

effects. NSAL-16-1 further stated that, when a baffle plate is loose, there is no significant 

mechanism to excite the plate. NSAL-16-1 stated that there will be minor pressure fluctuations 

in the baffle-former-barrel region, due to the relative motion of components; however, these 

pressure fluctuations have been shown to be very small. Therefore, NSAL-16-1 stated that any 

pressure results are bounded by those reviewed in the LOCA analysis. NSAL-16-1 also stated 

that vertical seismic effects are not expected to be significant, and intermittent impacts with the 

upper or lower core plates will not have significant inertia to damage the core plates when 

considering an SSE. NSAL-16-1 stated that the upper internals holddown forces have 

significant margin to accommodate any incidental impacts. Therefore, for the purposes of the 

evaluation described in NSAL-16-1, seismic effects were considered decoupled from a LOCA 

event. 

The risk assessment of this event (Attachment 3) was performed to determine whether the 

seismic risk attributable to BFB degradation approached the LIC-504 thresholds that might 

indicate that an imminent safety concern exists. The risk assessment concludes that the BFB 

degradation issue does not represent an imminent safety concern, using a bounding seismic 

hazard estimate that envelops the latest hazard estimates across all U. S. nuclear power plant 

sites. The seismic risk assessment considers the full range of earthquakes, of both lesser and 

greater intensity than the SSE.

Summary—Initiating Events 

Based on the above discussion, the initiating events that could result in baffle deformation or 

displacement are the large-break LOCA, medium-break LOCA, and an earthquake equal to or 

exceeding the SSE. 

5. Potential Consequences of BFB Cracking 

BFB cracking has the following potential consequences: 

• Excessive Fuel Grid Impact Loading from Baffle Plate Detachment or Deflection: The 

most significant consequence of a large number of BFB failures is the potential for 

excessive impact loadings on peripheral fuel assemblies from deflection of the baffle 

plates or complete or partial detachment of a baffle plate from the former plates. The 

minimum bolting pattern analyses in Licensee Event Report 16-004-00 (Ref. 1) 

determined the minimum number of bolts necessary to perform their function; it did not 
- 9 - 

determine whether plates detach. Even if plates do not detach, degraded BFBs could 

cause increased deflection of the plates during a LOCA or seismic event, which could 

cause increased fuel grid impact loading. Excessive fuel grid impact is most likely during 

a large LOCA but could occur during a medium-break LOCA or seismic event, 

particularly if defective BFBs are clustered. If fuel grid allowable loads are exceeded, 

grid crush could occur leading to localized degradation of cooling for affected fuel 

assemblies, and the capability to insert control rods in peripheral locations could be 

affected. Such localized degradation to cooling could be evaluated via a coolable 

geometry evaluation. With respect to a LOCA, all Westinghouse plants are licensed for 

leak-before-break (LBB) for the main reactor coolant loop piping, so for some plants, the 

largest line without LBB is a pressurizer surge line, accumulator line, or residual heat 

removal line (14- or 10-inch line). Some Westinghouse plants have gained NRC 

approval through a license amendment for expanded LBB, in which LBB is applied to the 

large branch lines, such that the largest line without LBB is in the range of a 3-inch to 

6-inch diameter line. The NRC-approved minimum bolting pattern methodology in 

WCAP-15029-P-A takes into account LBB such that only line sizes not covered by LBB 

need to be considered for evaluating acceptable BFB patterns. Should a plant be able 

to demonstrate applicability of expanded LBB, the LOCA loads on the baffle plates are 

significantly reduced. For plants to which expanded LBB applies, a fuel grid impact that 

is severe enough to damage non-peripheral fuel assemblies or affect the ability to insert 

a control rod is considered extremely unlikely. Most Westinghouse 4-loop plants do not 

have control rods in peripheral locations or can demonstrate that sufficient margin exists 

such that peripheral rods are not needed for safe shutdown. 

• Baffle Jetting: Baffle jetting can cause localized fuel cladding damage. Failed BFBs 

may allow gaps to form at the corners of the plates that can cause baffle jetting 

(i.e., high-velocity flow through the corner gaps driven by the differential pressure 

between the bypass flow and the main flow regions). Baffle jetting is not considered a 

safety issue because, although localized fuel cladding damage could cause increased 

radioactivity levels in the reactor coolant, the type of damage would be limited to isolated 

cladding damage, not melting. This type of damage, which occurs slowly, would cause 

an increase in coolant activity that would be detectable during normal operation by 

technical-specification-required monitoring of coolant activity, typically to be performed 

every 7 days5. Baffle-edge bolts help prevent the formation of gaps between the plate 

corners, and no degradation of baffle-edge bolts has been found in operating plants, 

which lessens the potential for baffle jetting. Based on OE, there has been no observed 

link between BFB degradation and baffle jetting. 

• Increased Bypass Flow: Core bypass flow is defined as the total amount of reactor 

coolant flow that bypasses the core region, which is not considered effective in the core 

heat transfer process. Gaps at plate corners from degraded BFBs could also cause an 

increase in core bypass flow. The potential core bypass flowpaths in the Westinghouse 

4-loop plants are the baffle-former-barrel region, head cooling spray nozzles, outlet 

nozzles, baffle plate-core cavity gap, and fuel assembly thimble tubes. However, only 

the baffle-former-barrel region bypass flow is significantly affected by the variations in 

the bolting configurations. The vertical baffle plates follow the bounding periphery of the 

core. They are joined to the core barrel by horizontal former plates spaced vertically 

along the baffle plates. For the downflow configuration, the flow that leaks through the 

 

5 Surveillance Requirement 3.4.16.1 is from NUREG-1431, “Standard Technical 

Specifications—Westinghouse Plants: Specifications,” Revision 4, Volume 1, dated April 30, 2012 

(Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A222).
- 10 - 

baffle joints and into the core before it reaches the bottom of the baffle-former-barrel 

region constitutes core bypass flow. Bolting pattern analyses evaluate the increase in 

core bypass flow for normal operation. 

• Fatigue: Failed BFBs could also cause increases in fatigue usage as a result of 

low-cycle fatigue and increased stresses from vibrational fatigue. Both types of fatigue 

are considered long-term issues and are unlikely to lead to failure during normal 

operation over one cycle. 

• Loose Parts: Failed BFBs could also produce loose parts in the form of detached bolt 

heads and locking bars during normal operation. In Westinghouse plants that have 

recently found extensive BFB degradation (i.e., Indian Point Unit 2 and Salem Unit 1), 

locking bar failures have been observed in conjunction with bolt fracture. When a BFB 

completely fractures at the head-to-shank transition, resulting in the separation of the 

bolt head from the shank, either the bolt head can vibrate against the locking bar or the 

baffle plate can locally separate slightly from the former and cause fatigue failure of the 

locking bar, allowing locking bars and bolt heads to become loose parts. The bolt 

shanks remain threaded into the holes and have not been observed to become loose 

parts. It should be noted that the clearance between the baffle plates and peripheral fuel 

assemblies is so small that bolt heads cannot become loose parts unless the fuel is 

removed. Loose bolt heads and locking bars would be small and would not be expected 

to significantly affect core cooling. Detached baffle plates would constitute a large loose 

part, but the potential for travel of these plates is not credible because of the small 

clearance between the plates and the fuel assemblies. In addition, given that clearance, 

when a bolt head separates from the bolt shaft, the bolt head will not fall out of its 

location, even if the locking bar later fails, as described above. Therefore, a bolt head 

trapped in the gap can cause fretting of the adjacent cladding. Localized fuel cladding 

damage caused by fretting can also be detected by monitoring reactor coolant activity. 

Based on the above, the safety-significant potential consequences of bolt degradation are 

detachment or increased deflection of the baffle plates, leading to increased fuel grid impact 

loads, and increases in core bypass flow. 

6. Evaluation of the As-Found Condition—Cracked BFBs 

Attachment 2 describes the risk-informed approach to making decisions regarding emergent 

issues. This approach has five key principles: 

(1) Compliance with Existing Regulations 

(2) Consistency with the Defense-in-Depth Philosophy 

(3) Maintenance of Adequate Safety Margins 

(4) Demonstration of Acceptable Levels of Risk 

(5) Implementation of Defined Performance Measurement Strategies 

The discussion that follows covers each of the five key principles as it relates to the “as-found” 

condition of degraded BFBs. Note that the same key principles are used to differentiate among 

the options considered for addressing the BFB degradation issue; the “Evaluation and 

Assessment of Options” section of this report includes that discussion. 

 
- 11 - 

Compliance with Existing Regulations 

For reactor vessel internals (RVIs), the following regulations are applicable: 

• Title 10 of the Code of Federal Regulations (10 CFR) 50.46, “Acceptance Criteria for 

Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors” 

• 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” 

Appendix A, “General Design Criteria [GDC] for Nuclear Power Plants,” 

o GDC 4, Environmental and Dynamic Effects Design Bases 

o GDC 27, Combined Reactivity Control Systems Capability 

o GDC 35, “Emergency Core Cooling” (considered together with 10 CFR 50.46) 

• 10 CFR 50.55a, “Codes and Standards” 

• 10 CFR 54.21, “Contents of Application—Technical Information” 

The following discusses the relationship of the BFB degradation issue to each of these 

regulations. 

10 CFR Part 50, General Design Criteria (GDC) for Nuclear Power Plants 

The staff notes that potential degradation of BFB does not automatically result in noncompliance 

with the GDC below. A licensee assesses OE related to BFB degradation and evaluates the 

impact on its plant(s) as necessary. If a licensee determines its plant is in noncompliance with 

any of the Commission’s regulations, including the GDC, the licensee is responsible for taking 

appropriate corrective action, which may include shutting down the plant. 

It should be noted that the Atomic Energy Commission published the final rule that added 

10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," in the Federal 

Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In 

accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, 

"SECY-92-223 - Resolution of Deviations Identified during the Systematic Evaluation Program," 

dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided 

not to apply the Appendix A GDC to plants with construction permits issued prior to May 21, 

1971. As a result, a number of operating plants were not licensed to, nor required to meet, the 

GDC. However, since the GDC generally bounds the licensing basis for the pre-GDC plants, 

the staff concludes that use of the current GDC will provide a conservative approach in 

evaluating the impact of BFB degradation on plant safety. 

GDC 4 Environmental and dynamic effects design bases. 

GDC 4 states the following: 

“Structures, systems, and components important to safety shall be designed to 

accommodate the effects of and to be compatible with the environmental conditions 

associated with normal operation, maintenance, testing, and postulated accidents, 

including loss-of-coolant accidents. These structures, systems, and components shall be 

appropriately protected against dynamic effects, including the effects of missiles, pipe 
- 12 - 

whipping, and discharging fluids, that may result from equipment failures and from 

events and conditions outside the nuclear power unit. However, dynamic effects 

associated with postulated pipe ruptures in nuclear power units may be excluded from 

the design basis when analyses reviewed and approved by the Commission 

demonstrate that the probability of fluid system piping rupture is extremely low under 

conditions consistent with the design basis for the piping.” 

The RVI were designed to accommodate these environmental effects, such as loads from 

LOCAs and seismic events. Degradation of BFBs could result in increased stresses on the 

bolts which could exceed the acceptance criteria of the original analyses. However, in 

acceptable bolting pattern analyses reviewed by the NRC staff, licensees have been able to 

show compliance with acceptance criteria for many configurations with a substantial percentage 

of degraded BFBs. Therefore, BFB degradation does not automatically indicate noncompliance 

with GDC 4.

GDC 27 Combined reactivity control systems capability. 

In GDC 27, the NRC states that the reactivity control systems shall be designed to have a 

combined capability, in conjunction with poison addition by the emergency core cooling system, 

of reliably controlling reactivity changes to assure that under postulated accident conditions and 

with appropriate margin for stuck rods the capability to cool the core is maintained. If an 

accident such as a LOCA occurs, failure of degraded BFBs could result in fuel grid allowable 

loads being exceeded and could affect the capability to insert control rods in certain peripheral 

locations. Plants with peripheral control rods may be able to demonstrate sufficient shutdown 

margin without these rods. Therefore, BFB degradation does not automatically indicate 

noncompliance with GDC 27. The licensees of plants that may have BFB degradation are 

responsible for determining operability based on plant-specific configurations and 

assumptions/observations regarding BFB degradation.

10 CFR 50.46 and GDC 35 

In 10 CFR 50.46(b)(4), the NRC states that calculated changes in core geometry shall be such 

that the core remains amenable to cooling. GDC 35 states, “A system to provide abundant 

emergency core cooling shall be provided. The system safety function shall be to transfer heat 

from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad 

damage that could interfere with continued effective core cooling is prevented and (2) clad 

metal-water reaction is limited to negligible amounts.” Degradation of BFBs could cause 

localized damage to peripheral fuel assemblies if an accident such as a LOCA causes fuel grid 

allowable loads to be exceeded. It may be possible to demonstrate that a coolable geometry is 

maintained such that ECCS function is essentially unimpaired even if such damage occurs. 

Other parts of the ECCS, such as piping, pumps, and heat exchangers, are not affected by BFB 

degradation. Therefore, the presence of BFB degradation would not automatically imply 

noncompliance with 10 CFR 50.46 and GDC 35. 

10 CFR 50.55a 

ASME Code Section XI, which is incorporated by reference in 10 CFR 50.55a, requires visual 

VT-3 examinations of all RVIs classified as “core support structures” once each 10-year 

inspection interval. The determination of which RVI components are considered core support 

structures is licensee specific. As a result, some licensees have considered the baffle-former 

assembly and, thus, the BFBs, to be core support structures and others have not. Further, even 
- 13 - 

if the baffle-former assembly is inspected under ASME Code Section XI, the VT-3 inspection is 

not necessarily detailed enough to detect bolt degradation, nor can it detect partially cracked 

bolts because of the location of these cracks at the head-to-shank transition. For this reason, 

the MRP-227-A guidelines for aging management (Ref. 3) specify augmented inspections using 

UT. Therefore, the presence of BFB degradation alone would not imply noncompliance with 

10 CFR 50.55a. 

10 CFR 54.21 

The requirements in 10 CFR 54.21 specify that the aging of systems, structures, and 

components (SSCs) within the scope of license renewal are to be managed to maintain 

intended functions consistent with the current licensing basis for the period of extended 

operation. Inspections specified in MRP-227-A (Ref. 3), along with the associated corrective 

actions, will ensure that the RVIs will be maintained consistent with the current licensing basis 

for the period of extended operation. Therefore, the presence of BFB degradation would not 

imply noncompliance with 10 CFR 54.21. 

Consistency with the Defense-in-Depth Philosophy 

One factor for assessing how an event might degrade defense in depth is to see how it affects 

the balance among the layers of defense. It is useful to consider the following layers of defense 

(successive measures) when evaluating the impact of degraded BFBs on defense in depth: 

(1) a robust plant design to survive hazards and minimize challenges that could result in an 

event occurring 

(2) prevention of a severe accident (core damage) should an event occur 

(3) containment of the source term should a severe accident occur 

(4) protection of the public from any releases of radioactive material (e.g., through siting in 

low-population areas and the ability to shelter or evacuate people, if necessary) 

The likelihood of undetected BFB degradation severe enough to cause baffle plate detachment 

during normal operation is low. As discussed earlier, the plant may be able to continue 

operating under such a condition or would have to shut down due to increased coolant activity. 

For plate detachment to occur during normal operation, extremely severe BFB degradation 

would be required. Such degradation would likely cause cladding damage detectable via the 

RCS activity sampling which is required by technical specifications, before this degradation 

became severe enough to cause baffle failure during normal operation. Degradation of BFBs 

may necessitate a reactor shutdown but does not increase the likelihood of any other event. 

The types of accidents that could cause deflection or detachment of baffle plates when BFBs 

are degraded are a large-break LOCA, a medium-break LOCA, or a seismic event. The SSCs 

that provide key safety functions following such accidents are not expected to be affected by 

BFB degradation. Reactivity control is provided by control rods and soluble poison; the BFB 

issue does not affect the latter. As stated earlier, the failure of the baffle could result in an 

impact on the peripheral fuel assemblies, but these locations do not typically have control rods, 

or they can be demonstrated to be unnecessary for safe shutdown. If degradation of BFBs 

causes localized fuel grid allowable loads to be exceeded, it may be possible to demonstrate 

that a coolable geometry is maintained such that ECCS function is essentially unimpaired even 
- 14 - 

if such damage occurs. Pumps, piping and valves associated with decay heat removal systems 

are not expected to be affected by BFB degradation. 

BFB degradation does not affect the containment structure, such that the containment response 

to these accidents would not be different whether the BFBs were degraded or not. Finally, the 

BFB degradation issue has no effect on the emergency preparedness functions, such that the 

fourth layer of defense is not affected. 

Another factor of defense in depth to consider involves the effect of the issue on the fission 

product barriers (i.e., fuel cladding, RCS pressure boundary, and containment). As stated 

above, BFB degradation does not affect the containment response. BFB degradation could 

affect fuel cladding in peripheral fuel assemblies as a result of fretting or baffle jetting, as 

previously discussed. This would affect only a small percentage of the cladding. If this 

occurred during normal operation, the RCS would remain intact. If baffle failure were the result 

of a LOCA, the damage to the fuel cladding would be isolated to the peripheral fuel assemblies 

and limited to a small fraction of the total. A seismic event that resulted in baffle failure would 

not result in the RCS pressure boundary failure except for earthquakes of very large magnitude, 

which have a very low frequency. 

In summary, BFB degradation does not affect the containment function. The effect on cladding 

is small and limited to peripheral fuel assemblies. The RCS would remain intact except possibly 

for a LOCA or large seismic event, both of which are very unlikely. The BFB degradation issue 

does not indicate that an imminent safety concern exists from any effect on defense in depth, 

because it does not significantly affect the four layers of defense nor the containment fission 

product barrier.

Maintenance of Adequate Safety Margins 

Acceptable bolt pattern analyses, in accordance with the methodology of WCAP-15029-P-A, 

use the allowable stresses from the ASME Code. Acceptance criteria for other types of 

analyses, such as fuel grid crush, core bypass flow, high- and low-cycle fatigue, and baffle 

jetting, are based on vendor design criteria. Degradation of significant numbers of BFBs can 

cause these acceptance criteria to be exceeded, which represents a potential reduction in 

safety margins. 

The safety analysis acceptance criteria for RVIs are plant specific and are documented in each 

plant’s UFSAR. Typically, the safety analysis criteria for RVIs are based on keeping RVI 

deflections within a specified limit during concurrent LOCA and SSE conditions. The analyses 

typically treat the internals as a few major subassemblies (core barrel, upper internals) and do 

not have specific criteria for the baffle-former assembly or BFBs. While BFB degradation 

represents a decrease in the safety margin for the baffles, the corresponding decrease in 

reactor safety is not so significant as to indicate that an imminent safety concern exists. 

Demonstration of Acceptable Levels of Risk 

The primary purpose of the risk assessment of the “as-found” condition is to determine whether 

the risk of the issue is sufficient to create an imminent safety concern, such that prompt action 

must be taken to provide reasonable assurance of adequate protection of public health and 

safety. The bounding discussion below supports a conclusion that the risk of this issue does not 

represent an imminent safety concern. 
- 15 - 

As stated above, baffle deformation or displacement could be expected to result in localized fuel 

cladding damage under three conditions—a medium LOCA, a large LOCA, or a seismic event. 

NUREG-1829 (Ref. 10) estimates plant-wide LOCA frequency. Table 7.19 of that NUREG 

provides the total PWR LOCA frequencies without contribution from steam generator tube 

rupture events. The smallest medium LOCA in that study was taken as 1,500 gpm. The mean 

frequency of a LOCA of that size or larger is given as 5.8x10-4 per year. Because “1,500 gpm or 

larger” includes medium and large LOCA events, it represents an upper bound for LOCA 

challenges to degraded BFBs. As stated previously in the section titled “Evaluation of Events 

that Could Result in Failure of Baffles with Degraded BFBs,” baffle failure due to a LOCA is not 

expected to result in fuel damage but could result in a baffle plate mechanically impinging on the 

cladding, which could result in localized fuel cladding damage. While the LOCA itself may lead 

to core damage, the contribution to core damage cannot be any larger than the frequency at 

which these events are expected to occur and is arguably much smaller because BFB 

degradation does not affect the SSCs designed to mitigate the LOCA, other than localized 

changes to coolable geometry resulting from damage to peripheral fuel assemblies. Since it is 

possible that coolable geometry can be demonstrated in such cases, core damage will not 

necessarily occur even though some peripheral fuel assemblies sustain damage. LIC-504 

provides guidance for when the risk of an emergent condition may indicate an imminent safety 

concern; the acceptance guideline is 1x10-3 per year. Therefore, the risk of BFB degradation 

does not represent an imminent safety hazard from potential LOCAs. 

Attachment 3 presents a notional seismic assessment of degraded BFBs. A bounding seismic 

hazard curve was generated that should envelop any nuclear plant in the United States. A 

generic fragility value for RVIs, taken from one of the NRC’s simplified plant analysis risk 

(SPAR) models, was used to determine a “baffle failure frequency,” assuming no degradation. 

Then an assumed percent degradation in the median capacity of the BFBs was used to 

compute a new failure frequency for the baffles. The 1x10-3 per year threshold would be 

approached if the median capacity of the BFBs were reduced by 75 percent; that is, with only 

25-percent median capacity, the baffle failure frequency would be near the guideline that would 

indicate a possible imminent safety hazard. Such a large reduction in median capacity, before 

the condition is detected and corrected, is considered unlikely because it would require levels of 

bolt degradation far in excess of that seen in the worst affected units. The worst affected units 

observed that approximately 25% of its bolts were degraded, some of which likely retain some 

load bearing capacity. The staff considers it very unlikely that plants in the most susceptible 

group that have not yet inspected the BFB would have a 50% greater level of degradation then 

has been observed to date, given the relatively similar operating time (EFPY) for the group of 

plants. Also, as stated above, baffle deformation or displacement does not automatically lead to 

core damage or fuel melt but could result in localized fuel cladding damage. 

A seismic event can also induce a LOCA. The analysis in Attachment 3 concludes that the 

frequency of such events would be about two orders of magnitude less than the frequency of a 

medium or larger LOCA. Therefore, these events do not contribute significantly to BFB 

degradation risk. 

LIC-504 also includes an acceptance guideline for when a large early release frequency (LERF) 

attributable to an emergent condition might indicate a potential imminent safety concern. The 

dominant contributors to a PWR LERF are bypass events, such as an intersystem LOCA and 

steam generator tube ruptures. The BFB issue does not affect either of these scenarios. 

However, the occurrence of one of these scenarios could lead to baffle failure, which could 

result in some localized fuel cladding damage, even in scenarios where fuel melt was avoided. 
- 16 - 

Typical initiating event frequencies for intersystem LOCAs are in the mid-10-6 per year range. 

Table 7.18 in NUREG-1829 (Ref. 10) estimates steam generator tube rupture events with 

leakage rates greater than 1,500 gpm to occur at a frequency of 6.5x10-6 per year. Therefore, a 

bounding estimate of LERF from intersystem LOCAs and steam generator tube ruptures is well 

below the 1x10-4 per year guideline in LIC-504, so that the LERF attributable to BFB 

degradation would also not indicate an imminent safety concern. 

Implementation of Defined Performance Measurement Strategies 

All the licensees of plants with high susceptibility to BFB degradation have at least verbally 

committed to conduct UT examinations of their BFBs by the next refueling outage. As 

previously stated, the EPRI-MRP interim guidance (Ref. 6) calls for these examinations. 

Therefore, the time frame for potentially elevated risk is short (less than 18 months); these 

plants will have either fall 2016, spring 2017, or fall 2017 refueling outages. 

The UT inspection specified by MRP-227-A for BFBs is the most effective means of 

performance monitoring of BFBs. Visual VT-3 inspections, which are specified by the ASME 

Code, Section XI, for core support structures, may detect broken BFBs. However, these 

strategies require a reactor shutdown and defueling. For online performance monitoring, the 

most effective method is RCS activity monitoring, which can detect an increase in coolant 

activity that may be indicative of fuel cladding failures. The most likely mechanism for fuel 

cladding failures related to BFB degradation is fretting by loose bolt heads. BFB degradation 

could theoretically lead to baffle jetting, which could also cause fuel cladding failures. However, 

there is currently no OE linking BFB degradation to baffle jetting. 

Loose parts monitoring is another means of monitoring BFB performance online. Loose parts 

monitoring systems may not be sufficiently sensitive to detect bolt heads and lock bars. The 

systems may detect loose plates. However, the staff considers loose parts monitoring to have 

limited effectiveness for detecting BFB degradation. 

 
- 17 - 

OPTIONS CONSIDERED 

Decision Options 

# Option Evaluation and Criteria Used 

1 Require immediate 

shutdown of potentially 

susceptible plants for UT 

examination or 

replacement of minimum 

bolt pattern. 

This option would be required if an imminent safety concern 

were identified, such as the following: 

• Defense in depth is significantly degraded. 

• There is significant loss of safety margin. 

• CDF or conditional CDF (CCDF) is high (e.g., greater 

than or on the order of 1x10-3/yr). 

• LERF or conditional LERF (CLERF) is high (e.g., greater 

than or on the order of 1x10-4/yr). 

2 Require potentially 

susceptible plants to 

perform UT 

examinations or replace 

minimum bolt pattern by 

the end of the next 

refueling outage. 

This option would be appropriate if the issue is not an 

imminent safety concern but the following are true: 

• Margin has been substantially reduced. 

• Condition gets worse over time. 

• Uncertainty is high such that margin needs to be restored in 

relatively short time frame. 

3 Develop a generic 

communication and plan 

to gather additional 

information to support a 

future need for further 

regulatory action. 

This option would be appropriate if the issue is not an 

imminent safety concern and the following are true: 

• Margin, while below codes and standards, is likely 

sufficient. 

• The condition is relatively stable; that is, not worsening over 

time. 

• Additional information would be useful before making a 

regulatory decision. 

4 Maintain status quo. This option would be appropriate if the issue is not an 

imminent safety concern and the following are true: 

• Margin is judged acceptable. 

• The condition is stable; that is, not worsening over time. 

• Any new examples of BFB degradation will be bounded by 

current operating experience. 

EVALUATION AND ASSESSMENT OF OPTIONS 

As stated above, Attachment 2 summarizes the risk-informed approach to decisionmaking. 

Each of the identified options is discussed below in terms of the five key principles. Other 

factors are also considered, as warranted. 

Option 1—Require immediate shutdown of potentially susceptible plants for UT 

examination or replacement of minimum bolt pattern. 

Synopsis—Under this option, the NRC would use a regulatory process, such as an order, to 

compel licensees of plants with high susceptibility to BFB degradation to shut down until UT 

examinations are conducted to determine the condition of the BFBs and to remain shut down 

until necessary bolt replacements are performed to restore adequate safety margins. 

Alternatively, licensees could replace a minimum pattern of bolts to meet all design 

requirements for normal and accident conditions, or take other corrective actions, such as an 

upflow conversion modification. These plants would also be required to justify the time of 
- 18 - 

operation until the next inspection of the BFBs, if some original bolts with no UT indications 

were left in service and relied upon for functionality of the baffle-former assembly. Licensees 

would need to justify an adequate safety margin to account for additional service-induced 

failures of original bolts. 

Compliance with Existing Regulations 

Immediate shutdown and inspection would find any degraded BFBs, and licensees would be 

required to replace a sufficient number of BFBs to ensure design limits would be met for all 

normal and design-basis operating conditions. With the baffle-former assembly restored to a 

fully qualified condition, compliance with the regulations related to the ECCS (GDC 35 and 

10 CFR 50.46) would be ensured, since these regulations implicitly assume the baffle-former 

assembly is capable of performing its intended function. 

Consistency with the Defense-in-Depth Philosophy 

Option 1 is consistent with the defense-in-depth philosophy because it would ensure BFB 

degradation is detected and corrected before the plant starts up. Thus, the likelihood of core 

damage is minimized with this option and all three barriers to fission product release (fuel 

cladding, reactor coolant pressure boundary, and containment) will have the best chance of 

remaining intact if a LOCA or SSE occurs. 

Maintenance of Adequate Safety Margins 

Since degraded BFBs would result in degraded safety margins, immediate shutdown for UT 

inspection would determine the extent of degradation. Corrective actions would restore or 

demonstrate adequate safety margins by the replacement of some or all of the degraded bolts, 

combined with an analysis of the as-left condition to demonstrate that safety margins are 

maintained. Therefore, Option 1 would ensure that adequate safety margins are restored. 

Demonstration of Acceptable Levels of Risk 

The bounding risk assessment of operating with significant BFB degradation indicates that this 

issue does not rise to the level of an imminent safety concern because the estimated CDF is 

less than 1x10-3 per reactor year and the estimated LERF is less than 1x10-4 per reactor year, 

as detailed previously. The level of risk for Option 1 would represent no change from the 

condition of having non-degraded BFBs because the susceptible plants would be shut down 

immediately. 

Implementation of Defined Performance Measurement Strategies 

Implementation of Option 1 would allow UT examinations, which are the most effective 

performance monitoring strategy for BFB degradation. 

Option 2—Require potentially susceptible plants to perform UT examinations or replace 

minimum bolt pattern by next refueling outage. 

Synopsis—This option is identical to Option 1, except plants would be allowed to implement the 

inspection or other corrective action at the next refueling outage rather than be required to shut 

down immediately. The industry has recommended this course of action voluntarily. MRP 

Letter 2016-022 (Ref. 6) transmitted interim guidance from the EPRI MRP that recommended 
- 19 - 

that all the plants in the most susceptible group (Westinghouse 4-loop, downflow, with 

Type 347 stainless steel BFBs) perform a UT volumetric inspection of all BFBs at the next 

scheduled refueling outage. For these plants, the next refueling outage is no later than 

fall 2017. This guidance is classified as “needed” under the industry materials protocol, 

NEI 03-08 (Ref. 11). In accordance with NEI 03-08, if a utility wishes to deviate from “needed” 

guidance, it must have a technical justification, which is sent to the responsible issue program 

within 45 days of approval; the deviation must be entered in the utility’s corrective action 

program; an internal or external independent review must be performed; and the responsible 

utility executive must concur. The NRC must be notified of the deviation at the same time that 

the technical justification is sent to the issue program. If the NRC becomes aware that a 

licensee intended to deviate from the interim guidance on BFBs, it could take regulatory action 

to ensure that the licensee inspected its BFBs at the next refueling outage. However, the staff 

does not anticipate a deviation by any of the licensees of the susceptible plants. Therefore, it is 

likely that use of a regulatory process to require implementation of Option 2 will not be 

necessary. 

Compliance with Existing Regulations 

Implementation of this option results in licensees for the most susceptible plants not knowing 

the physical integrity of the BFBs until the next scheduled refueling outage. Inspection would 

identify any degraded BFBs and licensees would be expected to replace a sufficient number of 

BFBs to ensure design limits will be met for all normal and design-basis operating conditions. 

With the baffle-former assembly restored to a fully qualified condition, compliance with the 

regulations related to the ECCS (GDC 35 and 10 CFR 50.46) would be ensured, since these 

regulations implicitly assume the baffle-former assembly is capable of performing its intended 

function. 

Consistency with the Defense-in-Depth Philosophy 

Defense in depth could potentially be degraded for these plants until the inspection is conducted 

or replacements are performed, or both, because localized fuel cladding damage could occur 

during normal operation from baffle jetting, fretting, or baffle plate impact loads on the fuel 

during a LOCA or seismic event. However, based on the determination of acceptable risk for 

this option, the potential reduction in defense in depth is acceptable. 

Maintenance of Adequate Safety Margins 

As discussed above, operation with degraded BFBs could mean safety margins are degraded. 

However, based on the observed extent of condition for the two plants, even with clustering of 

degraded bolts, it is not likely that the margins would be so eroded as to result in deflection or 

displacement of the baffle. Because of the gradual nature of the degradation, safety margins 

will be adequately maintained under this option. 

 

Demonstration of Acceptable Levels of Risk 

As stated in Option 1, operating with degraded BFBs has acceptably low risk levels. In addition, 

BFB degradation occurs slowly over time; thus, any increase in risk attributable to BFB 

degradation is judged an acceptable increase for the relatively short-term operation of one fuel 

cycle or less. 

 
- 20 - 

Implementation of Defined Performance Measurement Strategies 

As discussed above, monitoring of RCS activity and loose parts is an online performance 

measurement strategy that could give some indication of BFB degradation. Such an online 

performance measurement strategy is acceptable, given the acceptably low risk associated with 

this option. For susceptible plants, the most effective performance monitoring strategy, UT 

examination, will be performed at the next refueling outage. 

Option 3—Develop a generic communication and plan to gather additional information to 

support a future need for further regulatory action. 

Synopsis—Under this option, the NRC staff would issue a generic communication, such as a 

generic letter or bulletin, to request information from licensees. Such a generic communication 

would likely request confirmation of the type of plant design (upflow/downflow), upflow 

conversion, bolt material, previous inspections of BFBs, and planned implementation dates for 

UT examinations of the BFBs. The generic communication could request a justification that the 

planned implementation dates for the examinations are adequate, given the operating 

experience with extensive bolt degradation in certain plants. 

Compliance with Existing Regulations 

Implementation of this option results in licensees not knowing the physical integrity of the BFB 

for an indeterminate period. Once the licensee performs the UT examination or bolt 

replacements, or both, restoring the baffle-former assembly to a fully qualified condition, 

compliance with the regulations related to the ECCS (GDC 35 and 10 CFR 50.46) would be 

ensured, since these regulations implicitly assume the baffle-former assembly is capable of 

performing its intended function. 

Consistency with the Defense-in-Depth Philosophy 

Defense-in-depth could be degraded for these plants until the inspection or replacements are 

performed because cladding damage could occur during normal operation either due to baffle 

jetting or to baffle plate impact loads on the fuel during a LOCA or seismic event. Based on the 

determination of acceptable risk for this option, the potential reduction in defense-in-depth is 

acceptable for the short term but may not be acceptable for longer term operation. 

 Maintenance of Adequate Safety Margins 

As discussed above, operation with degraded BFBs could mean safety margins are degraded. 

While it is true that BFB degradation occurs gradually over time, there could be operating plants 

with more severe degradation than that observed to date. 

Demonstration of Acceptable Levels of Risk 

As stated in Option 1, operating with degraded BFBs has acceptably low risk levels. In addition, 

BFB degradation occurs slowly over time. For Option 3, it is not clear how long susceptible 

plants would operate before UT examinations would be conducted. Additional time to resolve 

this issue increases the uncertainty with how this issue might progress in the future and with the 

associated level of risk. Therefore, the increase in CDF may not be acceptable for longer term 

operation. 
- 21 - 

Implementation of Defined Performance Measurement Strategies 

As discussed above, monitoring RCS activity and loose parts is an online performance 

measurement strategy that could give some indication of BFB degradation and is acceptable, 

given the low risk associated with this option. Under Option 3, it is not clear when the most 

effective performance monitoring strategy, UT examination, will be conducted for susceptible 

plants. 

Option 4—Maintain status quo. 

Synopsis—Under this option, the NRC staff would take no action to require that the BFB margin 

to baffle failure be restored. Response to the operating experience with BFB degradation would 

be left to the discretion of the EPRI MRP and individual licensees. Licensees would be free to 

perform BFB inspections in accordance with the original recommended schedule in MRP-227-A 

(Ref. 3). Note that EPRI has already issued interim guidance for BFB inspections for the 

susceptible group of plants (Ref. 6). 

Compliance with Existing Regulations 

Implementation of this option results in licensees not knowing the physical integrity of the BFBs 

for an undetermined period. Once UT examinations or bolt replacements are performed, 

restoring the baffle-former assembly to a fully qualified condition, compliance with the 

regulations related to the ECCS (GDC 35 and 10 CFR 50.46) would be ensured, since these 

regulations implicitly assume the baffle-former assembly is capable of performing its intended 

function. 

Consistency with the Defense-in-Depth Philosophy 

Defense-in-depth could potentially be degraded for these plants until the inspection or 

replacements are performed, because localized fuel cladding damage could occur during 

normal operation from baffle jetting or to baffle plate impact loads on the fuel during a LOCA or 

seismic event. Based on the determination of acceptable risk for this option, the potential 

reduction in defense-in-depth is acceptable for the short term but may not be acceptable for 

longer term operation. 

Maintenance of Adequate Safety Margins 

As discussed above, operation with degraded BFBs could mean safety margins are degraded. 

While it is true that BFB degradation occurs gradually over time, there could be operating plants 

with more severe degradation than that observed to date. This option relies on normal 

inspections to identify plants with significant degradation. The amount of time until plants have 

been inspected could be longer than the time to gather information in Option 3. Waiting long 

periods of time to gather additional information results in greater exposure for a LOCA or 

seismic event or failure during normal operation. 

 

Demonstration of Acceptable Levels of Risk 

As with Option 3, it is not clear how long susceptible plants would operate before conducting UT 

examinations. Additional time to resolve this issue increases the uncertainty with how this issue 

might progress in the future, and the associated level of risk. Therefore, the increase in CDF 

may not be acceptable for longer term operation. 
- 22 - 

Implementation of Defined Performance Measurement Strategies 

As discussed above, monitoring RCS activity and loose parts is an online performance 

measurement strategy that could give some indication of BFB degradation and is acceptable, 

given the low risk associated with this option. Under Option 4, susceptible plants may not 

conduct the most effective performance monitoring strategy, UT examinations, until 3–10 years 

from 2016. 

Summary of Option Assessment 

The following table summarizes the four options against the five key principles of risk-informed 

regulation. 

Comparison of Options for Potentially Susceptible Plants 

Key Principle 

Option 1—

Immediate 

Shutdown 

Option 2—

Restore to 

Acceptable 

Level within 2 

Years 

Option 3—

Generic 

Communication 

& Information 

Gathering 

Option 4— 

Status Quo 

Regulations No compliance 

issues are 

identified. 

No compliance 

issues are 

identified. 

No compliance 

issues are 

identified. 

No compliance 

issues are 

identified. 

Defense in Depth There is no 

impact, once 

plants are shut 

down and fully 

restored before 

startup. 

There is a minor 

impact given the 

low likelihood of 

baffle failure 

before the next 

refueling outage. 

It may be 

degraded if baffle 

failure results in 

localized fuel 

cladding damage 

during normal 

operation. 

This is the same 

as Option 3. 

Safety Margins The extent of 

degradation would 

be determined 

and margin 

restored before 

any events occur. 

There would be 

no additional 

erosion of safety 

margins. 

There is a minor 

potential for 

further reduction 

in safety margins 

in this option, but 

this is not 

considered 

significant for 

such a short 

interval. 

This is similar to 

Option 2, except 

there is increased 

uncertainty about 

how much margin 

will remain over 

the longer period 

to gather 

information. This 

represents an 

increased 

exposure to a 

LOCA or seismic 

event. 

This is similar to 

Option 3 

Risk This condition is 

judged to 

represent a very 

small increase in 

risk over the “asfound” condition, 

because plants 

would be shut 

down in the near 

term. 

Degradation 

occurs slowly over 

time, so Option 2 

is judged 

acceptable from a 

risk perspective. 

The risk during 

normal operations 

remains small, 

and the exposure 

The longer time 

frame to resolve, 

compared to 

Option 2, results in 

increased 

uncertainty and 

risk exposure. 

This would result 

in an even longer 

time frame and 

greater 

uncertainty than 

Option 3 
- 23 - 

Comparison of Options for Potentially Susceptible Plants 

Key Principle 

Option 1—

Immediate 

Shutdown 

Option 2—

Restore to 

Acceptable 

Level within 2 

Years 

Option 3—

Generic 

Communication 

& Information 

Gathering 

Option 4— 

Status Quo 

to a possible 

LOCA or seismic 

event is limited to 

one cycle.

Monitoring This is not 

applicable, since 

susceptible plants 

would shut down 

in the near term. 

RCS activity and 

loose parts would 

be monitored. UT 

examinations 

would be 

conducted at the 

next refueling 

outage. 

This is the same 

as Option 2, 

except the 

schedule for the 

next UT, which is 

the best 

monitoring 

method, is 

indeterminate. 

As in Option 3, UT 

would be delayed 

as much as 

10 years for some 

susceptible 

plants. 

RECOMMENDATION 

The NRC staff recommends Option 2 as the one that restores margin in a timely manner with 

fewer burdens on licensees than Option 1. Options 3 and 4 are not preferred, in part because 

of the extended time that plants may operate with BFBs that continue to degrade. More detailed 

discussion follows. 

Option 1, under which immediate shutdown and inspection or bolt replacement would be 

required, is the option that ensures the following five key regulatory principles are maintained: 

(1) compliance with existing regulations 

(2) consistency with the defense-in-depth philosophy 

(3) maintenance of adequate safety margins 

(4) demonstration of acceptable levels of risk 

(5) implementation of defined performance measurement strategies 

While Option 1 would ensure that the risk associated with degraded BFBs is minimized, 

comparison with the LIC-504 guidelines does not indicate that this issue represents an imminent 

safety hazard. Therefore, an immediate shutdown is not required. 

Option 2, which would allow inspection or bolt replacement at the next refueling outage, may 

result in some degradation in defense-in-depth because of the potential for localized fuel 

cladding damage if a LOCA or SSE occurs before the inspection and result in a possible 

reduction in safety margins because the plant would operate with potentially degraded bolts. 

Because of the relatively short period of operation with potentially degraded BFBs and the 

staff’s evaluation that the risk of core damage from BFB degradation over this time period is low, 

Option 2 is acceptable, based on risk. 

Option 3, which would involve issuing a generic communication to gather information, and 

Option 4, which would maintain the status quo, may also result in some degradation of defense 

in depth and reduced safety margins, but the risk change per reactor year is assumed to be the 

same as for Option 2. Note, however, that risk-informed decisions should be made with 
- 24 - 

consideration of uncertainty. It is not known, for example, whether there may be currently 

operating plants with more severe BFB degradation than that observed in the two recent 

examples. As a result, the NRC staff is not comfortable with allowing an indeterminate time 

period before the UT inspections are conducted. The difference is that for Options 3 and 4, the 

increase in CDF would be for a longer time period than for Option 2. 

Limited tooling currently exists in the nuclear industry to perform the UT examinations and bolt 

replacements. Therefore, if Option 1 were chosen, it is likely that many of the susceptible plants 

would remain shut down for an extended period of time while waiting for tooling. Given that 

Option 2 has an acceptably low increase in CDF associated with it, imposing Option 1 would 

likely represent an unnecessary hardship on the licensees, considering the extended outages 

that could result. Although Option 3 and Option 4 initially have the same increase in CDF per 

reactor year as Option 2, the time frame under which inspections and bolt replacements would 

occur under these options is extended, so the increase in CDF becomes more uncertain with 

time. Based on these considerations, the staff considers Option 2 to be the best option for 

addressing potential BFB degradation. 

MRP Letter 2016-022 (Ref. 6) transmitted interim guidance from the EPRI MRP that 

recommended that the plants in the most susceptible group (Westinghouse 4-loop, 

downflow, with Type 347 stainless steel BFBs) conduct a UT examination of all BFBs at the 

next scheduled refueling outage. For these plants, the next refueling outage is no later than 

fall 2017. This guidance is classified as “needed” under the industry materials protocol, 

NEI 03-08 (Ref. 11). The NRC staff will review the interim guidance and document its review in 

a publicly available safety assessment. The staff does not expect that any of the plants in the 

susceptible group will deviate from the interim guidance. Therefore, since the industry is 

essentially mandating Option 2 through the NEI 03-08 materials protocol, it is not expected that 

the staff will need to take regulatory action to ensure that licensees implement Option 2. 

However, NEI 03-08 requires the NRC to be notified of deviations from “needed” guidance. 

Therefore, if any of the plants in the most susceptible group intends to deviate from the interim 

guidance, the NRC would be notified and could take regulatory action to ensure that the plant 

would conduct a UT examination at the next refueling outage. 

 
- 25 - 

REFERENCES 

1. Licensee Event Report (LER) 2016-004-00, “Unanalyzed Condition due to Degraded 

Reactor Baffle-Former Bolts,” Indian Point Unit No. 2, Docket No. 50-247, DPR-46, 

May 31, 2016 (ADAMS Accession No. ML16159A219). 

2. LER 272/2016-001-00, “Unanalyzed Condition due to Degraded Reactor Baffle to Former 

Bolts,” Salem Nuclear Generating Station Unit 1, Renewed Facility Operating License 

No. DPR-70, NRC Docket No. 50-272, July 5, 2016 (ADAMS Accession No. ML16187A330). 

3. “Materials Reliability Program: Pressurized Water Reactor Internals Inspection and 

Evaluation Guidelines (MRP-227-A),” 2011 Technical Report, December 2011, transmitted 

to the NRC by MRP letter MRP-2011-036, dated January 9, 2012 (ADAMS Accession 

No. ML120170453). 

4. Westinghouse Technical Bulletin TB-12-5, “Baffle Bolt Degradation in a Westinghouse 

NSSS Plant with Downflow Reactor Internal Design,” March 7, 2012. 

5. Westinghouse Nuclear Safety Advisory Letter (NSAL)-16-1 Revision 1, “Baffle-Former 

Bolts,” Westinghouse Electric Co. LLC, August 1, 2016. 

6. Letter from Bernie Rudell and Anne Demma to the NRC, Subject: “Transmittal of NEI-03-08, 

‘Needed’ Interim Guidance Regarding Baffle Former Bolt inspections for Tier 1 plants as 

Defined in Westinghouse NSAL-16-01,” EPRI Materials Reliability Program, transmitted by 

letter MRP 2016-022, July 27, 2016 (ADAMS Accession No. ML16211A054). 

7. Westinghouse Topical Report ENT000654, WCAP-15270, “Determination of Acceptable 

Baffle-Barrel-Bolting for Four-Loop Westinghouse Downflow and Converted Upflow 

Domestic Plants,” November 30, 2001 (ADAMS Accession No. ML15222B242). 

8. Westinghouse Topical Report WCAP-15029-P-A, “Westinghouse Methodology for 

Evaluating Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load 

Conditions,” Revision 1, December 31, 1998 (ADAMS Legacy Library Accession 

No. 9905070164). 

9. ENT000655—Westinghouse, WCAP 15030-NP-A, “Westinghouse Methodology for 

Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions under Faulted Load 

Conditions,” March 2, 1999. 

10. NUREG-1829, “Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the 

Elicitation Process,” U.S. Nuclear Regulatory Commission, April 2008 (ADAMS Accession 

No. ML080630013). 

11. NEI 03-08, “Guideline for the Management of Materials Issues,” Revision 2, Nuclear Energy 

Institute, January 2010 (ADAMS Accession No. ML101050337). 

ATTACHMENTS 

Attachment 1: Design and Function of Baffle-Former Assembly 

Attachment 2: Principles of Risk-Informed Decisionmaking 

Attachment 3: Seismic Risk Assessment of Degraded Baffle-Former Bolts 
 

Attachment 1 

ATTACHMENT 1 

DESIGN AND FUNCTION OF BAFFLE-FORMER ASSEMBLY 

Baffle-former bolts (BFBs) are part of the baffle-former assembly (Figures 1 and 2). BFBs secure 

the vertical baffle plates to the horizontal former plates, which in turn are bolted to the core barrel 

by the barrel-former bolts. The baffle plates thus form a shroud around the core that closely 

follows the outline of the core. The main function of the baffle plates is to direct and concentrate 

the flow of coolant though the core. Baffle plates span the whole height of the active core and vary 

in width. The plates are 1.5-inch-thick, Type 304 stainless steel. There are eight horizontal rows 

of BFBs, with the number of bolts across the plate varying with the plate width, from 2 bolts on the 

narrowest plates to 12 on the largest plates. There are also edge bolts that secure the corners of 

the plates together to help minimize gaps at the plate corners. Figure 3 shows a cross section of 

the baffle-former assembly and core barrel showing the locations of these various bolts. Figure 4 

shows the appearance of a BFB head as seen from inside the core baffle. Figure 5 contains a 

plan outline of the BFBs for one plant, showing the variation in plate size. 

The main flowpath is down the downcomer outside the core barrel then up through the lower core 

plate and through the core (Figure 6). A small portion of the flow (bypass flow) is directed between 

the core barrel and the baffle plate for cooling purposes. In some plants, this flow enters through a 

hole near the top of the core barrel and flows downward, joining the main core flow when it exits 

through a gap between the baffle plates and the lower core plate. This is referred to as a 

“downflow” configuration. In other plants, mostly newer Westinghouse units, the bypass flow 

enters at the bottom and flows upward, parallel to the main core flow (“upflow” configuration). 

Figure 6 shows both flow configurations. 
- 2 - 

Figure 1 Typical Westinghouse reactor vessel internals general arrangement and 

major components 
- 3 - 

Figure 2 Typical Westinghouse baffle-former assembly 

Figure 3 Cross section of one octant of a typical Westinghouse baffle-former 

assembly showing locations of bolt types 
- 4 - 

Figure 4 Appearance of BFB head as seen from core side 

Figure 5 Schematic of core cross section from a Westinghouse 4-loop plant updated final 

safety analysis report, showing variation in width of baffle plates 
- 5 - 

Figure 6 Schematic of flowpaths in Westinghouse-designed reactor internals, 

comparing downflow and upflow configurations 
 

Attachment 2

ATTACHMENT 2 

PRINCIPLES OF RISK-INFORMED DECISIONMAKING 

U.S. Nuclear Regulatory Commission (NRC) Office Instruction LIC-504, Revision 4, “Integrated 

Risk-Informed Decision-Making Process for Emergent Issues,” effective June 2, 2014 

(Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML14035A143), includes guidance on how to use the five key principles of risk-informed 

decisionmaking to evaluate an emergent issue. The discussion below is excerpted from that 

document for the convenience of the reader. 

As defined in the Staff Requirements Memorandum for SECY-98-0144, “White Paper on RiskInformed and Performance-Based Regulation,” dated March 1, 1999 (Ref. 2-1), a risk-informed 

approach to regulatory decisionmaking represents a philosophy whereby risk insights are 

considered together with other factors to establish requirements that better focus licensee and 

regulatory attention on design and operational issues commensurate with their importance to 

public health and safety. For nuclear power reactors, five key principles of risk-informed 

regulation serve as the foundation of sound risk-informed decisionmaking: 

(1) compliance with existing regulations 

(2) consistency with the defense-in-depth philosophy 

(3) maintenance of adequate safety margins 

(4) demonstration of acceptable levels of risk 

(5) implementation of defined performance measurement strategies 

Various risk-informed processes generally differ in the manner by which they demonstrate that 

the key principles are satisfied. When LIC-504 is employed, the five key principles are used to 

differentiate among various options for resolving the issue. Use of the key principles in this way 

is contrasted with their use in Regulatory Guide (RG) 1.174, “An Approach for Using 

Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the 

Licensing Basis,” Revision 2, issued May 2011 (Ref. 2-2), in the following paragraphs. 

Compliance with existing regulations (unless an exemption is sought) is fairly straightforward 

when a licensee requests a change to the licensing basis of its facility—the subject of RG 1.174. 

However, NRC regulations may not fully address the potential issues revealed when new 

information comes to light. In some cases, the new information may reveal that a licensee is not 

in compliance with a regulation. A risk-informed decision may involve choosing among options 

that involve how long a situation may exist before action must be taken to achieve full regulatory 

compliance. 

Note that there is a presumption of adequate protection of public health and safety afforded by 

compliance with the NRC’s regulations. New information about an unforeseen or substantially 

more likely hazard could raise “special circumstances” that could rebut this presumption. 

Appendix D to Section 19.2, “Review of Risk Information Used to Support Permanent PlantSpecific Changes to the Licensing Basis: General Guidance,” of NUREG-0800, “Standard 

Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR 

Edition,” issued June 2007 (Ref. 2-3), provides additional information on “special 

circumstances.” 

Consistency with the defense-in-depth philosophy in RG 1.174 starts with the presumption that 

defense in depth is adequate for the existing plant and evaluates the requested change to 
- 2 - 

determine whether defense in depth is maintained. For emergent issues, this key principle is 

used to compare various options proposed for addressing the issue. In some cases, defense in 

depth may be at the heart of the issue—that is, the new information may reveal insufficient 

defense in depth for the existing facility. 

Maintenance of adequate safety margins in RG 1.174 includes compliance with codes and 

standards and meeting the safety analysis acceptance criteria in a plant’s licensing basis. The 

LIC-504 approach to this key principle is no different. Appendix D to this office instruction 

provides questions to assist in evaluating how well each proposed option for addressing an 

emergent issue maintains safety margins. 

Demonstration of acceptable levels of risk in RG 1.174 concentrates on the change in risk 

associated with the proposed amendment to a plant’s licensing basis. For emergent issues, the 

total risk associated with the new or increased hazard may provide a basis for concluding that 

an immediate safety concern exists, such that the NRC must take prompt action to put a nuclear 

power plant in a safe condition. The risk associated with each proposed option for resolving the 

issue, in conjunction with the other key principles, may help the decision authority in choosing 

from among the options. Note that a quantitative risk assessment may prove difficult to obtain 

for many emergent issues. Qualitative risk assessments may be employed, as appropriate. 

Implementation of defined performance measurement strategies in RG 1.174 means to ensure 

monitoring is put in place so that (1) the desired outcomes of the change are obtained and 

(2) unintended adverse consequences of the change can be detected in time to take corrective 

action. This is the same for LIC-504. The decision regarding an emergent issue may apply for 

a finite time period (e.g., until a licensee can effect a permanent repair), so that this office 

instruction includes compensatory measures within the scope of this key principle. Appendix D 

to the licensing instruction includes guidance for performance monitoring and compensatory 

measures. 

Attachment 2 References 

2-1 Staff Requirements Memorandum SRM-SECY-98-0144, “White Paper on Risk-Informed 

and Performance-Based Regulation,” U.S. Nuclear Regulatory Commission, 

Washington, DC, March 1, 1999 (ADAMS Accession No. ML003753601). 

2-2 Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in 

Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 2, 

U.S. Nuclear Regulatory Commission, Washington, DC, May 2011 (ADAMS Accession 

No. ML100910006). 

2-3 NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for 

Nuclear Power Plants: LWR Edition,” Section 19.2, “Review of Risk Information Used to 

Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance,” 

U.S. Nuclear Regulatory Commission, Washington, DC, June 2007 (ADAMS Accession 

No. ML071700658).
 

Attachment 3

ATTACHMENT 3 

SEISMIC RISK ASSESSMENT OF DEGRADED BAFFLE-FORMER BOLTS 

The purpose of this attachment is to document an assessment of the risk of degraded 

baffle-former bolts (BFBs) from a seismic event and to determine whether the BFB degradation 

represents an imminent safety concern for operating U.S. nuclear power plants. 

A number of simplifying assumptions are made to show that the frequency of a seismic event 

resulting in displacement of the baffle itself is less than the 1x10-3 core damage frequency 

(CDF) criterion in LIC-504, Revision 4, “Integrated Risk-Informed Decision-Making Process for 

Emergent Issues,” effective June 2, 2014 (Agencywide Documents Access and Management 

System (ADAMS) Accession No. ML14035A143), which is the risk level at which the staff of the 

NRC would begin to question whether the issue presented an imminent safety hazard. As 

discussed below, BFBs would need to be degraded to the point where only 25-percent capacity 

remained for the seismic “baffle failure frequency” to reach 1x10-3 per year. Baffle failure itself is 

not expected to result in core melt, so the CDF from such degradation is judged to be much less 

than 1x10-3 per year. 

Also, a seismic event may lead to a loss-of-coolant accident (LOCA). Using generic reactor 

coolant system (RCS) pipe fragility information and a bounding seismic hazard curve, the 

analysis below shows that seismic-induced LOCAs are likely to be several orders of magnitude 

below the frequencies for the plant-wide LOCAs given in NUREG-1829, “Estimating Loss-ofCoolant Accident (LOCA) Frequencies Through the Elicitation Process,” issued April 2008. 

Therefore, seismically induced LOCAs do not indicate an imminent safety hazard, either. 

Summary of Approach—Seismically Induced Baffle Failure (without LOCA) 

The annual probability of failure of a structure, system, or component because of a seismic 

event can be estimated by convolving the seismic hazard for the given site with the site-specific 

fragility of the structure, system, or component. Such a calculation requires plant-specific 

information and a great deal of work to obtain the necessary curves (hazard and fragility). This 

simplified approach, which is to judge whether the additional risk attributable to BFB 

degradation approaches 1x10-3 per year, employs the following steps: 

(1) Develop a “bounding” seismic hazard curve for the United States. This was done by 

considering the recent updated seismic hazard submittals from four nuclear power plant 

sites—Sequoyah, Watts Bar, Columbia, and Diablo Canyon (Refs. 3-1, 3-2, 3-3, and 3-4, 

respectively). These sites were chosen for their relatively high seismic hazard. The 

“bounding curve” was taken as the maximum annual exceedance probability from the 

four plant sites between accelerations of 0.02 acceleration due to gravity (g) and 5g. 

This study uses peak ground acceleration. 

(2) Assume a representative fragility curve for the baffle assembly. A fragility curve is 

characterized by two parameters—the median capacity (Am) and the composite 

uncertainty (βc). NUREG/CR-6544, “A Methodology for Analyzing Precursors to 
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Earthquake-Initiated and Fire-Initiated Accident Sequences,” issued April 1998 

(Ref. 3-5), provides generic fragility information in Table 6-1 for “reactor vessel 

internals.” The resulting fragility parameters are Am = 1.8 and βc = 0.5. Use of generic 

fragility parameters ignores the transmission of the seismic energy from the location 

where the seismic hazard was developed (“control point”) to the specific location of the 

RCS piping, but the use of this generic estimate is judged acceptable for this notional 

study (see “assumptions,” below). 

(3) For each earthquake severity “bin” above 0.02g, determine the annual probability of an 

earthquake occurring in that bin (using the bounding hazard curve) and the probability of 

failure of the baffle assembly (its “fragility”) using the average earthquake intensity in that 

bin, where the average is taken as the geometric average of the upper and lower 

acceleration end points. 

(4) For each earthquake severity bin, multiply the annual probability of occurrence by the 

probability of baffle failure for that intensity earthquake. The “baffle failure frequency” is 

given by the sum of the annual failure frequencies of each bin. 

(5) Perform sensitivity studies using reduced values of Am corresponding to a percent 

reduction in the ability of the baffle assembly to withstand a seismic event. 

Summary of Approach—Baffle Failure from a Seismically Induced LOCA 

The approach for determining the frequency of a seismically induced LOCA is similar to that 

described above for baffle assembly failure: 

(1) Use the same bounding hazard curve described above. 

(2) The fragility parameters for piping also come from NUREG/CR-6544 and are taken as 

Am = 3.8 and βc = 0.61. 

(3) The annual frequency of a pipe-break LOCA resulting from a seismic event is estimated 

in the same manner as steps 3 and 4 above for the baffle assembly failure calculation. 

However, in this case, no sensitivity studies are performed, as the overall seismically 

induced LOCA frequency is what is desired for this analysis. 

Assumptions—Both Methods 

The analysis described above makes a number of assumptions. Since the goal of this analysis 

is to determine whether an imminent safety concern might exist, the use of bounding 

assumptions is useful to add confidence to the result. However, not all of the assumptions can 

be demonstrated to be bounding. On the other hand, baffle deformation or displacement is not 

expected to result in core melt, although limited cladding damage might occur if a loose baffle 

plate affected a peripheral fuel assembly. The assumptions include the following: 

• The peak ground acceleration seismic hazard curves were used. No spectral 

information was used. 

• No amplification or attenuation is assumed. 
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• A generic fragility for “reactor vessel internals” was used for the baffle assembly. 

• A generic fragility for “piping” was used to determine when a seismic event might cause 

a LOCA. 

• Baffle failure frequency is used as the metric for the risk of baffle failure in the event of 

an earthquake; the CDF attributable to such an event is expected to be much lower. 

• The convolution integral is approximated by summing over the discrete bins. 

• No failures were assumed for accelerations less than 0.02g. 

Bounding Assessment of Reduction in BFB Capacity because of Degradation 

The purpose of this section is to support the bounding seismic risk analysis by demonstrating 

that, with a reduction in the overall median seismic capacity of the BFBs by 75 percent, it is 

likely that the baffle would remain in place. 

The staff compared and extrapolated analysis results for two different levels of BFB degradation 

in WCAP-15270, “Determination of Acceptable Baffle-Barrel-Bolting for Four-Loop 

Westinghouse Downflow and Converted Upflow Domestic Plants,” dated November 30, 2001, to 

estimate the remaining structural capacity of a case in which only 25 percent of the BFBs were 

intact. Based on this comparison, the staff concluded that acceptance criteria would still be met 

for loads from a safe-shutdown earthquake only, for a case with 25 percent of BFBs intact. 

Based on an examination of as-found patterns of intact and degraded bolting at the two plants, 

overall the highest percentage of degraded bolts found in a reactor is 27 percent. Some 

individual baffle plates at one of the plants had high percentages of the bolts (approaching 

95 percent) potentially degraded, considering visually degraded bolts, untestable bolts, and 

bolts with ultrasonic testing indications. However, even on such plates, it is likely that 

25 percent of load-bearing capacity remains, because untestable bolts are not necessarily 

degraded and many cracked bolts retained some load-bearing capacity. Torque values to 

remove degraded bolts at one of the plants indicated many of those bolts retained significant 

load-bearing capability. In addition, no evidence of edge bolt degradation was found in the 

plants that have been inspected. Therefore, edge bolts probably contribute significant structural 

support to resist displacement of the baffle plates, even with a significant percentage of 

degraded BFBs. 

Based on the above discussion, it is reasonable to assume no more than 75 percent reduction 

in the seismic capacity of the BFBs. 

Results of the Seismic Risk Study 

Using the bounding hazard curve and the generic fragility given by Am = 1.8 and βc = 0.5, there 

is a 3.1x10-5 chance per year that a nondegraded baffle would fail because of a seismic event. 

If the seismic capacity of the baffle assembly were reduced by 50 percent, the frequency of 

baffle failure because of a seismic event would increase to 2.0x10-4 per year, which is an 

increase in failure frequency of 1.7x10-4 per year. If the seismic capacity of the baffle assembly 
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were reduced by 75 percent, such that only 25-percent capacity remained, the new seismic 

failure frequency for the baffle would be 9.2x10-4 per year and the increase over the 

nondegraded case would be 8.9x10-4 per year. This final number is less than but very close to 

the 1x10-3 per year threshold in LIC-504. As stated above, even with as much as 75-percent 

reduction in the overall median seismic capacity, the baffle is likely to remain supported. 

Therefore, it is concluded that, even with 75-percent reduction in the seismic capacity of the 

BFBs due to degradation, this issue does not rise to the level of an imminent safety concern. 

Regarding a seismically induced LOCA, using the bounding hazard curve and the generic 

fragility given by Am = 3.8 and βc = 0.61, taken from NUREG/CR-6544, there is a 6.3x10-6

chance per year of a LOCA because of a seismic event. The plant-wide LOCA frequency 

estimate for a 1,500 gallons per minute or larger LOCA is given in NUREG-1829, Table 7.19, as 

5.8x10-4 per year. The seismically induced LOCA frequency is about two orders of magnitude 

below the nonseismic LOCA frequency, so that this contribution to the risk of BFB degradation 

can be ignored when considering whether an imminent safety concern is indicated by this issue. 

Bounding Seismic Hazard Curve 

Peak Ground Acceleration (g) Exceedance Frequency 

0.02 7.00E-02 

0.05 2.30E-02 

0.1 8.40E-03 

0.15 4.60E-03 

0.2 2.80E-03 

0.25 2.00E-03 

0.3 1.40E-03 
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Bounding Seismic Hazard Curve 

Peak Ground Acceleration (g) Exceedance Frequency 

0.40 7.30E-04 

0.5 4.30E-04 

0.60 2.80E-04 

0.70 1.70E-04 

0.85 8.60E-05 

1. 4.90E-05 

2 8.52E-06 

3. 2.79E-06 

5. 5.43E-07 

 
- 6 - 

Attachment 3 References 

3-1 Letter, J.W. Shea to NRC, “Tennessee Valley Authority’s Seismic Hazard and Screening 

Report (CEUS Sites), Response to NRC Request for Information Pursuant to 

10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of 

Insights from the Fukushima Dai-ichi Accident,” CNL-14-038, March 31, 2014; 

Enclosure 3, “Seismic Hazard and Screening Report for Tennessee Valley Authority’s 

Sequoyah Nuclear Plant,” Table A-1g (ADAMS Accession No. ML14098A478). 

3-2 Letter, J.W. Shea to NRC, “Tennessee Valley Authority’s Seismic Hazard and Screening 

Report (CEUS Sites), Response to NRC Request for Information Pursuant to 

10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of 

Insights from the Fukushima Dai-ichi Accident,” CNL-14-038, March 31, 2014; 

Enclosure 4, “Seismic Hazard and Screening Report for Tennessee Valley Authority's 

Watts Bar Nuclear Plant,” Table A-1g (ADAMS Accession No. ML14098A478). 

3-3 Letter, David A. Swank to NRC, “Columbia Generating Station, Docket No. 50-397, 

Seismic Hazard and Screening Report, Response to NRC Request for Information 

Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task 

Force Review of Insights from the Fukushima Dai-ichi Accident,” GO2-15-045, 

March 12, 2015; Table 2.2.2-2a: “Annual Frequency of Exceedance Mean and Fractile 

Base Rock Seismic Hazard Curves for PGA at Columbia Generating Station 

(PNNL, 2014),” (ADAMS Accession No. ML15078A243). 

3-4 Letter, Barry S. Allen to NRC, “Response to NRC Request for Information Pursuant to 

10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the 

Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident: Seismic 

Hazard and Screening Report,” PG&E Letter DCL-15-035, March 11, 2015; Volume 2, 

Table A2.0-1: “Mean and Fractiles of Hazard for the Control Point for Peak Acceleration” 

(ADAMS Accession No. ML15070A607 (Vol. 1) and ML15070A608 (Vol. 2)). 

3-5 NUREG/CR-6544, “A Methodology for Analyzing Precursors to Earthquake-Initiated and 

Fire-Initiated Accident Sequences,” April 1998 (ADAMS Accession No. ML071650470).