ML16266A224

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Integrated Inspection Report 05000272/2016002 and 05000311/2016002
ML16266A224
Person / Time
Site: Salem  PSEG icon.png
Issue date: 09/22/2016
From: Fred Bower
Reactor Projects Branch 3
To: Sena P
Public Service Enterprise Group
References
IR 2016002
Download: ML16266A224 (66)


See also: IR 05000272/2016002

Text

T. Joyce

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION I

2100 RENAISSANCE BLVD., SUITE 100

KING OF PRUSSIA, PA 19406-2713

September 22, 2016

Mr. Peter Sena, III

President and Chief Nuclear Officer

PSEG Nuclear LLC - N09

P.O. Box 236

Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 -

INTEGRATED INSPECTION REPORT 05000272/2016002 AND

05000311/2016002

Dear Mr. Sena:

On June 30, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at

the Salem Nuclear Generating Station, Units 1 and 2 (Salem). The enclosed inspection report

documents the inspection results, which were discussed with Mr. Robert DeNight on July 28,

2016, and with Mr. Eric Carr on August 11, 2016, as well as other members of your staff.

NRC Inspectors examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed

personnel.

The inspectors documented two findings of very low safety significance (Green) in this report.

Further, inspectors documented a licensee-identified violation which was determined to be of

very low safety significance in this report. The NRC is treating these issues as one finding (FIN)

and as two non-cited violations (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the NCV in this report, you should provide a response within 30 days of the date

of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission,

ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional

Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory

Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Salem. In

addition, if you disagree with the cross-cutting aspect assigned to any finding, or a finding not

associated with a regulatory requirement in this report, you should provide a response within

30 days of the date of this inspection report, with the basis for your disagreement, to the

Regional Administrator, Region I, and the NRC Resident Inspector at Salem.

P. Sena

- 2 -

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390 of the NRCs

Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be

available electronically for public inspection in the NRCs Public Document Room or from

the Publicly Available Records component of the NRCs Agencywide Documents Access and

Management System (ADAMS). ADAMS is accessible from the NRC website at

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Fred L. Bower, III, Chief

Reactor Projects Branch 3

Division of Reactor Projects

Docket Nos. 50-272 and 50-311

License Nos. DPR-70 and DPR-75

Enclosure:

Inspection Report 05000272/2016002 and

05000311/2016002

w/Attachment: Supplementary Information

cc w/encl: Distribution via ListServ

ML16266A224

SUNSI Review

Non-Sensitive

Sensitive

Publicly Available

Non-Publicly Available

OFFICE

RI/DRP

RI/DRP

RI/DRS

RI/DRP

RI/DRP

NAME

PFinney/RB

RBarkley

MGray

MScott

FBower

DATE

9/16/16

9/14/16

9/16/16

9/22/16

9/22/16

1

Enclosure

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket Nos.

50-272 and 50-311

License Nos.

DPR-70 and DPR-75

Report Nos.

05000272/2016002 and 05000311/2016002

Licensee:

PSEG Nuclear LLC (PSEG)

Facility:

Salem Nuclear Generating Station, Units 1 and 2

Location:

P.O. Box 236

Hancocks Bridge, NJ 08038

Dates:

April 1, 2016 through June 30, 2016

Inspectors:

P. Finney, Senior Resident Inspector

A. Ziedonis, Resident Inspector

E. Burket, Emergency Preparedness Specialist

G. DiPaolo, Senior Reactor Inspector

M. Draxton, Project Engineer

J. Kulp, Senior Reactor Inspector

M. Modes, Senior Reactor Inspector

R. Nimitz, Senior Health Physicist

T. OHara, Reactor Engineer

D. Orr, Senior Reactor Inspector

R. Vadella, Project Engineer

J. Poehler, Senior Materials Engineer

Approved By:

Fred L. Bower, III, Chief

Reactor Projects Branch 3

Division of Reactor Projects

2

TABLE OF CONTENTS

REPORT DETAILS ....................................................................................................................... 5

1.

REACTOR SAFETY .............................................................................................................. 5

1R01

Adverse Weather Protection ...................................................................................... 5

1R04

Equipment Alignment .................................................................................................. 7

1R05

Fire Protection ............................................................................................................. 7

1R07

Heat Sink Performance .............................................................................................. 7

1R08

In-service Inspection Activities ................................................................................... 7

1R11

Licensed Operator Requalification Program ............................................................ 12

1R12

Maintenance Effectiveness ...................................................................................... 12

1R13

Maintenance Risk Assessments and Emergent Work Control ................................ 13

1R15

Operability Determinations and Functionality Assessments .................................... 14

1R18

Plant Modifications ................................................................................................... 19

1R19

Post-Maintenance Testing ....................................................................................... 20

1R20

Refueling and Other Outage Activities ...................................................................... 20

1R22

Surveillance Testing ................................................................................................. 21

1EP6

Drill Evaluation ........................................................................................................ 22

2. RADIATION SAFETY .......................................................................................................... 22

2RS1

Radiological Hazard Assessment and Exposure Controls ....................................... 22

2RS2

Occupational ALARA Planning and Controls ........................................................... 24

2RS3

In-Plant Airborne Radioactivity Control and Mitigation ............................................. 25

2RS4

Occupational Dose Assessment .............................................................................. 26

2RS5

Radiation Monitoring Instrumentation ...................................................................... 27

2RS7

Radiological Environmental Monitoring Program (REMP) ....................................... 28

4.

OTHER ACTIVITIES ............................................................................................................ 29

4OA1

Performance Indicator Verification ............................................................................ 29

4OA2

Problem Identification and Resolution ..................................................................... 29

4OA3

Follow-Up of Events and Notices of Enforcement Discretion.................................... 39

4OA5

Other Activities .......................................................................................................... 43

4OA6

Management Meetings ............................................................................................. 45

4OA7

Licensee-identified Violations ................................................................................... 45

ATTACHMENT: SUPPLEMENTARY INFORMATION ............................................................... 46

SUPPLEMENTARY INFORMATION ........................................................................................ A-1

KEY POINTS OF CONTACT .................................................................................................... A-1

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED ......................................................... A-2

LIST OF DOCUMENTS REVIEWED ........................................................................................ A-3

LIST OF ACRONYMS ............................................................................................................. A-16

3

SUMMARY

Inspection Report (IR) 05000272/2016002, 05000311/2016002; 04/01/2016 - 06/30/2016;

Salem Nuclear Generating Station Units 1 and 2; Operability Determinations and Functionality

Assessments; Follow-Up of Events and Notices of Enforcement Discretion.

This report covered a three-month period of inspection by resident inspectors and announced

inspections performed by regional inspectors. The inspectors documented one self-revealing

finding of very low safety significance (Green), one non-cited violation (NCV), one finding (FIN)

and one licensee identified violation. The significance of most findings is indicated by their color

(i.e., greater than Green, or Green, White, Yellow, Red) and determined using Inspection

Manual Chapter (IMC) 0609, Significance Determination Process (SDP), dated April 29, 2015.

Cross-cutting aspects are determined using IMC 0310, Aspects Within Cross-Cutting Areas,

dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance

with the NRCs Enforcement Policy, dated February 4, 2015. The NRCs program for

overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.

Cornerstone: Mitigating Systems and Initiating Events

Green. The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code

of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, "Instructions, Procedures,

and Drawings," because, from June 15, 2016 until July 26, 2016, PSEG did not accomplish

actions necessary to provide adequate confidence that a structure, system, and component

(SSC) would perform satisfactorily in service (an activity affecting quality) as prescribed by a

documented procedure. Specifically, although PSEG had concluded Salem Unit 2 is

susceptible to baffle bolt failure due to its design and operating life (but less susceptible than

Salem Unit 1), PSEG inadequately implemented Procedure OP-AA-108-115, "Operability

Determinations & Functionality Assessments," Sections 4.7.14 followed by Sections

4.7.18-4.7.20 to perform an operability evaluation (OpEval) to justify continued operation of

the unit until the next refueling outage. PSEGs immediate corrective actions included

entering the issue into its corrective action program (NOTF 20736630) and documenting an

operability evaluation to support the basis for functionality of the baffle structure and the

operability of the emergency core cooling system (ECCS) and reactivity control systems.

This finding is more than minor because it is associated with the equipment performance

attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to

ensure the availability, reliability, and capability of systems that respond to initiating events to

prevent undesirable consequences, in that degradation of a significant number of baffle bolts

could result in baffle plates dislodging following an accident. This issue was dispositioned as

more than minor because it was also similar to example 3.j of IMC 0612, Appendix E,

Examples of Minor Issues, in that the condition resulted in reasonable doubt of operability

of the ECCS and additional analysis was necessary to verify operability. In accordance with

IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A,

The Significance Determination Process for Findings At-Power, issued June 19, 2012, the

inspectors screened the finding for safety significance and determined it to be of very low

safety significance (Green), since the finding did not represent an actual loss of system or

function. After inspector questioning, PSEG performed OpEval 2016-015, which provided

sufficient bases to conclude the Unit 2 baffle assembly would support ECCS and control rod

system operability until the next refueling outage. This finding is related to the cross-cutting

4

aspect of Operating Experience because PSEG did not effectively evaluate relevant internal

and external operating experience. Specifically, PSEG did not adequately evaluate the

impact of degraded baffle bolts in Unit 2 when directly relevant operating experience was

identified at Unit 1. [P.5] (Section 1R15)

Green. A Green, self-revealing finding (FIN) was identified against MA-AA-716-010,

Maintenance Planning Process, Revision 18, when PSEG work orders (WOs) did not

specify the appropriate procedure to perform satisfactory modification testing of the main

generator automatic voltage regulator (AVR) protective relay (model STV1). Consequently,

the relay actuated below its design setpoint on February 4, 2016, resulting in an automatic

trip of the Unit 2 main turbine and reactor. PSEG entered the issue in their Corrective

Action Program (CAP) and performed a root cause evaluation (RCE), replaced the failed

STV1 relay with a properly tested relay, verified other STV relays were appropriately tested

as an extent of condition, and initiated an action to revise Laboratory Testing Services (LTS)

department relay test procedures to ensure all applicable acceptance criteria will be

incorporated.

The inspectors determined that a performance deficiency existed because PSEG WOs did

not specify the appropriate procedure to perform satisfactory modification testing of the main

generator AVR protection relay. This issue was more than minor since it was associated

with the procedure quality attribute of the Initiating Events cornerstone and adversely

impacted its objective to limit the likelihood of events that upset plant stability (turbine and

reactor trip) and challenge critical safety functions. Using IMC 0609, Attachment 4 and

Appendix A, Exhibit 1, the inspectors determined that this finding was of very low safety

significance, or Green, since mitigating equipment relied up to transition the plant to stable

shutdown remained available. The finding had a cross-cutting aspect in the area of Human

Performance, Work Management, in that the PSEG did not adequately implement the work

process to coordinate with engineering and maintenance departments as needed to

appropriately plan the STV1 relay modification test WO. [H.5] (Section 4OA3.3)

Other Findings

A violation of very low safety significance that was identified by PSEG was reviewed by the

inspectors. Corrective actions taken or planned by PSEG have been entered into PSEGs CAP.

This violation and corrective actions tracking number are listed in Section 4OA7 of this report.

5

REPORT DETAILS

Summary of Plant Status

Unit 1 began the inspection period at 100 percent power. The unit was shut down for a

refueling outage on April 14.

Unit 2 began the inspection period at 100 percent power. The unit remained at or near

100 percent power until June 28, when the unit tripped due to actuation of the main generator

protection system. The unit remained shut down at the end of the inspection period.

1.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01 - 1 sample)

.1

Summer Readiness of Offsite and Alternate Alternating Current Power Systems

a. Inspection Scope

The inspectors reviewed plant features and procedures for the operation and continued

availability of the offsite and alternate alternating current (AC) power system to evaluate

readiness of the systems prior to seasonal high grid loading on May 31. The inspectors

reviewed PSEGs procedures affecting these areas and the communications protocols

between the transmission system operator and PSEG. This review focused on changes

to the established program and material condition of the offsite and alternate AC power

equipment. The inspectors assessed whether PSEG established and implemented

appropriate procedures and protocols to monitor and maintain availability and reliability

of both the offsite AC power system and the onsite alternate AC power system. The

inspectors evaluated the material condition of the associated equipment by interviewing

the responsible system manager, reviewing condition reports and open WOs, and

walking down portions of the offsite and AC power systems including the 500 kilovolt

(kV).

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1

Partial System Walkdown (71111.04Q - 4 samples)

a. Inspection Scope

The inspectors performed partial walkdowns of the following systems:

Unit 1, 1A and 1C 125V direct current (DC) system during 1B 125V DC battery

inoperability on April 6

Unit 1, Containment penetrations during irradiated fuel moves on April 19

6

Unit 2, Service water (SW) system during 21 SW pump emergent repairs on June 7

Unit 2, Auxiliary building ventilation with damper 2ABV2 failed open on June 16

The inspectors selected these systems based on their risk-significance relative to the

reactor safety cornerstones at the time they were inspected. The inspectors reviewed

applicable operating procedures, system diagrams, the Updated Final Safety Analysis

Report (UFSAR), technical specification(s) (TSs), WOs, notifications (NOTFs), and the

impact of ongoing work activities on redundant trains of equipment in order to identify

conditions that could have impacted the systems performance of its intended safety

functions. The inspectors also performed field walkdowns of accessible portions of the

systems to verify system components and support equipment were aligned correctly and

were operable. The inspectors examined the material condition of the components and

observed operating parameters of equipment to verify that there were no deficiencies.

The inspectors also reviewed whether PSEG staff had properly identified equipment

issues and entered them into the CAP for resolution with the appropriate significance

characterization.

b. Findings

No findings were identified.

.2

Full System Walkdown (71111.04S - 1 sample)

a. Inspection Scope

On June 22, 2016, the inspectors performed a complete system walkdown of accessible

portions of the Unit 2 safety injection (SI) to verify the existing equipment lineup was

correct. The inspectors reviewed operating procedures, surveillance tests, drawings,

equipment line-up check-off lists, and the UFSAR to verify the system was aligned to

perform its required safety functions. The inspectors also reviewed electrical power

availability, component lubrication and equipment cooling, hanger and support

functionality, and operability of support systems. The inspectors performed field

walkdowns of accessible portions of the systems to verify as-built system configuration

matched plant documentation, and that system components and support equipment

remained operable. The inspectors confirmed that systems and components were

aligned correctly, free from interference from temporary services or isolation boundaries,

environmentally qualified, and protected from external threats. The inspectors also

examined the material condition of the components for degradation and observed

operating parameters of equipment to verify that there were no deficiencies.

Additionally, the inspectors reviewed a sample of related notifications and WOs to

ensure PSEG appropriately evaluated and resolved any deficiencies.

b. Findings

No findings were identified.

7

1R05 Fire Protection

.1

Resident Inspector Quarterly Walkdowns (71111.05Q - 5 samples)

a. Inspection Scope

The inspectors conducted tours of the areas listed below to assess the material

condition and operational status of fire protection features. The inspectors verified that

PSEG controlled combustible materials and ignition sources in accordance with

administrative procedures. The inspectors verified that fire protection and suppression

equipment was available for use as specified in the area pre-fire plan, and passive fire

barriers were maintained in good material condition. The inspectors also verified that

station personnel implemented compensatory measures for out of service, degraded, or

inoperable fire protection equipment, as applicable, in accordance with procedures.

Unit 2, Spent fuel (SF) and component cooling heat exchangers (HXs) on May 12

Unit 2, Boric acid evaporator unit and chemistry area on May 20

Unit 2, SW pump bays during 21 SW pump maintenance on June 8

Unit 2, 2B and 2C emergency diesel generator (EDG) rooms on June 16

Unit 2, Chiller room while protected on June 16

b. Findings

No findings were identified.

1R07 Heat Sink Performance (711111.07A - 1 sample)

a. Inspection Scope

The inspectors reviewed the 12 SI pump lube oil cooler readiness and availability to

perform its safety functions. The inspectors reviewed the design basis for the

component and verified PSEGs commitments to NRC Generic Letter 89-13, Service

Water Requirements Affecting Safety-Related Equipment. The inspectors performed

inspection of the as-found conditions, and discussed the results of previous inspections

with PSEG staff. The inspectors verified that PSEG initiated appropriate corrective

actions for identified deficiencies. The inspectors also verified that the number of tubes

plugged within the HX did not exceed the maximum amount allowed.

b. Findings

No findings were identified.

1R08 In-service Inspection Activities (71111.08 - 1 sample)

a.

Inspection Scope

Inspectors from the NRC Region I Office, specializing in materials and in-service

examination activities, observed portions of PSEGs activities involving baffle bolt

examinations and replacements during the Salem Unit 1 spring 2016 refueling outage

(1R24). PSEG notified the NRC of problems with baffle bolts in Event

8

Notification 51902, Anomalies Identified during Visual Inspection of Reactor Vessel

Internals. During May 17-19, 2016, and June 20-23, 2016, inspectors conducted an

inspection of PSEGs evaluation of the baffle bolt ultrasonic testing results and visual

examination performed during 1R24. The inspectors reviewed documentation,

interviewed personnel, and reviewed video recordings of visual examinations performed

during the current and previous refueling outages. The inspectors also observed in-

progress baffle bolt replacement activities.

Nondestructive Examination and Welding Activities (Section 02.01)

The inspectors conducted a review of PSEGs implementation of in-service inspection

(ISI) program activities for monitoring degradation of the reactor coolant system

boundary, risk significant piping and components, and containment systems during

Salem Unit 1 refueling outage 1R24. The sample selection was based on the inspection

procedure objectives and risk priority of those pressure retaining components in these

systems where degradation would result in a significant increase in risk. The inspectors

observed in-process nondestructive examination (NDE), reviewed records, and

interviewed personnel to verify the following: a) that non-destructive activities were

performed in accordance with American Society of Mechanical Engineers (ASME) Boiler

and Pressure Vessel Code Section XI, 2004 Edition, no Addenda, requirements; b) that

indications and defects, if present, were dispositioned in accordance with the ASME

Code or an NRC approved alternative; and, c) that relevant indications were compared

to previous examinations to determine if any changes occurred.

The inspectors reviewed the ultrasonic testing (UT) procedure used for the examination

of the Unit 1 baffle bolts to verify it met the requirements of the ASME Boiler and

Pressure Vessel Code and the applicable guidance in the Electric Power Research

Institutes Materials Reliability Program (MRP-227 and 228). The inspectors reviewed

the UT data records for the examinations performed during the 1R24 refueling outage to

verify that activities were performed in accordance with applicable examination

procedures.

The inspectors reviewed video from the visual examination of the baffle bolts performed

in the current refueling outage (RFO). The inspectors also reviewed video of visual

examinations performed during Unit 1 RFOs in 2001, 2013, and 2014 to assess the

as-found conditions of the baffle bolts. The inspectors reviewed certifications of the NDE

technicians performing the examinations to verify the examinations were performed by

qualified individuals in accordance with approved procedures and the results reviewed

and evaluated by certified Level III NDE personnel.

The inspectors performed a sample of observations of NDE activities and reviewed

records of NDE activities. The review sample consisted of two or three types of NDE

activities, including at least one volumetric examination.

ASME Code Required Examinations

Salem Unit 1, Liquid Penetrant Report No. PT-16-002, 11-RHRHEX Vessel Support,

4/15/16, (Summary No.205170) [record review]

Salem Unit 1, Liquid Penetrant Report No. PT-16-001, Pipe Lugs 8-RH-2116-10PL-1

through 4, 4/15/16, (Summary No. 263631) [record review]

9

Salem Unit 1, Liquid Penetrant Report No. PT-16-004, Pipe to Penetration IA,

Component 12 SJ-2152-36PS-4, 4/19/16, (Summary No. 263904) [record review]

Salem Unit 1, Liquid Penetrant Report No. PT-16-003, Inlet Nozzle To 11

Charging Pump, Component 6-CV-2111-14R1, 4/15/16,

(Summary No. 220757) [record review]

Salem Unit 1, Liquid Penetrant Report No. PT-16-005, Pipe-to-Valve (11CS48)

[record review] Component ID: 8-CS-2114-60, 4/15/16, (Summary No. 56640)

Salem Unit 1, Ultrasonic examination (Summary #006325) Report UT-16-039,

Component ID: 1-PZR-20, Pressurizer, shell J weld [Observed]

Component ID: 16-BFN-2111-IRS, Inside Radius Section Ultrasonic

Examination, 16-BF-2111, Report UT-16-013, Steam Generator #11,

(Summary #204201) [Observed]

Component 4-PRN-1100-IRS, Pressurizer Relief Nozzle, inside Radius Section,

Ultrasonic Examination, (Summary #007000), UT-16-031, [Observed]

Observation of Baffle Bolt Replacement Activities

The inspectors observed electrical discharge machining activities on a baffle bolt

location. The inspectors observed the bolt hole milling activities for a baffle bolt. The

inspectors verified that bolt replacement activities were being performed in accordance

with approved procedures.

Other Augmented, License Renewal or Industry Initiative Examinations

PSEG did not schedule augmented inspections in the outage scope for 1R24.

Review of Relevant Indication(s) Evaluated and Accepted for Continued Service

PSEG did not have any originally rejectable indications since the end of their prior

outage, which were later accepted for continued use after evaluation.

Modifications, Repairs, or Replacements Consisting of Welding on Pressure Boundary

Risk Significant Systems

The inspectors reviewed Design Change Package 80092579, Salem Unit 1 - Steam

Generator (SG) Bowl Drain Repair, for SGs 11, 12, 13, and 14. This change removed

Alloy 600 and associated 82/182 weld material from each SG channel head bowl drain

plug to reduce the potential for primary water stress corrosion cracking. The inspectors

determined overall whether the modifications were completed in accordance with ASME

Section XI as a repair/replacement activity. Specifically, the inspectors reviewed the

machining and welding procedures used to complete the modifications, reviewed the

training of the machinists, welders and laborers qualified on a mockup of the channel

heads, and reviewed the mockup training completed by all craft personnel on the project.

The inspectors reviewed the in-process NDE and the final NDE procedures to determine

whether the change was implemented in accordance with ASME Section XI

repair/replacement requirements.

10

PWR Vessel Upper Head Penetration Inspection Activities (Section 02.02)

The Salem Unit 1 reactor pressure vessel head was replaced with an Alloy 690 head in

2005. The inspectors determined that reactor pressure vessel head examinations (per

ASME Code Case N-729) were not required during 1R24.

Boric Acid Corrosion Control Inspection Activities (Section 02.03)

The inspectors reviewed the Boric Acid Corrosion Control program and implementing

PSEG procedures, and discussed the outage inspections with program engineers. The

inspectors also reviewed documentation, corrective action process notifications,

including photographic records, of the conditions identified during the plant shutdown.

The inspectors also reviewed a sample of notifications recommending repairs to

identified conditions and a sample of boric acid engineering evaluations performed to

determine the priority of repair of identified boric acid corrosion on safety significant

piping and components. Boric acid inspections were conducted on safety significant

piping and components inside the containment structure during walk downs conducted

by PSEG staff with the plant at normal pressure and temperature conditions. The

inspectors reviewed a sample of photos and visual inspection records to verify that boric

acid leakage was being appropriately identified and non-conforming conditions of boric

acid leaks were documented in the CAP with a focus on areas that could cause

degradation of safety significant components.

The inspectors verified that potentially more significant boric acid deficiencies were

being adequately dispositioned by reviewing a sample of evaluations documented in the

following PSEG condition reports: 20682192, 20699859, 20699820, 20699910,

20704139, 20707125, 20712774, 20713572, 20722494, 20682192, 20699859,

20707125, 20722494, 70179375, 20699820, 20704139, 70185980, 20712774,

20713573, 20713572.

These reviews verified whether the corrective actions were consistent with the

requirements of the ASME Code and 10 CFR Part 50, Appendix B, Criterion XVI. The

inspectors reviewed the engineering evaluations associated with these condition reports

to verify whether equipment or components wetted or impinged upon by boric acid

solutions were properly analyzed for degradation that might impact their function.

Steam Generator Tube Inspection Activities (Section 02.04)

PSEGs Base Eddy Current Test (ECT) program consisted of: (a) 100 percent bobbin

probe inspection of straight and U-bend tubes, (b) 50 percent Hot Leg coverage of Top

of Tubesheet area with an array probe, (c) 3 tube periphery tube array testing, and

various + Point sampling strategies (for U-bend and Dent/Ding inspections) of in-service

tubes were completed in each SG. The inspectors reviewed the 1R24 SG tube

Degradation Assessment, ECT examination scope and expansion criteria to verify that it

met TS requirements, Electric Power Research Institute (EPRI) guidelines, and

commitments made to the NRC. The inspectors also verified that the ECT scope

included areas of degradation that were known to represent potential ECT challenges

such as the top of tube sheet, tube support plates, and U-bends. Upon completion of

eddy current (EC) examinations and the evaluation of all data, PSEG staff determined

that six tubes required plugging. The affected tubes were plugged during 1R24. The

11

inspectors verified that the affected tubes were properly screened against the in situ

screening criteria and that none of the tube indications required in-situ pressure testing.

The inspectors observed portions of the ECT being performed and verified whether:

(1) the appropriate probes were used for identifying the expected types of degradation,

(2) calibration requirements were adhered to, and (3) probe travel speed was in

accordance with procedural requirements. The inspectors performed a review of the

site-specific qualifications for the techniques being used, and verified whether the ECT

data analyses were adequately performed per EPRI and PSEG specific guidelines. The

inspectors selected a sample of degraded tubes and compared them to the previous

outage operational assessment to assess PSEGs prediction capabilities. The inspectors

also reviewed a sample of EC data, and verified, through discussion with the data

analyst that the analytical techniques used to evaluate the inspection data were

adequate. The inspectors further verified that the assumed NDE flaw sizing accuracy

was consistent with data from EPRI examination technique specification sheet or

applicable performance demonstration. Finally, the inspectors reviewed the

qualifications for the EC data collection personnel, a sample of the inspection

supervision personnel qualifications and a sample of the qualifications of staff

responsible for interpretation and resolution analysis to determine whether the records

were complete.

The inspectors observed a portion of a plug integrity visual examination per procedure

81DP-9RC40, Steam Generator Channel Head Video Inspection, to verify that those

tubes that had been previously plugged did not exhibit any leakage. No evidence of plug

leakage was identified. Additionally, the inspectors observed a portion of the secondary

sludge lancing and foreign object search and retrieval (FOSAR) inspections. No

significant foreign materials or quantity of sludge were identified.

During the prior operating cycle previous to the current refueling outage 1R24, the

inspectors determined whether leakage from each SG was measured, via sampling of

each SG, for the complete prior operating cycle (leakage was not measured).

PSEG staff completed secondary side inspections and sludge lancing of all SGs. The

inspectors reviewed the results to determine that no loose parts affecting tube integrity

were noted and that other SG related inspections were performed without repairs.

PSEG staff performed a plug integrity visual examination to verify that those tubes that

had been previously plugged did not exhibit leakage. From this visual exam, PSEG staff

documented excessive boron buildup around tube plug 43-34 in the SG 11 cold leg and

initiated CR-2016-29172 to track the evaluation of the condition. PSEG staff also

initiated Notification 20726743 to track the condition. PSEG Engineering staff review of

the plug concluded that no evidence of plug leakage had occurred. Additionally,

secondary sludge lancing and FOSAR inspections were performed in each SG. No

foreign materials, which could damage SG tubes, were identified. The inspectors

reviewed the PSEG evaluations and information to determine the conclusions were

technically supported.

Identification and Resolution of Problems (Section 02.05)

The inspectors reviewed a sample of condition reports, which identified NDE indications,

deficiencies and other nonconforming conditions since the previous, 1R23, refueling

outage. The inspectors verified that nonconforming conditions were properly identified,

12

characterized, evaluated, corrective actions identified and dispositioned, and

appropriately entered into the CAP.

b. Findings

Introduction. The inspectors determined the level of degradation of Unit 1 baffle bolts

reported to the NRC as a condition not previously analyzed is an issue of concern that

warrants additional inspection to determine whether a performance deficiency exists. As

a result, the NRC opened a unresolved item (URI).

Description. Additional inspection is warranted to determine whether a performance

deficiency exists related to Event Notification 51902, dated May 3, 2016, in which PSEG

reported to the NRC that the level of degradation of baffle bolts was a condition not

previously analyzed. The baffle bolts secure plates in the reactor core barrel to form a

shroud around the fuel core to direct reactor coolant flow upward through the fuel

assemblies. In order to determine if a performance deficiency exists, the inspectors will

review the results of PSEGs RCE which will be completed at a later date.

(URI 05000272/2016002-01, Baffle-Former Bolts with Identified Anomalies)

1R11 Licensed Operator Requalification Program (71111.11Q - 1 sample)

Quarterly Review of Licensed Operator Requalification Testing and Training

a. Inspection Scope

The inspectors observed licensed operator simulator training on June 8, 2016, which

included a heater drain pump oil leak, a steam generator feed pump trip, and a steam

generator tube rupture. The inspectors evaluated operator performance during the

simulated event and verified completion of risk significant operator actions, including the

use of abnormal and emergency operating procedures. The inspectors assessed the

clarity and effectiveness of communications, implementation of actions in response to

alarms and degrading plant conditions, and the oversight and direction provided by the

control room supervisor. The inspectors verified the accuracy and timeliness of the

emergency classification made by the shift manager and the TS action statements

entered by the shift technical advisor. Additionally, the inspectors assessed the ability of

the crew and training staff to identify and document crew performance problems.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness (71111.12Q - 3 samples)

a. Inspection Scope

The inspectors reviewed the samples listed below to assess the effectiveness of

maintenance activities on SSC performance and reliability. The inspectors reviewed

system health reports, CAP documents, maintenance WOs, and maintenance rule (MR)

basis documents to ensure that PSEG was identifying and properly evaluating

performance problems within the scope of the MR. For each sample selected, the

inspectors verified that the SSC was properly scoped into the MR in accordance with

13

10 CFR 50.65 and verified that the (a)(2) performance criteria established by PSEG staff

was reasonable. As applicable, for SSCs classified as (a)(1), the inspectors assessed

the adequacy of goals and corrective actions to return these SSCs to (a)(2).

Additionally, the inspectors ensured that PSEG staff was identifying and addressing

common cause failures that occurred within and across MR system boundaries.

Unit 2, 22SW535, unsatisfactory stroke time of SW accumulator supply valve to

22 containment fan cooler unit (CFCU) on May 2

Unit 2, Circulating water system 125V DC battery degradation on May 23

Common, MR URI, 05000272;311/2015008-01: Inadequate MR System

Performance Criteria Selection, closeout on May 1

b. Findings

No findings were identified. Additional inspection results regarding the URI closeout are

documented in Section 4OA5.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 5 samples)

a. Inspection Scope

The inspectors reviewed station evaluation and management of plant risk for the

maintenance and emergent work activities listed below to verify that PSEG performed

the appropriate risk assessments prior to removing equipment for work. The inspectors

selected these activities based on potential risk significance relative to the reactor safety

cornerstones. As applicable for each activity, the inspectors verified that PSEG

personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the

assessments were accurate and complete. When PSEG performed emergent work, the

inspectors verified that operations personnel promptly assessed and managed plant risk.

The inspectors reviewed the scope of maintenance work and discussed the results of

the assessment with the stations probabilistic risk analyst to verify plant conditions were

consistent with the risk assessment. The inspectors also reviewed the TS requirements

and inspected portions of redundant safety systems, when applicable, to verify risk

analysis assumptions were valid and applicable requirements were met.

Unit 1, 11SW223, SW outlet valve to 11 CFCU, failure to close on April 7

Unit 1, Reactor core baffle-to-former bolt expanded inspection scope on April 22

Unit 2, Appendix R safe shutdown panel failed indication on May 9

Unit 2, 2A subcooling margin monitor failure on May 26

Unit 2, Yellow risk with one offsite power source unavailable on June 1

b. Findings

No findings were identified.

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1R15 Operability Determinations and Functionality Assessments (71111.15 - 9 samples)

a. Inspection Scope

The inspectors reviewed operability determinations for the following degraded or

non-conforming conditions based on the risk significance of the associated components

and systems:

Unit 1, Corrosion and metal loss identified during inspection of 11 SW header

on April 23

Unit 1, Immediate operability determination (IOD) of the degraded condition of the

baffle-former bolts on April 27

Unit 1, 1 Emergency control air compressor shutdown on April 27

Unit 1, SI thermal relief valve failures on May 2

Unit 1, 13 turbine-driven auxiliary feedwater (AFW) pump degraded performance

on May 8

Unit 1, 11 diesel fuel oil storage tank high particulates on May 18

Unit 2, IOD of the degraded condition of the baffle-former bolts identified from Unit 1

operating experience on April 27

Unit 2, 125V DC battery degraded cell post connections on May 2

Common, 10 CFR Part 21 issue related to safety-related 4kV breakers on May 16

The inspectors evaluated the technical adequacy of the operability determinations to

assess whether TS operability was properly justified and the subject component or

system remained available such that no unrecognized increase in risk occurred. The

inspectors compared the operability and design criteria in the appropriate sections of the

TSs and UFSAR to PSEGs evaluations to determine whether the components or

systems were operable. The inspectors confirmed, where appropriate, compliance with

bounding limitations associated with the evaluations. Where compensatory measures

were required to maintain operability, the inspectors determined whether the measures

in place would function as intended and were properly controlled by PSEG.

b. Findings

Introduction. The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B,

Criterion V, "Instructions, Procedures, and Drawings," because, from June 15, 2016

until July 26, 2016, PSEG did not accomplish actions necessary to provide adequate

confidence that an SSC would perform satisfactorily in service (an activity affecting

quality) as prescribed by a documented procedure. Specifically, although PSEG had

concluded Salem Unit 2 is susceptible to baffle bolt failure due to its design and

operating life (but less susceptible than Salem Unit 1), PSEG inadequately implemented

Procedure OP-AA-108-115, "Operability Determinations & Functionality Assessments,"

by not performing Section 4.7.14 followed by Sections 4.7.18-4.7.20 to perform an

operability evaluation (OpEval) to justify continued operation of the unit until the next

refueling outage. In particular, PSEG incorrectly exited their procedure on June 15,

2016, and re-entered it to complete these steps on July 26, 2016, based on discussions

with the NRC. The operability evaluation provided appropriate justification for the

licensees plans to examine the baffle-former bolts at the next Unit 2 RFO.

15

Description. On April 22, 2016, PSEG identified baffle-former (baffle) bolt degradation

at Salem Unit 1 that was determined to be unanalyzed because it did not meet the

minimum acceptable bolt pattern analysis developed to support plant startup. PSEG

staff identified that 192 baffle bolts out of a total population of 832 were considered

degraded. On May 4, 2016, due to the number of degraded baffle bolts discovered on

Unit 1, PSEG staff determined that it was necessary to perform an extent of condition

review for the baffle bolts on Unit 2. PSEG entered this issue into the corrective action

program as NOTF 20727590 and completed an immediate operability determination

(IOD) to evaluate the Unit 2 baffle bolts and baffle assembly structure in accordance with

PSEG procedure OP-AA-108-115, "Operability Determinations & Functionality

Assessments," Section 4.7.4.

The inspectors reviewed the design basis and current licensing basis documents for

Unit 2 to identify the specific safety functions of the baffle bolts. The inspectors identified

that the baffle bolts are part of the baffle assembly structure located in the reactor

pressure vessel. The bolts secure a series of vertical metal plates called baffle plates,

which help direct water up through the nuclear fuel assemblies to ensure proper cooling

of the fuel. A sufficient number of baffle bolts are required to secure the plates to ensure

proper core flow during normal and postulated accident conditions, and also to ensure

that control rods can be inserted to shut down the reactor.

On June 21, 2016, the inspectors reviewed the IOD as part of a detailed review of the

ongoing baffle bolt activities at Salem and noted that the IOD concluded that there was

reasonable assurance that the Unit 2 reactor assembly was operable, but required

additional evaluation due to the conditions observed in Unit 1. Specifically, the IOD

concluded that there was reasonable assurance that the Unit 2 reactor assembly was

operable pending further evaluation based upon the following factors: (1) Unit 2 had

fewer effective full power years of operation than Unit 1; (2) a baffle bolt visual

examination completed during the most recent Unit 2 2R21 refueling outage (fall 2015)

did not identify any visual deficiencies; and, (3) there was no current indication of reactor

fuel pin leakage in Unit 2, which could be caused by baffle bolt failure and subsequent

fretting. The inspectors review of PSEGs IOD concluded that the IOD provided

sufficient technical detail to support the initial conclusion that there was reasonable

assurance, based on the limited information available, that the Unit 2 baffle bolts would

retain sufficient capability to perform their intended functions. PSEG procedure OP-AA-

108-115, Section 4.7.11 directs that if there is a reasonable expectation that the SSC is

operable, but a more rigorous evaluation is deemed warranted, then update the current

notification or initiate a notification for Engineering to prepare a Technical Evaluation to

support the prompt determination of operability. The immediate actions section of

NOTF 20727590 requested a work order be generated to perform an extent of condition

review for Unit 2 baffle bolts. The Station Ownership Committee (SOC) screening of

NOTF 20727590 on May 6, 2016, assigned a work order to Engineering to ensure that

Operations is provided the Technical Evaluation product. This will allow review for

assessment of operability as required. From review of the daily running log of baffle

bolt action items spreadsheet, the inspectors noted that on May 4, 2016, action EOC.2

to perform an operability evaluation for Unit 2 was closed to EOC.7-9, to complete an

adverse condition monitoring plan, an operational decision making document, and a

Technical Evaluation in lieu of an OpEval. Consistent with this decision, on May 26,

2016, the Salem plant manager discussed with the senior resident inspector PSEGs

views that an operability evaluation was not required or being developed. In response,

16

the inspectors shared their understanding of PSEG procedure guidance and regulatory

requirements in this regard.

Between May 6 and June 15, 2016, PSEG engineering performed Technical Evaluation

70187161, Extent of Condition Review for Salem Unit 2 Susceptibility to Baffle Bolt

Failure. The purpose of the Technical Evaluation was to determine the potential for

baffle bolt degradation in Unit 2 based upon the results of visual and ultrasonic

examination results observed in Unit 1, and to identify and evaluate key factors that

could potentially impact the safe operation of Unit 2 for the remainder of the current

operating cycle. The Technical Evaluation evaluated the key factors that affect

irradiation assisted stress corrosion cracking (IASCC). Additionally, the Technical

Evaluation assessed the safety consequences of the degraded baffle bolts in the as-

found condition in Unit 1. The Technical Evaluation conclusion summary indicated that

Unit 2 is susceptible to baffle bolt failure due to its design and operating life; that any

degradation in Unit 2 would be less advanced that that observed in Unit 1; and that

PSEG should exercise heightened awareness and monitoring of Unit 2 due to this

vulnerability. The Technical Evaluation also concluded that Unit 1 could have safely

shut down and the core would be cooled by demonstrating that control rod insertability is

assured and a core coolable geometry was maintained. Thus the Technical Evaluation

concluded that Unit 2 could also be shut down and cooled based upon the conclusion

reached regarding Unit 1. Following completion of the Technical Evaluation on June 15,

PSEG did not continue on in the operability determination process.

The inspectors assessed PSEGs Technical Evaluation 70187161 during an onsite

inspection which took place from June 21-23, 2016. PSEG concluded in Technical

Evaluation 70187161, that Salem Unit 2 is susceptible to baffle bolt failure due to its

design and operating history, but less so than observed in Salem Unit 1. The inspectors

determined this conclusion met PSEGs definition of a degraded condition as defined in

procedure OP-AA-115-108, Section 2.4. Section 2.4 defines a degraded condition as A

condition in which the qualification of an SSC or its functional capability is reduced.

Section 2.4 lists reduced reliability as an example of a degraded condition and aging

as an example of a condition that can reduce the capability of a system. The inspectors

noted that IASCC is a time dependent aging degradation mechanism and baffle bolt

failures reduce the functional capability and reliability of the baffle assembly.

Consequently the Technical Evaluation describes a degraded condition in the Unit 2

baffle assembly. Since the Technical Evaluation concluded that the reactor could be

shut down and cooled based upon the assessment of safety consequences, the

inspectors concluded that PSEG considered that the reactivity control and emergency

core cooling systems were operable. As a result, the inspectors concluded that PSEG

should have continued on in the operability determination process as described in

Section 4.7.14, Operable but Degraded or Nonconforming, and declared both the

reactivity control and emergency core cooling systems operable but degraded. Once a

SSC is determined to be operable but degraded, Section 4.7.18 directs that An

OpEval will be requested based on a declaration of operable but degraded or

nonconforming. Section 4.7.19 directs Engineering to Prepare and review and

OpEval. Section 4.7.20 directs Operations to approve or disapprove the OpEval when

Engineering completes it. Sections 4.7.14, 4.7.18, 4.7.19 and 4.7.20 were not

implemented by PSEG.

The inspectors acknowledged that licensees apply judgment in these decisions and can

use a graded approach regarding the level of detail. In this particular instance, the

17

inspectors considered that operating experience was available that showed the Unit 2

baffle bolts were subject to IASCC and that plants of similar design (4-loop

Westinghouse pressurized water reactors with a down-flow configuration and baffle bolts

of 347 stainless steel material and similar dimensions) were subject to greater amounts

of bolt degradation compared to other reactor designs. Furthermore, the inspectors

noted the baffle bolts had experienced levels of neutron radiation exposure above the

threshold for IASCC initiation as referenced in NUREG/CR-7027, Degradation of LWR

Core Internal Materials due to Neutron Irradiation.

The inspectors conducted an exit meeting on June 23, 2016, describing a potential

violation of 10 CFR Part 50 Appendix B, Criterion 5, Instructions, Procedures, and

Drawings, for PSEG not completing the OpEval and assessing the effect of the

operability of the ECCS and rod control system based upon the functionality of the baffle

former assembly. Consistent with the change made by PSEG staff to the Salem action

item list on May 4, 2016, to not perform an OpEval, the PSEG Compliance Director

indicated that an operability evaluation was not required and therefore they disagreed

with this finding.

The inspectors determined that Engineering did not perform an OpEval as directed by

OP-AA-108-115 Section 4.7.19, which states PREPARE and REVIEW an OpEval. The

OpEval Form (Attachment 1), or a facsimile, may be used to document the engineering

evaluation (Engineering). Because an OpEval was not prepared, Operations did not

have the opportunity to approve or disapprove an OpEval as required by

OP-AA-108.115, Section 4.7.20 which states: When Engineering completes the

OpEval, then APPROVE or DISAPPROVE.

In summary, Technical Evaluation 70187161 concluded Unit 2 is susceptible to IASCC

baffle bolt degradation and that the expected degradation should be less than that

observed in Unit 1. The inspectors assessed that PSEGs conclusions concerning the

susceptibility and expected degradation in Unit 2 was adequately supported. However,

the inspectors concluded that the Technical Evaluation did not provide adequate

confidence that SSCs (baffle bolts supporting ECCS) would perform satisfactorily in

service to justify continued operation of Unit 2 until the next refueling outage in the

spring of 2017 in that line break size assumptions were not adequately supported.

Following discussions with NRC Region I management and the inspectors, PSEG staff

subsequently completed an operability evaluation (OpEval 2016-015) on July 26, 2016.

The OpEval compared the differences in the operating history and parameters between

Unit 1 and Unit 2 and again concluded that Unit 2 was less susceptible than Unit 1

primarily due to significantly fewer thermal cycles and fewer effective full power years

(EFPY) of operation. The OpEval concluded that operability was supported although

the Unit 2 baffle assemblies are considered degraded since Unit 2 is susceptible to

degraded baffle bolts. Based upon a qualitative analysis, PSEGs OpEval stated that

Unit 2 can accommodate 38 percent degraded baffled-former bolts (distributed across

the assembly) and remain within the acceptable bolting pattern analysis patterns

assuming the dynamic loads of a large break loss of coolant accident. The inspectors

concluded that PSEGs OpEval 2016-015 provided an adequate basis to conclude that

the Unit 2 baffle assembly would support ECCS and rod control system continued

operation until the planned refueling outage in spring 2017. In particular, the inspectors

considered that PSEGs visual examinations of approximately 70 percent of the baffle

bolts, in the fall 2015 refueling outage (2R21), did not identify any bolts that were

18

missing or visually degraded. Considering the collective results from Salem Unit 1 and 2

baffle bolt visual examination results, the inspectors determined this evidence, in

conjunction with a review of other operating factors (EFPY and thermal cycles), provided

a reasonable expectation of the Salem Unit 2 baffle assemblys capability to perform its

supporting TS functions.

Analysis. The inspectors determined that a performance deficiency resulted when PSEG

did not implement Procedure OP-AA-108-115, "Operability Determinations &

Functionality Assessments," Section 4.7.14 followed by Sections 4.7.18-4.7.20 to

perform an OpEval to justify continued operation of the unit until the next refueling

outage for the Unit 2 baffle bolt degraded condition until questioned by NRC inspectors.

PSEGs initial documentation did not provide sufficient basis for continued operation until

the next refueling outage. Specifically, based upon the Technical Evaluation 70187161

conclusion that the Salem Unit 2 design and operating life make it susceptible to baffle

bolt failures, the inspectors determined that PSEG, in effect, concluded that a degraded

condition exists in Unit 2. Therefore, PSEG should have continued on in the operability

determination process as described in Section 4.7.14, Operable but Degraded or

Nonconforming.

This finding is more than minor because it is associated with the equipment performance

attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to

ensure the availability, reliability, and capability of systems that respond to initiating

events to prevent undesirable consequences, in that, degradation of a significant

number of baffle bolts could result in baffle plates dislodging following an accident. This

issue was dispositioned as more than minor because it was also similar to example 3.j of

IMC 0612, Appendix E, Examples of Minor Issues, in that, the condition resulted in

reasonable doubt of operability of the ECCS and additional analysis was necessary to

verify operability. In accordance with IMC 0609.04, Initial Characterization of Findings,

and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for

Findings At-Power, issued June 19, 2012, the inspectors screened the finding for safety

significance and determined it to be of very low safety significance (Green), since the

finding did not represent an actual loss of system or function. After inspector

questioning, PSEG performed OpEval 2016-015, which provided sufficient bases to

conclude the Unit 2 baffle assembly would support ECCS and control rod system

operability until the next RFO. This finding is related to the cross-cutting aspect of

Operating Experience because PSEG did not effectively evaluate relevant internal and

external operating experience. Specifically, PSEG did not adequately evaluate the

impact of degraded baffle bolts at Unit 2 when directly relevant operating experience

was identified at Unit 1. [P.5]

Enforcement. 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and

Drawings, states, in part, that activities affecting quality shall be prescribed by

documented procedures of a type appropriate to the circumstances and shall be

accomplished in accordance with those procedures. The Introduction to Appendix B

states that quality assurance comprises all those planned and systematic actions

necessary to provide adequate confidence that a SSC will perform satisfactorily in

service. PSEG Procedure OP-AA-108-115, "Operability Determinations & Functionality

Assessments," prescribes PSEGs process to assess the operability of SSCs that are

required to be operable by TSs, or that perform required support functions for SSCs that

are required to be operable by TSs. Section 4.7 prescribes the operability determination

process. Section 4.7.14 states that if an SSC described in TSs is determined to be

19

operable even though a degraded or nonconforming condition is present, then the SSC

is considered operable but degraded or nonconforming. Sections 4.7.18 - 4.7.20

describe how the Operations Shift Manager should request the site engineering staff to

perform an OpEval upon a declaration of operable but degraded, or nonconforming.

The OpEval is completed to justify continued operation during the period of time while

operable but degraded or nonconforming conditions exist.

Contrary to the above, from June 15, 2016, until July 26, 2016, PSEG did not

accomplish actions necessary to provide adequate confidence that an SSC would

perform satisfactorily in service (an activity affecting quality) as prescribed by a

documented procedure. Specifically, although PSEG had concluded the Salem Unit 2

design and operating life make it susceptible to baffle former bolt failures, PSEG

inadequately implemented Procedure OP-AA-108-115, to perform an OpEval to justify

continued operation of the unit. PSEGs corrective actions included entering the issue

into its corrective action program (NOTF 20736630) and documenting an adequate

operability evaluation (OpEval 2016-015 on July 26, 2016) to support the basis for

functionality of the baffle structure and its ability to support the operability of the ECCS

and reactivity control systems. This violation is being treated as an NCV, consistent with

Section 2.3.2 of the Enforcement Policy. (NCV 05000311/2016002-02, Failure to

Follow Operability Determination Procedure for Unit 2 Baffle-Former Bolts)

1R18 Plant Modifications (71111.18 - 2 samples)

.2

Permanent Modifications

a. Inspection Scope

The inspectors reviewed Design Change Package (DCP) 80117136, Salem Unit 1

Baffle to Former Bolt Replacement. This modification documents the replacement of

189 degraded and potentially degraded baffle bolts with a new design baffle bolt made of

an improved material. Additionally the modification documented the locations of the

replacement bolts and the location of three degraded or potentially degraded bolts which

were left in place and are described below. The inspectors also reviewed modification

documents (DCP 80117378) associated with the equivalency evaluation of the material

change from Type 347 stainless steel to Type 316 stainless steel, and the bolt head

design change from a slot to a hex configuration. Thus this inspection involved two

samples - 1) the bolting pattern analysis for the replacement bolts, and 2) a review of

the bolting material change.

This modification was completed during the spring 2016 refueling outage (1R24) and

involved the replacement of 189 baffle bolts out of a total of 832 located in the Unit 1

reactor vessel. PSEG replaced 189 either degraded or potentially degraded baffle bolts

as observed by visual indications of missing or protruding bolt heads, missing or broken

lock bar, bolts that did not pass ultrasonic testing or bolts that were inaccessible for

ultrasonic testing. PSEG did not remove and replace three bolts that were potentially

degraded due to difficulties encountered during the removal/replacement process. One

bolt had an indication during ultrasonic testing but was not visibly damaged. The second

bolt was inaccessible for ultrasonic testing, which would have required replacement.

The third bolt had successfully passed an ultrasonic test but had a visual indication on

one of the lock bar welds which may have indicated a crack in the weld.

20

The inspectors reviewed PSEGs analysis and the Westinghouse minimum bolting

analysis and determined that leaving the one degraded and two potentially degraded

bolts installed was technically acceptable and that the baffle assembly was functional as

a system support component. Details of the NRC assessment of the final configuration

of the baffle bolts and the minimum bolting analysis can be found in Section 4OA2 of this

report.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing (71111.19 - 9 samples)

a. Inspection Scope

The inspectors reviewed the post-maintenance tests for the maintenance activities listed

below to verify that procedures and test activities adequately tested the safety functions

that may have been affected by the maintenance activity, that the acceptance criteria in

the procedure were consistent with the information in the applicable licensing basis

and/or design basis documents, and that the test results were properly reviewed and

accepted and problems were appropriately documented. The inspectors also walked

down the affected job site, observed the pre-job brief and post-job critique where

possible, confirmed work site cleanliness was maintained, and witnessed the test or

reviewed test data to verify quality control hold point were performed and checked,

and that results adequately demonstrated restoration of the affected safety functions.

Unit 1, 13 Station power transformer tap changer did not function in automatic on

May 4

Unit 1 11SJ45, residual heat removal (RHR) to SI motor-operated valve failure to

stroke closed on May 5

Unit 1, 12 containment fan cooling unit (CFCU) motor cooler HX failed leak test on

May 6

Unit 1, Reactor coolant pump flow channel III degraded on May 6

Unit 1, Turbine-driven AFW room cooler cycling on May 10

Unit 1, Reactor vessel level indication system capillary repair on May 13

Unit 2, 24 SW strainer trip on thermal overloads on April 7

Unit 2, 24 SG flow channel 1 drop to 93 percent on May 4

Unit 2, 21 Chiller thermal expansion valve failure on May 24

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities (71111.20 - 1 sample)

a. Inspection Scope

The inspectors reviewed the stations work schedule and outage risk plan for the Unit 1

maintenance and refueling outage (1R24), conducted April 14 through the end of the

quarter. The inspectors reviewed PSEGs development and implementation of outage

21

plans and schedules to verify that risk, industry experience, previous site-specific

problems, and defense-in-depth were considered. During the outage, the inspectors

observed portions of the shutdown and cooldown processes and monitored controls

associated with the following outage activities:

Configuration management, including maintenance of defense-in-depth,

commensurate with the outage plan for the key safety functions and compliance with

the applicable TSs when taking equipment out of service

Implementation of clearance activities and confirmation that tags were properly hung

and that equipment was appropriately configured to safely support the associated

work or testing

Installation and configuration of reactor coolant pressure, level, and temperature

instruments to provide accurate indication and instrument error accounting

Status and configuration of electrical systems and switchyard activities to ensure that

TSs were met

Monitoring of decay heat removal operations

Impact of outage work on the ability of the operators to operate the SF pool cooling

system

Reactor water inventory controls, including flow paths, configurations, alternative

means for inventory additions, and controls to prevent inventory loss

Activities that could affect reactivity

Maintenance of secondary containment as required by TSs

Refueling activities, including fuel handling and fuel receipt inspections

Fatigue management

Tracking of startup prerequisites, walkdown of the drywell (primary containment) to

verify that debris had not been left which could block the emergency core cooling

system suction strainers, and startup and ascension to full power operation

Identification and resolution of problems related to refueling outage activities

Foreign Object Search and Retrieval (FOSAR) for missing baffle bolts and locking

tabs

During this outage, PSEG replaced 189 degraded baffle bolts in the Unit 1 reactor vessel

baffle assembly. This emergent project resulted in the extension of the outage schedule

from 36 days to 106 days.

b. Findings

No findings were identified.

1R22 Surveillance Testing (71111.22 - 5 samples)

a. Inspection Scope

The inspectors observed performance of surveillance tests and/or reviewed test data of

selected risk-significant SSCs to assess whether test results satisfied TSs, the UFSAR,

and PSEG procedure requirements. The inspectors verified that test acceptance criteria

were clear, tests demonstrated operational readiness and were consistent with design

documentation, test instrumentation had current calibrations and the range and accuracy

for the application, tests were performed as written, and applicable test prerequisites

were satisfied. Upon test completion, the inspectors considered whether the test results

22

supported that equipment was capable of performing the required safety functions. The

inspectors reviewed the following surveillance tests:

Unit 1, Manual SI on April 17

Unit 1, 11CA360, control air header supply check valve, as-found local leak rate test

(LLRT) on April 22

Unit 2, 21 RHR In-service Testing on April 1

Unit 2, 22SW223, SW outlet valve to 22 CFCU, stroke time in the required evaluation

range on May 3

Unit 2, Reactor coolant system (RCS) elevated leakrate on May 17

b. Findings

No findings were identified.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation (71114.06 - 1 sample)

Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of a routine PSEG emergency drill on June 16 to

identify any weaknesses and deficiencies in the classification, notification, and protective

action recommendation development activities. The inspectors observed emergency

response operations in the simulator, technical support center, and emergency

operations facility to determine whether the event classification, notifications, and

protective action recommendations were performed in accordance with procedures. The

inspectors also attended the drill critique to compare inspector observations with those

identified by PSEG staff in order to evaluate PSEGs critique and to verify whether the

PSEG staff was properly identifying weaknesses and entering them into the CAP.

b. Findings

No findings were identified.

2.

RADIATION SAFETY

Cornerstones: Occupational and Public Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01 - 6 samples)

a. Inspection Scope

The inspectors reviewed PSEGs performance in assessing and controlling radiological

hazards in the workplace. The inspectors used the requirements contained in 10 CFR

Part 20, TSs, applicable Regulatory Guides (RGs), and the procedures required by TSs

as criteria for determining compliance.

23

Inspection Planning

The inspectors reviewed the PIs for the occupational radiation safety cornerstone,

radiation protection (RP) program audits, and reports of operational occurrences in

occupational radiation safety since the last inspection.

Radiological Hazard Assessment (1 sample)

The inspectors conducted independent radiation measurements during walk-downs of

the facility and reviewed the radiological survey program, air sampling and analysis,

continuous air monitor use, recent plant radiation surveys for radiological work activities,

and any changes to plant operations since the last inspection to verify survey adequacy

of any new radiological hazards for onsite workers or members of the public.

Instructions to Workers (1 sample)

The inspectors reviewed high radiation area work permit controls and use; observed

containers of radioactive materials and assessed whether the containers were labeled

and controlled in accordance with requirements.

The inspectors reviewed several occurrences where a workers electronic personal

dosimeter alarmed. The inspectors reviewed PSEGs evaluation of the incidents,

documentation in the CAP, and whether compensatory dose evaluations were

conducted when appropriate. The inspectors verified follow-up investigations of actual

radiological conditions for unexpected radiological hazards were performed.

Contamination and Radioactive Material Control

The inspectors observed the monitoring of potentially contaminated material leaving the

radiological controlled area and inspected the methods and radiation monitoring

instrumentation used for control, survey, and release of that material.

Radiological Hazards Control and Work Coverage (1 sample)

The inspectors evaluated in-plant radiological conditions and performed independent

radiation measurements during facility walk-downs and observation of radiological work

activities. The inspectors assessed whether posted surveys; radiation work permits

(RWPs); worker radiological briefings and RP job coverage; the use of continuous air

monitoring, air sampling, and engineering controls; and dosimetry monitoring were

consistent with the present conditions. The inspectors examined the control of highly

activated or contaminated materials stored within the SF pools and the posting and

physical controls for selected high radiation areas (HRAs), locked high radiation areas

(LHRAs) and very high radiation areas (VHRAs) to verify conformance with the

occupational PI.

Risk-Significant High Radiation Area and Very High Radiation Area Controls (1 sample)

The inspectors reviewed the procedures and controls for HRAs, VHRAs, and radiological

transient areas in the plant.

24

Radiation Worker Performance and Radiation Protection Technician Proficiency

(1 sample)

The inspectors evaluated radiation worker performance with respect to RP work

requirements. The inspectors evaluated RP technicians in performance of radiation

surveys and in providing radiological job coverage.

Problem Identification and Resolution (1 sample)

The inspectors evaluated whether problems associated with radiation monitoring and

exposure control (including operating experience) were identified at an appropriate

threshold and properly addressed in the CAP.

b. Findings

No findings were identified.

2RS2 Occupational As Low As is Reasonable Achievable Planning and Controls

(71124.02 - 3 samples)

a. Inspection Scope

The inspectors assessed PSEGs performance with respect to maintaining occupational

individual and collective radiation exposures as low as is reasonably achievable

(ALARA). The inspectors used the requirements contained in 10 CFR Part 20,

applicable RGs, TSs, and procedures required by TSs as criteria for determining

compliance.

Inspection Planning

The inspectors conducted a review of Salem Station collective dose history and trends;

ongoing and planned radiological work activities; previous post-outage ALARA reviews;

radiological source term history and trends; and ALARA dose estimating and tracking

procedures.

Radiological Work Planning

The inspectors selected the following radiological work activities based on exposure

significance for review:

RWP 13, Control Rod Drive Activities

RWP 14 , Pressurizer Activities

RWP 17, Primary SG Work

For each of these activities, the inspectors reviewed: ALARA work activity evaluations;

exposure estimates; and exposure reduction requirements.

25

Verification of Dose Estimates and Exposure Tracking Systems

The inspectors reviewed the current annual collective dose estimate; basis methodology;

and measures to track, trend, and reduce occupational doses for ongoing work activities.

The inspectors evaluated the adjustment of exposure estimates or re-planning of work.

Source Term Reduction and Control (1 sample)

The inspectors reviewed the current plant radiological source term and historical trend,

plans for plant source term reduction, and contingency plans for changes in the source

term as the result of changes in plant fuel performance or changes in plant primary

chemistry.

The inspectors observed radiological work activities and evaluated the use of shielding

and other engineering work controls based on the radiological controls and ALARA plans

for those activities.

Radiation Worker Performance (1 sample)

The inspectors observed radiation worker and RP technician performance during

radiological work to evaluate worker ALARA performance according to specified work

controls and procedures. Workers were interviewed to assess their knowledge and

awareness of planned and/or implemented radiological and ALARA work controls.

Problem Identification and Resolution (1 sample)

The inspectors evaluated whether problems associated with ALARA planning and

controls were identified at an appropriate threshold and properly addressed in the CAP.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03 - 3 samples)

a. Inspection Scope

The inspectors reviewed the control of in-plant airborne radioactivity and the use of

respiratory protection devices in these areas. The inspectors used the requirements in

10 CFR Part 20, RG 8.15, RG 8.25, NUREG/CR-0041, TS, and procedures required by

TS as criteria for determining compliance.

Inspection Planning

The inspectors reviewed the UFSAR to identify ventilation and radiation monitoring

systems associated with airborne radioactivity controls and respiratory protection

equipment staged for emergency use. The inspectors also reviewed respiratory

protection program procedures and current PIs for unintended internal exposure

incidents.

26

Engineering Controls (1 sample)

The inspectors reviewed operability and use of both permanent and temporary

ventilation systems, and the adequacy of airborne radioactivity radiation monitoring in

the plant based on location, sensitivity, and alarm set-points.

Use of Respiratory Protection Devices (1 sample)

The inspectors reviewed the adequacy of PSEGs use of respiratory protection devices

in the plant to include applicable ALARA evaluations, respiratory protection device

certification, respiratory equipment storage, air quality testing records, and individual

qualification records.

Problem Identification and Resolution (1 sample)

The inspectors evaluated whether problems associated with the control and mitigation of

in-plant airborne radioactivity were identified at an appropriate threshold and addressed

by PSEGs CAP.

b. Findings

No findings were identified.

2RS4 Occupational Dose Assessment (71124.04 - 2 samples)

a. Inspection Scope

The inspectors reviewed the monitoring, assessment, and reporting of occupational

dose. The inspectors used the requirements in 10 CFR Part 20, RGs, TSs, and

procedures required by TSs as criteria for determining compliance.

Inspection Planning

The inspectors reviewed: RP program audits; National Voluntary Laboratory

Accreditation Program (NVLAP) dosimetry testing reports; and procedures associated

with dosimetry operations.

Source Term Characterization (1 sample)

The inspectors reviewed the plant radiation characterization (including gamma, beta,

alpha, and neutron) being monitored. The inspector verified the use of scaling factors to

account for hard-to-detect radionuclides in internal dose assessments.

External Dosimetry

The inspectors reviewed: dosimetry NVLAP accreditation; onsite storage of dosimeters;

the use of correction factors to align electronic personal dosimeter results with NVLAP

dosimetry results; dosimetry occurrence reports; and CAP documents for adverse trends

related to external dosimetry.

27

Internal Dosimetry (1 sample)

The inspectors reviewed: internal dosimetry procedures; whole body counter

measurement sensitivity and use; adequacy of the program for whole body count

monitoring of plant radionuclides or other bioassay technique; adequacy of the program

for dose assessments based on air sample monitoring and the use of respiratory

protection; and internal dose assessments for any actual internal exposure.

Special Dosimetric Situations

The inspectors reviewed external dose monitoring of workers in large dose rate gradient

environments.

Problem Identification and Resolution

The inspectors evaluated whether problems associated with occupational dose

assessment were identified at an appropriate threshold and properly addressed in the

CAP.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation (71124.05 - 1 sample)

a. Inspection Scope

The inspectors reviewed performance in assuring the accuracy and operability of

radiation monitoring instruments used to protect occupational workers during plant

operations and from postulated accidents. The inspectors used the requirements in

10 CFR Part 20; RGs; applicable industry standards; and procedures required by TSs as

criteria for determining compliance.

Inspection Planning

The inspectors reviewed: Salem Station UFSAR; RP audits; records of in-service survey

instrumentation; and procedures for instrument source checks and calibrations.

Walkdowns and Observations

The inspectors checked the calibration and source check status of various portable

radiation survey instruments and contamination detection monitors for personnel and

equipment.

Calibration and Testing Program

The inspectors reviewed the calibration standards used for portable instrument

calibrations and response checks to verify that instruments were calibrated by a facility

that used National Institute of Science and Technology traceable sources.

28

Problem Identification and Resolution (1 sample)

The inspectors verified that problems associated with radiation monitoring

instrumentation (including failed calibrations) were identified at an appropriate threshold

and properly addressed in the CAP.

b. Findings

No findings were identified.

Cornerstone: Public Radiation Safety (PS)

2RS7 Radiological Environmental Monitoring Program (71124.07 - 2 samples)

a. Inspection Scope

The inspectors reviewed the Radiological Environmental Monitoring Program (REMP) to

validate the effectiveness of the radioactive gaseous and liquid effluent release program

and implementation of the Groundwater Protection Initiative (GPI). The inspectors used

the requirements in 10 CFR Part 20; 40 CFR Part 190; 10 CFR Part 50, Appendix I; TSs;

Offsite Dose Calculation Manual (ODCM); Nuclear Energy Institute 07-07; and

procedures required by TSs as criteria for determining compliance.

Inspection Planning

The inspectors reviewed: Salem and Hope Creek Stations 2015 annual radiological

environmental and effluent monitoring reports; REMP program audits; ODCM changes;

land use census; UFSAR; and inter-laboratory comparison program results.

Site Inspection (1 sample)

The inspectors walked down various passive dosimeter and air and water sampling

locations and reviewed associated calibration and maintenance records. The inspectors

observed the sampling of various environmental media as specified in the ODCM and

reviewed any anomalous environmental sampling events including assessment of any

positive radioactivity results. The inspectors reviewed any changes to the ODCM. The

inspectors verified the operability and calibration of the meteorological tower instruments

and meteorological data readouts. The inspectors reviewed environmental sample

laboratory analysis results, laboratory instrument measurement detection sensitivities,

laboratory quality control program audit results, and the inter- and intra-laboratory

comparison program results. The inspectors reviewed the groundwater monitoring

program as it applies to selected potential leaking structures, systems, or components;

and 10 CFR 50.75(g) records of leaks, spills, and remediation since the previous

inspection.

Groundwater Protection Initiative Implementation

The inspectors reviewed: groundwater monitoring results; changes to the Groundwater

Protection Initiative (GPI) program since the last inspection; anomalous results or

missed groundwater samples; leakage or spill events including entries made into the

decommissioning files (10 CFR 50.75 (g)); evaluations of surface water discharges; and

29

PSEGs evaluation of any positive groundwater sample results including appropriate

stakeholder notifications and effluent reporting requirements.

Identification and Resolution of Problems (1 sample)

The inspectors evaluated whether problems associated with the REMP were identified at

an appropriate threshold and properly addressed in PSEGs CAP.

b. Findings

No findings were identified.

4.

OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151)

Unplanned Scrams, Unplanned Power Changes, and Unplanned Scrams with

Complications (6 samples)

a. Inspection Scope

The inspectors reviewed PSEG submittals for the following Initiating Events Cornerstone

PIs for the period of July 1, 2015 through June 30, 2016.

Unit 1 & 2 Unplanned Scrams

Unit 1 & 2 Unplanned Power Changes

Unit 1 & 2 Unplanned Scrams with Complications

To determine the accuracy of the PI data reported during those periods, inspectors used

definitions and guidance contained in Nuclear Energy Institute (NEI) Document 99-02,

Regulatory Assessment Performance Indicator Guideline, Revision 7. The inspectors

reviewed PSEG operator narrative logs, maintenance planning schedules, condition

reports, event reports, and NRC integrated IRs to validate the accuracy of the

submittals.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution (71152 - 4 samples)

.1

Routine Review of Problem Identification and Resolution Activities

a. Inspection Scope

As required by Inspection Procedure 71152, Problem Identification and Resolution, the

inspectors routinely reviewed issues during baseline inspection activities and plant

status reviews to verify PSEG entered issues into their CAP at an appropriate threshold,

gave adequate attention to timely corrective actions, and identified and addressed

adverse trends. In order to assist with the identification of repetitive equipment failures

and specific human performance issues for follow-up, the inspectors performed a daily

30

screening of items entered into their CAP and periodically attended condition report

screening meetings. The inspectors also confirmed, on a sampling basis, that, as

applicable, for identified defects and non-conformances, PSEG performed an evaluation

in accordance with 10 CFR Part 21.

b. Findings

No findings were identified.

.2

Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a semi-annual review of site issues to identify trends that

might indicate the existence of more significant safety concerns. As part of this review,

the inspectors included repetitive or closely-related issues documented by PSEG in the

CAP and repetitive or closely-related issues that may have been documented by PSEG

outside of the CAP, such as trend reports, PIs, major equipment problem lists, system

health reports, MR assessments, and maintenance or CAP backlogs. The inspectors

also reviewed PSEG CAP database for the first and second quarters of 2016 to assess

notifications written in various subject areas (equipment problems, human performance

issues, etc.), as well as individual issues identified during the inspectors daily condition

report review (Section 4OA2.1). The inspectors reviewed the PSEG CAP trending data,

conducted under LS-AA-125, to verify that PSEG personnel were appropriately

evaluating and trending adverse conditions in accordance with applicable procedures.

a. Findings and Observations

No findings were identified.

Equipment Reliability (Steady)

The inspectors documented an adverse trend in either equipment reliability or unplanned

entries into TS shutdown limiting conditions for operation (LCO) in each of the previous

four semi-annual trend review periods (IRs 05000272; 311/2014003, 2014005, 2015002

and 2015004). In February 2016, in response to PSEGs unplanned LCO performance

goal not being met, PSEG performed Common Cause Evaluation (CCE) 70184208,

Unplanned Shutdown LCO Goal Not Met. The CCE was completed in April of 2016, with

the following results:

A trend of data over an 18-month period from August 2014 through January 2016

identified 68 unplanned shutdown LCOs, which far exceeded the station goal of

no more than 8 in a 12-month rolling average. PSEGs CCE concluded:

1) 15 LCO entries were attributed to faulty parts; 2) 10 entries were attributed to

equipment not being repaired in a timely manner; and 3) more follow up

evaluations were warranted:

o Work Group Evaluation (WGE) 70185245, Follow up Evaluation from

Unplanned shutdown LCOs, was performed to further evaluate the

10 entries attributed to equipment not being repaired in a timely manner.

PSEG attributed the cause to ineffective development and

31

implementation of equipment reliability strategies to ensure reliability until

long-term elimination or mitigating actions were in place. Actions were

assigned to develop bridging strategies for Plant Health Committee items

and rollout to Station Oversight Committee (SOC) and Management

Review Committee (MRC) an expectation that if an unplanned LCO

occurs, a causal evaluation should be performed.

The inspectors noted some improvement in the area of unplanned entries into TS LCOs

in recent months; specifically, 44 unplanned shutdown LCOs occurred from June 2015

to April 2016, but only seven occurred in the last 3 months of this 10 month period. The

inspectors determined that the adverse trend of equipment failures did not constitute a

performance deficiency, because the trend, by itself, did not constitute a violation of any

NRC requirement. The inspectors inspected individual equipment failures as ROP

baseline inspection samples documented in other sections of this report.

Main Control Room Deficiencies (Steady with recent improvement)

The inspectors noted an adverse trend in main control room deficiencies, as evident by a

Red station performance metric dating back to mid-2015, when the station metric was

redefined to align with the current industry metric. Specifically, in June of 2016, Unit 1

had 69 and Unit 2 had 45, versus a red performance metric threshold of 16 or more.

However, the inspectors noted recent improvements in this area. Specifically, Unit 1

reduced the backlog from 99 in January 2016 to 69 in June, and Unit 2 reduced the

backlog from 73 before the fall 2015 refueling outage to 45 in June 2016.

Untimely Reportability Determinations (Steady)

In Section 4OA2.2 of IR 2015-004, the inspectors identified that past operability

determinations were untimely in supporting conclusions of LER reportability in 60 days,

and listed multiple examples. In response to a LER 05000311/2016-001-000 being

submitted well beyond 60 days from the occurrence of the event (see Sections 4OA2.3

and 4OA7 of this report), PSEG performed a review under apparent cause evaluation

(ACE) 70183590, to determine the extent of condition relative to missed or late reports

under 10 CFR 50.72 and 50.73. PSEG concluded the following: 1) The execution of

CAP does not support timely completion of evaluation products to support 60-day LER

submittals; 2) SOC and MRC have a low threshold for requesting reportability reviews;

and 3) Salem has a high number of supplemental LERs relative to the industry (four in

2015 versus an industry average of less than one), indicating that CAP does not support

timely cause evaluation completion, which require LERs to be supplemented. The

inspectors noted that PSEGs conclusion 3 above is consistent with a previously

identified trend by the inspectors documented in Section 4OA2.2 of IR 2015002, which

listed a steady increase in CAP evaluation products and subsequent trend of CAP

products falling behind station timeliness goals. As a result of the ACE listed above,

PSEG issued a temporary standing order to develop interim guidance until process

improvements and controls were institutionalized for reportability, assigned corrective

actions to develop procedure improvements and controls for accompanying reportability

reviews, and to develop the appropriate change management plan for process changes

to perform reportability reviews. The inspectors did not identify any actual violations of

10 CFR 50.72 or 50.73 during the performance of this inspection. The timeliness of

reportability determinations remains a minor adverse trend.

32

Status Control and Human Performance Events (Improving)

The inspectors previously documented an adverse trend in status control in Section

4OA2.5 of IR 2014005. In December of 2015, Nuclear Oversight identified an adverse

trend in status control. In February of 2016, PSEG completed a CCE in response to the

adverse trend in plant status control. Additionally, status control was a focus area for the

station in 2016. Since that time, the inspectors noted considerable improvement in the

area of status control. Specifically, as of June 1, 2016, the station achieved 181 status

control event free days. However, in recent months, the inspectors noted several human

performance events that were not classified as status control events, though they reflect

many of the same behavioral breakdowns in standards and fundamentals. Examples

include:

April 17: 1B EDG invalid actuation: During the performance of solid state

protection system testing in Mode 6 (refueling), the 1B EDG unexpectedly started

while an operator in the field was attempting to replace a light bulb on the test

box. PSEG performed an investigation and determined that the most likely

cause was due to the operators finger bumping the block switch during the bulb

replacement, which was enough pressure to allow the test block signal to be

momentarily interrupted. PSEG reported this event as a telephone notification

under 10 CFR 50.73(a)(1) and (a)(2)(iv)(A) on June 15.

April 25: #1 Emergency Compressed Air Compressor trip during leak test -

PSEG performed Quick Human Performance Investigation (QHPI) 70186240 and

determined the operator in the control room did not understand the report from

the equipment operator in the field, and determined that three-way

communication was not used when it should have been.

April 19: 22B circulator bypass valve operated in the wrong direction - PSEG

performed QHPI 71085972 and determined that an equipment operator did not

fully open the 22B circulator outlet valve prior to attempting remote closure of the

22B circulator bypass, which resulted in the bypass valve failing to stroke closed.

March 27: Station Blackout (SBO) air compressor tripped - the equipment

operator did not follow procedure while testing the SBO air compressor, resulting

in a trip of the compressor (20723821).

The inspectors determined that none of the issues above were of more than minor

significance, because none of them resulted in a significant plant transient or loss of a

mitigating system. The inspectors determined that although the trend in events

classified as status control had improved, the behaviors that contributed to them were

still present.

.3

Annual Sample: Unit 2 Auxiliary Feedwater Loop Response Time Exceeded Technical

Specifications

a. Inspection Scope

The inspectors performed an in-depth review of PSEGs identification, evaluation, and

resolution following the discovery that a channel of the 21 AFW pump engineered

safety feature actuation system (ESFAS) automatic actuation logic was inoperable.

33

On November 18, 2015, maintenance personnel compiling test data, collected on

October 18, 2015, during the Unit 2 plant shutdown for the fall 2015 refueling outage,

determined that the pump instrumentation loop time response exceeded test acceptance

criteria. At the time, Unit 2 was shut down in a refueling outage and AFW was not

required. The cause of the slow loop response was due to the isolation valve to the

21 AFW pump discharge pressure transmitter (2PA3450) being closed. The pressure

transmitter provided input into the pump run-out protection and flow control circuit.

The closed isolation valve caused the pressure transmitter to take longer to sense pump

discharge pressure, which resulted in the slow opening of the pump SG flow control

valves (valves 23AF21 and 24AF21). PSEGs investigation determined that the

condition likely existed since April 20, 2015, following the completion of maintenance on

the pressure transmitter. On January 19, 2016, PSEG determined that the condition

was reportable to the NRC. PSEG initiated an ACE to determine the cause of the

untimely review and evaluation of the surveillance data collected on October 18, 2015,

and a WGE to determine the cause of the improperly positioned isolation valve to

pressure transmitter 2PA3450. The inspectors performed an in-depth review of the ACE

and WGE and corrective actions associated with the issues documented in Orders

70183590 and 70182519. PSEG submitted Licensee Event Report (LER)

05000311/2016-001-000, AFW Loop Response Time Exceeded TSs, on March 21,

2016, as an operation or condition which was prohibited by the plants TS. The

inspectors review of the LER is documented in Section 4OA3.1 of this report. Section

4OA7 documents the enforcement aspects related to the LER.

The inspectors assessed PSEGs problem identification threshold, causal analysis,

extent of condition reviews, compensatory actions, and the prioritization and timeliness

of corrective actions to determine whether PSEG was appropriately identifying,

characterizing, and correcting problems associated with these issues and whether the

planned or completed corrective actions were appropriate. The inspectors compared

the actions taken to the requirements of PSEGs CAP and 10 CFR Part 50, Appendix B.

In addition, the inspectors reviewed documentation associated with this issue, and

interviewed engineering and maintenance personnel to assess the effectiveness of

the implemented and planned corrective actions.

b. Findings and Observations

No findings were identified.

Maintenance personnel compiling 21 AFW pump loop time response test data identified

the slow response times for valves 23AF21 and 24AF21, and entered this issue into the

CAP as NOTF 20710947. During their review, PSEG identified that the instrument

isolation valve for the 21 AFW pump discharge pressure transmitter (2PA3450) was

closed versus the required position of open. The improperly positioned valve was

promptly placed into the required open position. PSEG entered the improperly

positioned valve into the CAP as NOTF 20709417, and performed a prompt investigation

and a WGE. The inspectors determined that action taken by PSEG upon discovery of

the slow response times for valves 23AF21 and 24AF21 were prompt and appropriate.

The inspectors reviewed Order 70182519, which documented the WGE for instrument

isolation valve for 2PA3450 being found in the incorrect position. Although the actual

cause of the improperly positioned isolation valve was indeterminate, PSEG concluded

that the condition most likely existed since April 20, 2015, when maintenance was last

34

performed on 2PA3450. Corrective actions included plans to install human factors tools

(i.e., additional measure devices) on all transmitter isolation valves located in both the

Unit 1 and 2 AFW instrumentation panels. The inspectors concluded that PSEGs

planned corrective action was appropriate.

The inspectors reviewed the timeline of events from the collection of test data on

October 18, 2015, until the submittal of the LER for the condition prohibited by TS

related to the slow instrument loop response time for the 21 AFW pump. The inspectors

concluded that information was available to PSEG personnel on November 20, 2015,

that the condition was potentially reportable when the cause was determined to be due

to the incorrectly positioned instrument isolation valve to 2PA3450. However, the

required LER was not submitted until March 21, 2016.

The inspectors reviewed PSEGs investigation into the reportability timeliness issue, as

documented in Order 70183590. PSEG determined that the cause was due to work

tracking assignments not being made to facilitate identification and completion of the

required past operability review in accordance with Engineering standard practice. The

normal practice to evaluate issues for potential past operability/reportability is for the

SOC to assign a technical evaluation to Engineering to review. In this case an action

item was assigned to Engineering versus a technical evaluation. The due dates for

action items are allowed to be extended by the assignee whereas, the process of

extending technical evaluations has more stringent controls. Therefore, the priority of

the action item was not established at the correct threshold by the assigned

engineering supervisor. This resulted in extensions of the due date for the past

operability/reportability review. PSEGs corrective actions taken or planned included

issuance of an Operations standing order, which provided additional interim guidance for

performing past operability and reportability reviews, and to develop process

improvements and controls for accomplishing past operability and reportability reviews.

The inspectors concluded that the actions taken or planned appeared to appropriately

address the reportability timeliness issue. In accordance with IMC 0612, "Power

Reactor Inspection Reports," the above timeliness of reportability issue constituted a

violation of minor significance that is not subject to enforcement action in accordance

with the Enforcement Policy.

As discussed in Order 70183590, PSEG recognized that the SOC inappropriately

assigned an action item versus the more appropriate technical evaluation to

Engineering for the past operability/reportability review. The inspectors observed that

actions taken by PSEG did not directly address the shortfall of the SOC in this case.

The inspectors noted that there was a low level assignment for the SOC to evaluate for a

human performance crew clock reset; however, the clock reset was determined to not be

necessary. The inspectors noted that the other actions taken or planned discussed

above appeared to be adequate to address the inappropriate extensions of past

operability and reportability reviews.

In NRC Inspection Report 05000272, 05000311/2015004, dated February 10, 2016, a

problem identification and resolution adverse trend was documented related to past

operability determinations being untimely in supporting conclusions of LER reportability

within sixty days. The inspectors concluded that the untimely past operability and

reportability review of the failed 21 AFW pump instrument loop time response test as an

additional example of the adverse trend identified in NRC IR 05000272,

35

05000311/2015004 and updated in Section 4OA2.2 of this report. At the end of this

inspection period, PSEG had not entered this adverse trend into their CAP.

.4

Annual Sample: Struthers-Dunn Relay Failures in Safety-Related Applications

a. Inspection Scope

The inspectors performed an in-depth review of PSEGs ACE and corrective actions

associated with NOTF 20681569 related to a 21 containment spray (CS) pump failure to

start. The 21 CS pump failed to start on October 2, 2015, during post-maintenance

testing following scheduled maintenance. The 21 CS pump failure to start was

investigated by PSEG during subsequent troubleshooting. Additionally, a failure modes

and causal table determined the most likely cause for the failure to start was from a

starting relay high contact resistance. PSEG postulated that contact contamination

created a high resistance condition that was subsequently cleared due to the wiping

action of the relay contact. The starting relay was a Struthers-Dunn Model 219BBX-240

and was replaced. The failed relay was sent for failure analysis to an offsite laboratory.

The lab was unable to repeat the high resistance contact operation that was observed at

Salem. The lab functional testing did not yield any deficiencies or failure mechanisms.

The inspectors assessed PSEGs problem identification threshold, causal analyses,

technical analyses, extent of condition reviews, and the prioritization and timeliness of

corrective actions to determine whether PSEG was appropriately identifying,

characterizing, and correcting problems associated with this issue. The inspectors

reviewed the circumstances of this relay failure issue to ascertain the appropriateness of

corrective actions. The inspectors also assessed PSEGs corrective actions to prevent

recurrence. The inspectors compared the actions taken to the requirements of PSEGs

CAP and 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. In addition, the

inspectors reviewed documentation associated with this issue, including condition

reports, and interviewed engineering personnel to assess the effectiveness of the

planned and implemented corrective actions.

b. Findings and Observations

No findings were identified.

The Struthers-Dunn relays in critical applications were all replaced in 1996 and 1997

during extended unit shutdowns. From about 2000 to 2015, Salem experienced

Struthers-Dunn relay failures in critical applications at about one MR functional failure

per year. In May 2013, after a Struthers-Dunn relay failure associated with the

15 containment fan cooling unit (CFCU), PSEG developed extensive corrective actions

to revise preventive maintenance (PM) templates and determine an appropriate

replacement periodicity. An accelerated testing program was a corrective action and

completed in March 2014 to determine the number of relay operations when the contacts

gold flashing began to wear away exposing the silver base. Exposing the silver contact

base leads to a corrosion condition called sulfidation creating a high resistance between

relay contacts. Offsite laboratory analysis of previous Struthers-Dunn relays had

identified worn gold flashing and sulfidation.

PSEG determined from the accelerated relay testing program that Struthers-Dunn relays

in CFCU applications should be replaced every 10 years. The CFCUs have more

36

frequent equipment on/off cycles compared to other critical Struthers-Dunn applications.

PSEG determined all other Struthers-Dunn relay replacements should be replaced at

20 years. PSEG established the 20 year replacement interval based on 400 relay

operations for the equipment considered. However, the inspectors noted that for some

relay applications, major gold flashing wear or wiping resulting in areas of exposed silver

was observed from the accelerated failure testing results at just 350 relay operations.

PSEG generated notification 20734284 in response to the inspectors observation for

resolution and to reevaluate the intended 20 year replacement periodicity.

The corrective action due dates for the final PM templates are due in August 2016.

PSEG accelerated and completed the Struthers-Dunn relay replacements in all CFCU

applications. The inspectors noted that if PSEG finalizes a 20 year replacement for

non-CFCU applications, considering that all Struthers-Dunn relays were replaced in

1996 to 1997, then all Struthers-Dunn relays would now or in the near term require

replacement. PSEG initiated notification 20734280 in response to the inspectors

observation for resolution.

.5

Annual Sample: Unexpected Number of Degraded Baffle-Former Bolts Discovered in

the Unit 1 Reactor Pressure Vessel

a. Inspection Scope

The inspectors performed an in-depth review of PSEGs technical evaluation and

corrective actions associated with NOTF 20726264 for baffle-former (baffle) bolts found

with indications of degradation during the spring 2016 Salem Unit 1 24th refueling outage

(1R24). PSEG performed ultrasonic examinations of the baffle bolts in accordance with

their procedures in response to recent industry operating experience and 1R24 visual

examination results indicating 18 visually damaged baffle bolts. After an unexpected

number of degraded baffle bolts were discovered, PSEG staff entered the issue into their

corrective action program as NOTF 20727538 and reported the issue to the NRC as

Event Notification 51902 on May 3, 2016, because the as-found number and

location of degraded bolts, which were mainly concentrated in three of the eight baffle

assemblies, represented an unanalyzed condition. PSEG staff completed corrective

actions to replace 189 of 192 potentially degraded baffle bolts on Unit 1. As

documented in Section 1R18, PSEG did not remove and replace three bolts that were

potentially degraded due to difficulties encountered during the removal/replacement

process.

The baffle bolts help secure vertical plates (also referred to as baffle plates) inside the

reactor vessel, which then forms a structure surrounding the reactor fuel assemblies to

orient the fuel and to direct coolant flow through the core. A sufficient number of baffle

bolts are required to remain intact to secure the baffle plates in place so as to not affect

control rod insertion or impede emergency core cooling flow during postulated accident

conditions. Bolt heads that separate and are no longer held in place by bolt lock-tabs

can also become a loose parts concern.

The inspectors assessed whether PSEG acceptable baffle bolt pattern analysis for

Unit 1 was completed in accordance with the NRC-approved methodology and provided

appropriate structural margin for the next cycle of operation to ensure the Unit 1 baffle

plates will remain in place during both normal operation and limiting postulated accident

conditions. The inspectors also assessed whether PSEGs evaluations of the baffle

37

bolts installed in Salem Unit 2 were technically sufficient to conclude the Unit 2 baffle

assembly will perform as intended until the next planned refueling outage, at which time

PSEG plans to examine the bolts. The inspectors reviewed PSEGs procedures for

determining the functionality and operability of degraded systems, components and

structures as they relate to Unit 2. Additionally, the inspectors interviewed PSEG

engineering personnel and contractor staff to discuss the results of PSEGs technical

evaluations and to assess the effectiveness of the implemented and planned corrective

actions.

The inspectors assessed PSEGs problem identification threshold, cause analyses,

extent of condition, compensatory actions, and the prioritization and timeliness of

PSEGs corrective actions to determine whether PSEG staff were properly identifying,

characterizing, and correcting problems associated with this issue and whether the

planned or completed corrective actions were appropriate. The inspectors compared the

actions taken to PSEGs corrective action program, operability determination process,

and the requirements of 10 CFR Part 50, Appendix B. The inspectors observed portions

of baffle bolt replacement activities at Unit 1 and reviewed the final visual examination of

the baffle bolts and plates once the work was completed.

b. Observations

The NRC responded to the initial discovery of an unexpected number of baffle bolts

found degraded at Salem Unit 1 by implementing a comprehensive inspection plan

consisting of various baseline inspection samples to assess the extent of the issue and

to determine the necessary NRC actions. A previously planned ISI sample (Refer to

Section 1R08) was expanded to include a review of the capability of the NDE techniques

for ultrasonically testing (UT) the baffle bolts, to evaluate the UT results, and to observe

a portion of bolt replacement activities on-site. Two permanent modification samples

(Refer to Section 1R18) were conducted to review the design change package and

evaluations associated with the new, replacement baffle bolts, and to review the PSEG

design change package documenting the as-left baffle bolting pattern in Unit 1. NRC

resident inspectors reviewed PSEGs foreign material controls and loose parts analysis

(Refer to Section 1R20) to address the potential for missing bolt heads and concluded it

would not impact safe operation of the plant.

NRC Region I based inspectors, accompanied by an expert from the NRC Office of

Nuclear Reactor Regulation, completed this annual problem identification and resolution

inspection sample, to verify that PSEGs evaluations and corrective action to replace

Unit 1 baffle bolts were completed in accordance with NRC approved methodology to

support a conclusion that the Unit 1 baffle assembly meets the plant design basis. The

inspectors also reviewed the adequacy of PSEGs technical evaluations completed to

determine whether there is a reasonable expectation the Unit 2 baffle assembly will

perform as intended during the current operating cycle. The results of this review are

discussed herein and in Section 1R15 of this report.

At the completion of this inspection, PSEGs conduct of a RCE to determine the causes

of the failure of the baffle bolts in Unit 1 was ongoing. The inspectors determined

PSEGs RCE will not be completed until after laboratory tests and analyses, planned for

fall 2016, are performed on a sample of the bolts removed from Unit 1. PSEGs

technical evaluation discussed the cause of the degraded baffle bolts as primarily due to

IASCC. This determination was based on industry operating experience related to baffle

38

bolt failure in both foreign and domestic plants, is a known degradation mechanism and

the operational and physical characteristics of both Salem plants indicate that they are

susceptible to this mechanism. The inspectors reviewed PSEGs technical evaluation

and the supporting operating experience related to baffle bolt failures at other plants.

IASCC is a cracking mechanism that occurs over a long period of time when susceptible

metals are exposed to neutron radiation from the reactor core and stresses as part of

normal design and operation. The inspectors determined PSEG identified the likely

cause of the baffle bolt degradation and their plans to complete a RCE when additional

metallurgical information was available was appropriate.

Following identification of the degraded baffle bolts on Unit 1, PSEGs immediate

corrective action was to analyze the as-found condition and begin replacing bolts that

either had visual indications of bolt failure (protruding bolt head for example), did not

pass UT examination, or were not accessible for UT examination. The as-found number

and pattern of these bolts exceeded the acceptance criteria in the plants analysis that

was prepared in advance of the baffle bolt examinations; therefore, PSEG reported this

discovery to the NRC as an unanalyzed condition in Event Notification 51902 on May 3,

2016. PSEG staff completed corrective actions to replace 189 of 192 potentially

degraded baffle bolts. PSEG did not remove and replace three bolts that were

potentially degraded due to difficulties encountered during the removal/replacement

process. As previously documented in Section 1R18, one bolt had an indication during

ultrasonic testing but was not visibly damaged. The second bolt was inaccessible for

ultrasonic testing, which would have required replacement. The third bolt had

successfully passed an ultrasonic test but had a visual indication on one of the lock bar

welds which may have indicated a crack in the weld.

The inspectors determined that PSEG staff performed an acceptable bolt pattern

analysis that evaluated the replacement bolt pattern for Unit 1. The inspectors found

the results of the analysis accounted for a conservative failure rate of bolts and provided

appropriate margin for one cycle of operation. The inspectors verified that PSEGs

methodology for its acceptable bolt pattern analyses, including its determination of

margin, was consistent with the NRC-approved methodology in topical report

WCAP-15029-NP-A (ML15222A882). The inspectors determined that PSEG staff

tracked corrective actions to re-examine the Unit 1 baffle bolts during the next planned

refueling outage. The inspectors noted the new baffle bolts were made of a material

(316 SS) with improved resistance to IASCC and included an improved design to reduce

the stresses at the head to shank transition, both of which are enhancements compared

to the original bolts.

As part of an extent of condition assessment, PSEG entered NOTF 20727590 in its

corrective action program to evaluate the potential for degraded baffle bolts on Unit 2.

PSEG operators performed an IOD and concluded that the baffle assembly was

operable. PSEG staff performed a subsequent technical evaluation that concluded

Unit 2 would experience less baffle bolt degradation than Unit 1 based on several plant

factors. The inspectors reviewed PSEGs technical evaluations, including the inputs for

the operability determination, and noted that PSEG staff concluded there was not a

degraded condition at Unit 2. In consideration of the guidance in PSEGs operability

procedure and operating experience from Unit 1 and other plants, the NRC issued an

NCV in this report because PSEG did not perform an OPEval for Unit 2 as a follow-up to

the IOD for the potential impact on supported systems controlled by the Technical

Specifications (Refer to Section 1R15).

39

As a corrective action, PSEG staff performed OpEval 2016-015 and demonstrated that

the Unit 2 baffle assembly remained operable. The inspectors concluded that this

supplemental evaluation provided adequate technical justification for the continued

operation of Unit 2 until the next refueling outage in spring 2017, at which time PSEG

plans to examine the baffle bolts. PSEG also implemented compensatory measures to

monitor the reactor coolant system for any signs of fuel leakage, which could be an

indicator of baffle bolt failures and to generate additional contingency actions in

response to indications of increased unidentified leakage or receipt of a metal impact

monitoring system alarm.

The inspectors reviewed Westinghouse Nuclear Safety Advisory Letter NSAL-16-1,

which discussed the results of recent baffle bolt inspections and provided

Westinghouses recommendations on this issue. The letter described the plants as most

susceptible (i.e. Tier 1a) to this degradation as Westinghouse 4-loop reactors limited to

those with a down-flow configuration and using Type 347 stainless steel. A non-

proprietary presentation on the contents of NSAL-16-1 can be found at ML16202A063.

The inspectors noted the recommendation was to complete UT volumetric examination

of the baffle bolts at the next scheduled refueling outage, and that PSEG had already

planned this action for Unit 2. The inspectors determined PSEGs overall response to

the issue was commensurate with the safety significance, was timely, and included

appropriate compensatory actions. The inspectors concluded that the actions completed

and planned were reasonable to address the ongoing aging management of baffle bolts.

4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153 - 4 samples)

.1

Plant Events (2 samples)

a. Inspection Scope

For the plant events listed below, the inspectors reviewed and/or observed plant

parameters, reviewed personnel performance, and evaluated performance of mitigating

systems. The inspectors communicated the plant events to appropriate regional

personnel, and compared the event details with criteria contained in IMC 0309, Reactive

Inspection Decision Basis for Reactors, for consideration of potential reactive inspection

activities. As applicable, the inspectors verified that PSEG made appropriate emergency

classification assessments and properly reported the event in accordance with 10 CFR

50.72 and 50.73. The inspectors reviewed PSEGs follow-up actions related to the

events to assure that PSEG implemented appropriate corrective actions commensurate

with their safety significance.

Unit 1, Baffle to former bolts found broken or degraded on May 3 (EN 51902)

Unit 2, Reactor trip from main turbine trip on June 28 (EN 52048)

b. Findings

No findings were identified.

40

.2

(Closed) LER 05000311/2016-001-000: Auxiliary Feedwater Loop Response Time

Exceeded Technical Specifications

a. Inspection Scope

While evaluating surveillance instrumentation loop time response test data associated

with the 21 AFW pump that was collected during the Unit 2 plant shutdown for the fall

2015 refueling outage, PSEG determined that a channel of the pumps ESFAS

automatic actuation logic was inoperable. In November 2015, PSEG personnel

identified the slow loop response time during surveillance testing. The cause of the slow

loop response was due to the isolation valve to the 21 AFW pump discharge pressure

transmitter (2PA3450) being closed. The pressure transmitter provided input into the

pump run-out protection and flow control circuit. The closed isolation valve caused the

pressure transmitter to take longer to sense pump discharge pressure which resulted in

slow opening of the pump steam generator flow control valves (valves 23AF21 and

24AF21). PSEGs investigation determined that the condition existed since April 20,

2015, following the completion of maintenance on the pressure transmitter. An

engineering review concluded that, although the AFW loop response time test results did

not satisfy TS requirements, the accident analysis assumptions remained valid and the

condition did not result in an unanalyzed condition. This issue is discussed in more

detail in Section 4OA2.1 of this report. No other issues were identified during the review

of the LER. This LER is closed.

b. Findings

The enforcement aspects of this violation are discussed in Section 4OA7.

.3

(Closed) LER 05000311/2016-002-00: Automatic Reactor Trip Due to Main Turbine Trip

a. Inspection Scope

On February 4, Salem Unit 2 automatically tripped from approximately 74 percent power.

Power had been reduced at the beginning of dayshift to support a 500 kV transmission

line outage. The reactor trip was due to a Main Turbine trip caused by a Main Generator

Protection signal initiated by a main generator AVR volts/hertz over excitation protection

relay. All emergency core cooling systems and emergency safeguards feature systems

functioned as expected. PSEG submitted this LER in accordance with 10 CFR 50.73

(a)(2)(iv)(A), "Any event or condition that resulted in manual or automatic actuation of

any of the systems listed in paragraph (a)(2)(iv)(B)," specifically automatic actuation of

the Reactor Protection System and the Auxiliary Feedwater System for this event. The

inspectors reviewed the LER, the associated cause evaluation, and interviewed PSEG

staff. This LER is closed.

b. Findings

Introduction. A Green, self-revealing FIN was identified against MA-AA-716-010,

Maintenance Planning Process, Revision 18, when PSEG WOs did not specify the

appropriate procedure to perform satisfactory modification testing of the main generator

AVR protective relay (model STV1). Consequently, the relay actuated below its design

setpoint on February 4, 2016, resulting in an automatic trip of the Unit 2 main turbine and

reactor.

41

Description. On February 4, 2016, Unit 2 experienced an automatic main turbine and

reactor trip from approximately 74 percent power, initiated by a trip of the main generator

AVR STV 1 relay. The STV1 is designed to protect the main generator, main power

transformers, and auxiliary transformer from over-excitation due to over-voltage

operation, and consists of an adjustable pickup dial setting between 1.8 and

2.5 voltz/hertz (V/Hz), ranging from 108 - 150 V at 60 Hz. PSEG design calculation

ES-7.007, Salem Unit 2 Generator and Transformer Protective Relay Setpoint

Determination, Revision 5, established a design setpoint for the STV1 relay of 138 V at

60 Hz, corresponding to a V/Hz dial setting of 2.3, with an associated time delay of

45 seconds. Just prior to the Unit 2 trip on February 4, the main generator was

operating at approximately 26.1 kV following a manual MVAR adjustment, which

corresponded to 2.175 V/Hz sensed at the STV1. After the Unit 2 trip, PSEG

troubleshooting determined the as-found pick-up value of the STV1 was 2.17 V/Hz. The

post-trip sequence-of-event data showed the STV1 time delay unit picked up 45 seconds

after exceeding 2.17 V/Hz, which tripped the AVR and resulted in a loss of field to the

main generator, thereby causing a turbine trip and coincident reactor trip.

In response to the Unit 2 reactor trip, PSEG performed RCE 70183932, Unit 2

Automatic Reactor Trip on Generator Protection, to determine why the STV1 relay

actuated below the design setpoint. PSEG identified two root causes: 1) setpoint drift

due to a damaged rheostat; and 2) the damaged rheostat was not identified due to an

inadequately planned work order that specified a less than adequate post-modification

test method. PSEG DCP 80109718, Salem Unit 2 AVR Replacement, supplement 10,

documented that a modification test was required for the STV1 relay in accordance

with Relay Department test procedures, which subsequently required the use of an

engineering-approved Relay Test Order (RTO). However, Maintenance Planning

prepared WO 60122561-0014 to perform STV1 modification testing without specifying

the applicable test procedures. MA-AA-716-010, step 4.5.7, states If approved

procedure(s) are available which cover all or part of the work scope, then specify in the

work package to perform work in accordance with the procedure(s). Additionally, step

3.1.1 states, in part, Maintenance Planners are responsible to interface with: System

Engineers for providing supplemental technical direction on a case by case basis as

needed; and Maintenance Shops to obtain information needed to produce an

adequately detailed work package.

Additionally, the RCE determined that WO 60122561-0014 directed the PSEG LTS

department to perform modification testing of the STV1 relay. However, LTS utilized

different testing procedures than the Relay department procedures specified in the DCP.

The LTS modification testing performed on October 5, 2015, did not functionally test the

STV1 relay at its design setpoint of 138 volts at 60 Hz, which corresponded to a dial

setting of 2.3 as discussed above. The RCE determined the manufacturer-specified

acceptance testing required verifying the V/Hz pick-up was within one percent of all V/Hz

adjustable dial settings, whereas the LTS procedure required the V/Hz pickup at a four

percent tolerance on the 2.0 dial setting, or four percent of 120 volts at 60 Hz. The

STV1 relay pickup value from the LTS testing on October 5, 2015, fell outside of the one

percent tolerance specified by the manufacturer, and LTS did not have a technical basis

to support an allowable tolerance of four percent. The RCE determined that returning

the relay to the manufacturer-specified setting of one percent would have required

adjusting the damaged rheostat to a position where the relay would not have functioned,

and therefore would have resulted in a failed acceptance test that would have prevented

42

the relay from being installed in the plant. The inspectors verified that the STV1 RTO

specified a one percent tolerance at the design setpoint of 138 volts at 60 Hz.

Analysis. The inspectors determined that a performance deficiency existed because

PSEG WOs did not specify the appropriate procedure to perform satisfactory

modification testing of the main generator AVR protection relay STV1. This issue was

more than minor since it was associated with the procedure quality attribute of the

Initiating Events cornerstone and adversely impacted its objective to limit the likelihood

of events that upset plant stability (main generator and turbine trip) and challenge critical

safety functions. Specifically, due to a work order that was not planned properly, PSEG

did not test the STV1 relay at the applicable design setpoint and manufacture-specified

tolerance. Consequently, the relay actuated below its design setpoint on February 4,

2016, resulting in an automatic trip of the Unit 2 main turbine and reactor. Using IMC 0609, Attachment 4 and Appendix A, Exhibit 1, the inspectors determined that this

finding was of very low safety significance, or Green, since mitigating equipment relied

up to transition the plant to stable shutdown remained available.

The finding had a cross-cutting aspect in the area of Human Performance, Work

Management, in that the organization implements a work process that includes the need

for coordination with different groups or job activities. Specifically, the PSEG process for

planning the STV1 relay modification test WO included the need for maintenance

planners to coordinate with engineering to provide design setpoint and tolerance

specifications, as well as electrical maintenance departments to verify appropriate test

procedures were specified in the WO. The inspectors determined that PSEG did not

adequately implement the work process in accordance with MA-AA-716-010. [H.5]

Enforcement. MA-AA-716-010, Maintenance Planning Process, Revision 18, step 4.5.7,

states If approved procedure(s) are available which cover all or part of the work scope,

then specify in the WO to perform work in accordance with the procedure(s). Contrary

to the above, PSEG did not specify in the WO to perform work in accordance with

approved Relay department test procedures, and the associated RTO, for modification

testing of the STV1 relay on October 5, 2015. Specifically, due to a work order that was

not planned properly, PSEG did not test the STV1 relay at the applicable design setpoint

and manufacturer-specified tolerance. Consequently, the relay actuated below its

design setpoint on February 4, 2016, resulting in an automatic trip of the Unit 2 main

turbine and reactor. PSEG entered the issue in CAP as notification 20717849 and

performed RCE 70183932. Planned corrective actions included replacing the failed

STV1 relay with a properly tested STV1 relay, verifying other STV relays were

appropriately tested as an extent of condition, and revising LTS department relay test

procedures to ensure all applicable acceptance criteria are incorporated. This finding

does not involve enforcement action because no violation of a regulatory requirement

was identified. Because this finding does not involve a violation and is of very low safety

significance, it is identified as a Finding. (FIN 05000311/2016002-03, Inadequate Work

Order Planning Results in Main Generator AVR STV Relay Trip)

43

4OA5 Other Activities

.1

(Closed) URI 05000272; 311/2015008-01: Inadequate Maintenance Rule System

Performance Criteria (PC) Selection

a. Inspection Scope

In IR 05000272; 311/2015-008, inspectors identified a URI associated with inadequate

Maintenance Rule Performance Criteria selection.

During this review the inspectors noted approximately 25 high safety significant systems

(HSS) with reliability PC greater than two maintenance preventable functional failures

(MPFFs). According to ER-AA-310-1003, Attachment 3, flowchart Process for Selecting

Reliability Performance Criteria, HSS SSCs, with reliability PC greater than or equal to

two MPFFs require SSC past performance documentation. When the inspectors

requested that PSEG provide past performance documentation for the HSS SSCs with

reliability PC greater than two MPFFs, PSEG provided documentation of HSS SSC PC

approval from 1997, when the MRule Program was first implemented by PSEG. The

inspectors determined this documentation did not support the assigned PC, because it

did not consider the last 18 years of SSC past performance.

The inspectors also reviewed ER-AA-310-1007, Maintenance Rule - Periodic (a)(3)

Assessment. Step 5.11.1.4 states to determine that the number of MPFFs allowed per

evaluation period is consistent with the assumptions in the probabilistic risk assessment

(PRA). Contrary to ER-AA-310-1007, step 5.11.4, the last two periodic (a)(3)

assessments performed by PSEG: April 1, 2011, through September 9, 2012; and

October 1, 2012 through June 30, 2014; did not verify that the number of MPFFs allowed

per evaluation period were consistent with the assumptions in the PRA. Additionally,

ER-AA-310-1003, step 4.3.2, states, in part, that unless justified and approved by the

Maintenance Rule Expert Panel, the number of MPFFs selected, as a Reliability PC,

may not be higher than the PRA-supplied number of functional failures.

The inspectors determined that the failure to meet ER-AA-310-1007, step 5.11.4, and

ER-AA-310-1003, step 4.3.2, was a performance deficiency. However, at the time of

inspection, as documented in the IR referenced above, the inspectors did not have the

information needed to determine whether the performance deficiency was more than

minor. The inspectors reviewed PSEGs actions in response to the URI, to determine

whether the performance or condition of HSS SSCs was effectively controlled through

the performance of appropriate preventive maintenance under 10 CFR 50.65(a)(2), and

also to determine if those HSS SSCs being monitored under 10 CFR 50.65(a)(1) were

assigned appropriate goals and monitoring when considered against the appropriate

reliability PC threshold.

b. Findings

No findings were identified.

PSEG captured the performance deficiency associated with the URI in the CAP under

notifications 20694641, 20699573, and 20716722. In response, the PSEG Engineer

performed detailed reviews of all the HSS reliability performance criteria against the

basic event failure assumptions in the most recent PRA model. For any systems that

44

were identified to have reliability performance criteria deviations from the PRA basic

event failure data, performance criteria changes were proposed to more closely align

with the PRA. Any proposed changes to system performance criteria were scheduled

for review by the Maintenance Rule Expert Panel, including a review of system

performance during the last 36 months. The inspectors observed a sampling of the

Expert Panel meetings, and reviewed meeting minutes for several others. Upon

completion of the PSEG system reviews and expert panel meetings, a total of 12 HSS

had reliability performance criteria reductions to more closely align with PRA failure data.

Five of the 12 systems were already being monitored under 10 CFR Part 50.65(a)(1)

prior to the reduction in performance criteria. None of the 12 systems were moved to

(a)(1) as a result of the performance criteria reductions. The inspectors sampled the

performance criteria adjustments to determine if HSS classified under (a)(2) were being

appropriately monitored, and to verify that (a)(1) systems had appropriate goals

assigned. No performance deficiencies were identified. The inspectors determined that

PSEGs scope of actions restored compliance with ER-AA-310-1007, step 5.11.4, and

ER-AA-310-1003, step 4.3.2.

This URI is closed.

.2

License Renewal Commitments Inspection - Phase I Observation of License Renewal

Activities (71003 - 1 sample)

a. Inspection Scope

License renewal inspections verify the license conditions added as part of the renewed

operating license, regulatory commitments, and selected aging management programs,

and are implemented in accordance with 10 CFR Part 54, Requirements for the

Renewal of Operating Licenses for Nuclear Power Plants. This inspection was

completed during 1R24 to observe the implementation of select aging management

program activities that are only available for observation during a refueling outage. This

inspection is described as Phase 1 in NRC Inspection Manual Procedure 71003, Post-

Approval Site Inspection for License Renewal and is intended to be completed during the

last refueling outage prior to a nuclear power facility entering the period of extended

operation.

As part of this review the inspectors observed the implementation of aging management

programs and activities described in the license conditions, and regulatory commitments,

as well as any testing or visual inspections of systems, structures, and components

which are only accessible at reduced power levels or during a refueling outage.

The inspectors observed the ultrasonic thickness inspection of 1S-FWR-P-21-L1, which

is a 6-inch diameter elbow in the Feedwater Recirculation system. The component is

part of the No. 12 SG Feed pumps 24-inch discharge header. The inspectors observed

the test grid being applied and the recording of measurements in accordance with test

procedure OU-AA-335-004 under the flow accelerated program guidance

ER-AA-430-1001 as directed by WO 30285966.

The inspectors also observed the preparation for the replacement of a Moisture

Separator Reheat Drain system 4-inch diameter piping section. The line is the drain

from the No. 11 West Moisture Separator Reheat Main Steam Coil going to the No. 11

West Main Steam Coil Drain Tank. This was the planned replacement of 27 feet of

45

piping with corrosion resistant P22/Chrome Moly material. The work was being

performed on the 140 Turbine deck, under WO 60123316.

The inspectors observed the No. 12C Miscellaneous Drains drain manifold replacement

spool piece. This 12-inch diameter manifold receives three drain lines from the No. 15A,

B, & C Bleed Steam lines and is being replaced with corrosion resistant P22 (Chrome

Moly) material. The replacement was in progress and performed under WO 60123347.

After reviewing WO 60120251, the inspectors observed the removal and evaluation of

random samples of inaccessible Salem Unit 1 containment liner covered by insulation.

The inspectors observed the containment interior liner insulation being removed,

unremediated containment liner sections, and containment liner sections that were

cleaned, brushed, and prepared for panel installation. The inspectors reviewed

ultrasonic thickness data to verify whether the program was in conformance with

American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,

Section XI.

b. Findings and Observations

No findings were identified.

4OA6 Meetings, Including Exit

On July 28, 2016, the inspectors presented the inspection results to Mr. Robert DeNight,

Salem Operations Director, and other members of the PSEG staff. On August 11, 2016,

an additional exit meeting was conducted and the inspectors presented inspection

results specific to the baffle bolt issues in this report to Mr. Eric Carr, Acting Station Vice

President. During the August 11, 2016 exit meeting, PSEG management stated they

may contest NCV 05000311/2016002-02 (Section 1R15), in a written response within

30 days of the date of this inspection report, using the process described in the cover

letter. Additionally, the inspectors verified that no proprietary information was retained

by the inspectors or documented in this report.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by PSEG

and is a violation of NRC requirements which meets the criteria of the NRC Enforcement

Policy, for being dispositioned as an NCV.

TS LCO 3.3.2.1 requires the ESFAS instrumentation channels and interlocks shown

in Table 3.3-3 shall be operable. Table 3.3-3, Function 8, requires two channels of

AFW automatic actuation logic to be operable in Modes 1, 2, and 3. With the

number of operable channels one less than the required number of channels, TS

LCO 3.3.2.1 requires the inoperable channel to be restored to operable status within

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or, be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least Hot

Shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Contrary to TS LCO 3.3.2.1, one less than

the required number of channels of AFW automatic actuation logic were operable

from April 20, 2015, until Unit 2 entered Mode 4 for a scheduled refueling outage on

October 23, 2015. This was due to the 21 AFW pump loop time response being

greater than the allowed TS value because the isolation valve for the pressure

46

override defeat pressure transmitter was in the closed position. PSEG entered this

issue into the CAP as NOTFs 20709417, 20716352, 20710947, and 20711796.

This performance deficiency was more than minor because it was associated with

the human performance attribute of the Mitigating System cornerstone, and

adversely affected the cornerstone objective of ensuring the reliability and capability

of systems that respond to initiating events to prevent undesirable consequences.

The inspectors evaluated this finding using IMC 0609, Appendix A, The Significance

Determination Process for Findings At-Power, Exhibit 2. The inspectors determined

that the finding was of very low safety significance (Green) because the finding did

not represent an actual loss of function of at least a single train for greater than its

TS allowed outage time.

ATTACHMENT: SUPPLEMENTARY INFORMATION

A-1

Attachment

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

J. Perry, Site Vice President

E. Carr, Acting Site Vice President

J. Barkhamer, PSEG Engineer

J. Bergeron, Superintendent of Instrumentation and Controls

T. Cachaza, Senior Regulatory Compliance Engineer

R. Cary, Environmental Coordinator

L. Clark, Instrument Supervisor

B. Daly, Nuclear Environmental Affairs, Sustainability

D. Denelsbeck, RP Support Supervisor

B. Down, PSEG Engineer

P. Essner, System Engineer

P. Fabian, Salem Steam Generator Engineer

T. Giles, Salem ASME Section XI Program Owner

F. Grenier, RP Supervisor, Dosimetry

M. Hassler, Salem Radiation Protection Manager

B. Kerkorian, Salem Steam Generator Supervisor

D. Kolasinski, Senior Engineer

A. Kraus, Manager, Nuclear Environmental Affairs

T. MacEwen, Principal Compliance Engineer

J. Mallon, Compliance Director

S. Markos, Manager, Design Engineering

J. Marooney, MPR Engineering Consultant

P. Martitz, Technical Support Superintendent

J. Melchionna, Engineering Services

R. Moore, System Engineering Branch Manager

D. Mora, Salem NDE Program Coordinator

G. Morrison, Mechanical Engineer

T. Mulholland, Shift Operations Manager

A. Ochoa, Senior Compliance Engineer

B. Ohmert, System Engineer

T. Oliveri, Salem Unit 1 and Unit 2, NDE Manager

J. ORourke, Regulatory Affairs

J. Owad, Design Engineering

M. Phillips, Regulatory Assurance

M. Pyle, Chemistry Manager

N. Ruvis, Westinghouse

B. Sebastian, Manager Fire Protection/Industrial Safety

J. Stairs, Manager Plant Engineering

C. Wend, Radiation Protection Manager

D. Yilgic, Lead Engineer Quality Control Chemistry

A-2

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Open 05000272/2016002-01

URI

Baffle-Former Bolts with Identified

Anomalies (Section 1R08)

Open and Closed 05000311/2016002-02

NCV

Failure to Follow Operability

Determination Procedure for Unit 2

Baffle-Former Bolts (Section 1R15)05000311/2016002-03

FIN

Inadequate Work Order Planning Results in

Main Generator AVR STV Relay Trip

(Section 4OA3.3)

Closed

05000272:311/2015-008-01

URI

Inadequate Maintenance Rule System

Performance Criteria Selection

(Section 4OA5)

05000311/2016-001-00

LER

Auxiliary Feedwater Loop Response Time

Exceeded Technical Specifications

(Section 4OA3.1)

05000311/2016-002-00

LER

Automatic Reactor Trip Due to Main

Turbine Trip (Section 4OA3.3)

A-3

LIST OF DOCUMENTS REVIEWED

  • Indicates NRC-identified

Section 1R01: Adverse Weather Protection

Procedures

SC.OP-SO.500-0001, Trip-A-Unit Scheme Operation, Revision 10

OP-AA-108-107-1001, Electric System Emergency Operations and Electric Systems Operator

Interface, Revision 4

Notifications

20731655*

20731657*

20731658*

20731659*

20731662

20731729*

20731735*

Section 1R04: Equipment Alignment

Procedures

SC.MD-ST.125-0003, Quarterly Inspection and Preventive Maintenance of Units 1, 2, & 3 125

Volt Station Batteries, Revision 30

S1.OP-ST.CAN-0007, Refueling Operations - Containment Closure, Revision 25

S2.OP-SO.SW-0005, Service Water System Operation, Revision 42

S2.OP-SO.ABV-0001, Auxiliary Building Ventilation System Operation, Revision 25

S2.OP-SO.SJ-00001, Preparation of the Safety Injection System for Operation, Revision 19

OP-SA-102-106, Salem Operations Master List of Timed Actions, Revision 0

OP-AA-108-103, Locked Equipment Program, Revision 4

Notifications

20702800

20707221

20724871

20729878*

20732182

20732551

20732785*

20732994*

20733091

Drawings

205337, Sheet 1, No. 2 Unit Auxiliary Building - Ventilation, Revision 43

205242, Sheet 1, No. 2 Unit Service Water Nuclear Area, Revision 81

205242, Sheet 2, No. 2 Unit Service Water Nuclear Area, Revision 76

Maintenance Orders/Work Orders

50180453

50182431

60125981

60129782

Section 1R05: Fire Protection

Procedures

FP-SA-2542, Pre-Fire Plan Unit 2 Spent Fuel/Component Cooling Heat Exchanger and Pump

Area, Revision 0

FP-SA-2552, Pre-Fire Plan Unit 2 Boric Acid Evaporator Unit & Chemistry Area, Revision 0

FP-SA-2651, Pre-Fire Plan Unit 2 Service Water Intake Structure, Revision 0

FP-SA-2555, Pre-Fire Plan Unit 2 Diesel Generator Area, Revision 0

FP-SA-2556, Pre-Fire Plan Unit 2 Inner Piping Penetration Area & Chiller, Revision 0

A-4

Notifications

20723743

20730150*

20732820*

20732836*

Section 1R07: Heat Sink Performance

Notifications

20726947

20727041

20727041

Maintenance Orders/Work Orders

30255437

Section 1R08: In-service Inspection

NDE Procedures

Liquid Penetrant Examination Procedure, OU-AA-335-002, Revision 3

Nondestructive Examination Procedure, Manual Ultrasonic Examination of Vessel Nozzle Inner

Radius Regions, Procedure Number 54-ISI-132-011, 1/27/2011

Nondestructive Examination Procedure, Ultrasonic Examination of Austenitic Piping Welds,

Procedure Number 54-ISI-836-014, 8/21/2013

Areva NP Inc., Nondestructive Examination Procedure, Multi-Frequency Eddy Current

Examination of Tubing, Procedure Number 54-ISI-400-021, 6/12/2013

Notifications

20682192

20694861

20697140

20697577

20697669

20699820

20699859

20699910

20704139

20707057

20707057

20707125

20712181

20712774

20713572

20713573

20713849

20713849

20714082

20716581

20720745

20722494

20724667

20725857

20726340

20726743

Maintenance Orders/Work Orders

60114705

60123261

60126260

Evaluations

70178672

70178814

70178821

70179375

70183001

70185980

Self Assessments

Check-In Self-Assessment, Salem INPO PWR Materials Review, 7/30/2015

NDE Records

Salem Unit 1, Liquid Penetrant Report No. PT-16-002, 11-RHRHEX Vessel Support, 4/15/16

(Summary No.205170)

Salem Unit 1, Liquid Penetrant Report No. PT-16-001, Pipe Lugs 8-RH-2116-10PL-1 thru 4,

4/15/16 (Summary No. 263631)

Salem Unit 1, Liquid Penetrant Report No. PT-16-004, Pipe to Penetration IA, Component 12

SJ-2152-36PS-4, 4/19/16 (Summary No. 263904)

A-5

Salem Unit 1, Liquid Penetrant Report No. PT-16-003, Inlet Nozzle-to-Pump (11 Charging

Pump), Component 6-CV-2111-14R1, 4/15/16 (Summary No. 220757)

Salem Unit 1, Liquid Penetrant Report No. PT-16-005, PIPE TO VALVE (11CS48)

component ID: 8-CS-2114-60, 4/15/16 (Summary No. 356640)

Design Change Package

80092579, Salem Unit 1 - Steam Generator Bowl Drain Repair, SG 11, 12, 13, and 14 (removal

of Alloy 600 and associated 82/182 weld material from each SG Channel Head (SGCH)

bowl drain plugs

PSEG NUCLEAR VTD NUMBER: 900013(019), Title Stress Analysis of Tube-Tubesheet Weld

AREVA RSG, 11/23/15; Calculation Summary Sheet, 7/25/2015.

PSEG Nuclear Work Order 70172201; Areva Reanalysis of Salem Steam Generator tube-to-

tubesheet joint as a friction joint and to provide a revised SG stress analysis to PSEG for

record purposes

WO #60123261, including weld history sheet; Replace SISJ - !SJ248 & 2SJ249

PSEG NUCLEAR LLC VTD NUMBER: AREVA 902739 (001); Salem Unit 1 SG Condition

Monitoring for 1R22 AND Final Operational Assessment for Cycles 23 & 24; 8/8/13

Drawings: 02-9124528D, Salem Unit 1 Steam Generator Channel Head Drain

Modification, Revision 001

Drawings: 1512E32, Salem REPLACEMENT Steam Generator General Layout; Salem

Unit 1 Steam Generator Channel Head Drain Modification, Revision 1

Drawing 02-9124526B, Revision 001, Steam Generator Channel Head Drain Plug

Document No.: 51-9207624-000, Salem Unit 1 SG Condition Monitoring for 1R22 and Final

Operational Assessment for Cycles 23 & 24

Other Documents

NRC Regulatory Issues Summary 2016-02, Design Basis Issues Related To Tube-To-

Tubesheet Joints in Pressurized-Water Reactor Steam Generators, March 23, 2016

PSEG NUCLEAR LLC VTD Number: 9000(019); AREVA Stress Analysis of Tube-Tubesheet

Weld-AREVA, Vendor Number 32-9235210-001

Section 1R11: Licensed Operator Requalification Program

Other Documents

SG-1624, Risk Management, SGFP Trip, SGTR, dated 05/21/16

Section 1R12: Maintenance Effectiveness

Procedures

ER-AA-310-1004, Maintenance Rule - Performance Monitoring, Revision 14

Notifications

20689987

20729117*

20730512*

20730513*

20731038*

20732228*

Drawings

265029, Circ Water Swgr Bldg. 125VDC DC Distribution System, Revision 5

A-6

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

Procedures

OP-AA-108-116, Protected Equipment Program, Revision 12

Notifications

20723781

20724495

20725030*

20725036

20726192

20727564

20727565

20728242

20731749

20733122

Maintenance Orders/Work Orders

60128649

Other Documents

ACE 20723873, 11 CFCU Low Speed Breaker Back-Flashed

Section 1R15: Operability Determinations and Functionality Assessments

Calculations, Analysis, Engineering Evaluations, and Specifications

MPR Associates Letter "Salem Service Water Discharge Header - Disposition of Degraded

Joints", (0108-0471-0007, Rev 1), 6/3/2016

MPR Associates Letter, Salem PCCP Bell-and-Spigot Joint Degradation-Supplemental

Information to (MPR-2650 Revision 0), 10/26/05

MPR Associates Letter, Salem Service Water Discharge Header - Disposition of Degraded

Joints (0108-0471-0007, Rev 0), 4/29/2016

MPR Calculation 0108-0333-JEM-01, Structural Evaluation of Service Water Piping Thinned

Joints, Revision 0

PSEG VTD 326511-001, "Structural Evaluation of Service Water Piping Thinned Joints"

PSEG VTD 326511-002, "Service Water

PSEG VTD 326511-003, "Service Water WEKO Seal Structural Repair Relief Request RAI

Response Technical Input"

PSEG VTD 326511-004, "Request for Use of Mechanical Repair System in Degraded Service

Water Pipe Joints - Input for Response to NRG Request for Additional Information dated

October 29, 2013"

S-C-SW-MEE-1975, Salem Units 1 & 2 Concrete Service Water Pipe Joints - Acceptance

Criteria, Revision 0

Drawings, Wiring Diagrams, and Piping and Instrumentation Diagrams

205243, Sheet 1, Auxiliary Building Control Air, Revision 49

0108-0471-0007, Salem Service Water Discharge Header - Disposition of Degraded Joints,

4/29/2016

Evaluations

70097092

70097514

70103845

70131286

70144770

Notifications

20724198

20726264

20727538

20727590

20726001

20726320

20727126

20727354

20727430

20727678

20729040

20730485*

20727242

20727261

A-7

Procedures

CC-AA-309-101, Engineering Technical Evaluation, Revision 10

OP-AA-108-115, Operability Determinations & Functionality Assessments, Revision 4

LS-AA-120, Issue Identification and Screening Process, Revision 13

LS-AA-125, Corrective Action Program, Revision 21

NO-AA-10, Quality Assurance Topical Report (QATR), Revision 84

S1.OP-PT.CA-0001, Emergency Control Air Compressor Functional Test, Revision 18

S1.OP-LR.CA-0005, Leak Rate Test 1CA920, Revision 1

SC.OP-LB.DF-0001, Diesel Fuel Oil Testing Program, Revision 3

Maintenance Orders/Work Orders

30265178

50140453

50154389

50154555

50158970

50172136

60115402

Miscellaneous

Inspection Manual Chapter 0326, Operability Determinations & Functionality Assessments for

Conditions Adverse to Quality or Safety, dated December 3, 2015

Event Notification 51902, Anomalies Identified during Visual Inspection of Reactor Vessel

Internals, dated May 3, 2016

70187161, Extent of Condition Review for Salem Unit 2 Susceptibility to Baffle Bolt Failure,

Revision 0

70187161, Extent of Condition Review for Salem Unit 2 Susceptibility to Baffle Bolt Failure,

Revision 1

OpEval 2016-015, Potentially Degraded Baffle-Former Bolts in Salem Unit 2, Revision 0

80117136, Salem Unit 1 Baffle to Former Bolt Replacement, Revision 0

80117136, Salem Unit 1 Baffle to Former Bolt Replacement, Revision 1

S2016-156, 50.50 Screen: DCP 80117136, Salem Unit 1 Baffle to Former Bolt Replacement,

Revision 0

S2016-156, 50.50 Screen: DCP 80117136, Salem Unit 1 Baffle to Former Bolt Replacement,

Revision 1

80117136 SUP01, Map of Degraded Bolt Locations, Revision 0

Westinghouse LTR-RIDA-16-125, Rev. 2, Salem Unit 1 Baffle Bolting One Cycle Replacement

Pattern Summary Letter, dated May 31, 2016

WCAP-18058-P, Determination of Acceptable Baffle-Former Bolting for Salem Units 1 and 2,

Revision 0

ML13093A382, Request for Relief from ASME Code Defect Removal for Service Water Buried

Piping, 4/3/2013

ML13227A338, PSEG Response to Request for Additional Information- Relief Request SC-14R-

133, Alternative Repair for Service Water System Piping, 8/15/13

ML14016A123, PSEG Response to Request for Additional Information (RAI 31 and RAI 32) -

Relief Request SC-14R-1 33, Alternative Repair for Service Water System Piping, 1/8/14

ML14058A228, PSEG Response to Request for Additional Information (RA133 - RAI36)-Relief

Request SC-14R-133, Alternative Repair for Service Water System Piping, 2/27/14

ML14085A482, PSEG Response to Request for Additional Information (RAJ 37) - Relief

Request SC-14R-133, Alternative Repair for Service Water System Piping, 3/26/14

ML14097A029, Salem Nuclear Generating Station, Units 1 And 2- Safety Evaluation of Relief

Request No. SC-14R-133 for the Alternative Repair for Service Water System Piping (TAC

NOS. MF1375 AND MF1376), 4/8/2014

A-8

Modifications

80110461

Other Documents

ML13093A382, Request for Relief from ASME Code Defect Removal for Service Water Buried

Piping, 4/3/2013

ML13227A338, PSEG Response to Request for Additional Information- Relief Request SC-14R-

133, Alternative Repair for Service Water System Piping, 8/15/13

ML14016A123, PSEG Response to Request for Additional Information (RAI 31 and RAI 32) -

Relief Request SC-14R-1 33, Alternative Repair for Service Water System Piping, 1/8/14

ML14058A228, PSEG Response to Request for Additional Information (RA133 - RAI36)-Relief

Request SC-14R-133, Alternative Repair for Service Water System Piping, 2/27/14

ML14085A482, PSEG Response to Request for Additional Information (RAJ 37) - Relief

Request SC-14R-133, Alternative Repair for Service Water System Piping, 3/26/14

ML14097A029, Salem Nuclear Generating Station, Units 1 And 2- Safety Evaluation of Relief

Request No. SC-14R-133 for the Alternative Repair for Service Water System Piping (TAC

NOS. MF1375 AND MF1376), 4/8/2014

Section 1R18: Plant Modifications

Condition Reports

20733528

20733526

20726264

20735142

Other Documents

80117136, Design Change Package for Salem Unit 1 Baffle-to-Former Bolt Replacement,

Revision 0

80117378, Item Equivalency Evaluation for Replacement Baffle Bolts, dated 6/2/2016

EVAL-16-19, Salem Unit 1 Baffle-Former Bolt Replacement 1R24, Revision 0

LTR-RIAM-16-39, Transmittal of Westinghouse Specification 70041 EB to PSEG, dated

5/4/2016

S2016-156, 50.59 Screening Form for DCP 80117136, Revision 0

WCAP-18144-P, Generic Replacement Type 316 Cold-Worked Baffle-Former Bolt Qualification

for 4-Loop Downflow Plants, Revision 0

Procedures

54-ISI-364-00, Remote Underwater In-Vessel Visual Inspection of Reactor Pressure Vessels,

Vessel Internals, and Components in Pressurized Water Reactors, dated August 22,

2000

54-ISI-372-005, Remote Underwater In-Vessel Visual Inspection of Baffle to Former Bolts and

Baffle Edge Bolts, dated September 23, 2011

54-UT-108-001, Ultrasonic Inspection of Internal Hex Head Baffle Bolts, dated April 24, 2011

GBRA 104650, Work Instruction Bolt Removal, Revision D

GBRA 173122, Repair and Inspection Sequence Plan for Baffle-former Bolt Replacement at

NPP Salem, Revision 00

A-9

Miscellaneous

180-9257342-000, NDE Services Final Report, Salem Unit 1, 1R24 Baffle to Former Plate Bolt

Inspection Report, dated June 2, 2016

51-9256526-000, Technical Justification for Internal Hex Head E Baffle to Former Bolts

Volumetric Examination at Westinghouse 4-Loop Reactors, dated April 25, 2016

IVVI-101, 01RF Examination Summary Record, VT-3 of Upper Core and Support Plate, dated

5/9/2001

Inservice Inspection Results, Bolt ID 5-55-C, dated May 3, 2016

Inservice Inspection Results, Bolt ID 6-75-C, dated April 30, 2016

NDE Personnel Qualification and Certification, VT-1, 2, & 3, Employee 16657, dated March 7,

2016

NDE Personnel Qualification and Certification, VT-1, 2, & 3, Employee 114882, dated March 4.

2015

MRP-228, Materials Reliability Program: Inspection Standard for PWR Internals- 2012 Update,

Revision 1

54-UT-108-001, Ultrasonic Inspection of Internal Hex Head Baffle Bolts, Revision 1

80117136, Salem Unit 1 Baffle to Former Bolt Replacement, Revision 0

80117136, Salem Unit 1 Baffle to Former Bolt Replacement, Revision 1

S2016-156, 50.59 Screen: DCP 80117136, Salem Unit 1 Baffle to Former Bolt Replacement,

Revision 0

S2016-156, 50.59 Screen: DCP 80117136, Salem Unit 1 Baffle to Former Bolt Replacement,

Revision 1

80117136 SUP01, Map of Degraded Bolt Locations, Revision 0

Westinghouse LTR-RIDA-16-125, Rev. 2, Salem Unit 1 Baffle Bolting One Cycle Replacement

Pattern Summary Letter, dated May 31, 2016

Westinghouse LTR-RIDA-16-125, Rev. 3, Salem Unit 1 Baffle Bolting One Cycle Replacement

Pattern Summary Letter, dated July 11, 2016

WCAP-18058-P, Determination of Acceptable Baffle-Former Bolting for Salem Units 1 and 2,

Revision 0

WCAP-18144-P, Generic Replacement Type 316 Cold-Worked Baffle-Former Bolt Qualification

for 4-Loop Downflow Plants, Revision 0

VEN-16-041, Remote Visual Examination: Baffle-former Bolts (Core Side), dated July 27, 2016

Section 1R19: Post-Maintenance Testing

Procedures

SC.MD-PM.CBV-0002, CFCU Motor Heat Exchanger Internal Inspection, Revision 20

SC.MD-PM.SW-0012, Enecon Tubesheet Cladding System, Revision 13

SC.IC-TI.ZZ-0104, Configuration Control for NUS Model MTH801 Summators, Revision 32

S2.IC-CC.RCP-0058, 2FT-542 #24 Steam Generator Flow Protection Channel I, Revision 42

Notifications

20273570

20670175

20672463

20723478

20723652

20723765

20724185

20724217

20725095

20725111

20726481

20727534

Maintenance Orders/Work Orders

30205173

60120462

60128697

60129161

A-10

Evaluations

70171681

Section 1R20: Refueling and Other Outage Activities

Procedures

LS-AA-119-1003, Calculating Work Hours, Revision 7

MA-AA-716-008-1010, Reactor Services Project FME Plan, Revision 2

S1.OP-IO.ZZ-0006, Hot Standby to Cold Shutdown, Revision 37

S1.OP-TM.ZZ-0001, Reactor Coolant System Pressure - Temperature Curves, Revision 4

SC.OP-DL.ZZ-0001, Reactor Coolant System Heatup/Cooldown Log, Revision 9

SC.OP-DL.ZZ-00012, Pressurizer Heatup/Cooldown Log, Revision 5

Notifications

20723957

20725589*

20725843

20725856

20725917

20726061*

20726121

20726355

20727113

20727298

20727697

20729566

Other Documents

1R24 Shutdown Safety Evaluation and Approval, dated 03/25/16

Section 1R22: Surveillance Testing

Procedures

S2.OP-ST.RHR-0001, Inservice Testing - 21 Residual Heat Removal Pump, Revision 29

S2.RA-ST.RHR-0001, Inservice Testing 21 Residual Heat Removal Pump Acceptance Criteria,

Revision 12

S1.OP-ST.SSP-0001, Manual Safety Injection - SSPS, Revision 32

Notifications

20725279*

20725282*

20725581

20725603

20725936

20726147

20726148

20726342

20728892*

20728962*

20728963*

Maintenance Orders/Work Orders

50182657

Other Documents

Unit 1 Operator logs for April 17 and 18, 2016

Section 1EP6: Drill Evaluation

Procedures

NC.EP-EP.ZZ-0405, Emergency Termination - Redaction - Recovery, Revision

S2.OP-AB.Fuel-0001, Fuel Handling Incident, Revision 5

S2.OP-AB.CW-0001, Circulating Water System Malfunction, Revision 36

S2.OP-AB.CVC-0001, Loss of Charging, Revision 9

Notifications

20733529

20733001

A-11

Other Documents

S16-01, Salem All Facilities Training Drill, 06/16/16

Section 2RS1: Access Control to Radiologically Significant Areas

Procedures

RP-AA-301, Radiological Air Sampling Program, Revision 6

RP-AA-460, Control for High and Very High Radiation Areas, Revision 17

RP-AA-463, High Radiation Area Key Control, Revision 4

RP-AA-401-1001, Special Instruction for Highly Radioactive In-core Components, Revision 0

RP-SA-103, Radiological Control of Reactor Cavity and Spent Fuel Pool Operations, Revision 1

RP-AA-210, Dosimetry Issue, Usage, and Control, Revision 13

RP-AA-401, Operational ALARA Planning and Control, Revision 13

Other Documents

Audits

Locked High Radiation Key Inventory Logs

Radiation Protection Job Guides (7 through 14)

Radiological Survey data (various)

Radiation Protection Plant Radionuclide Evaluation

Corrective Action Documents (various Notifications)

Section 2RS2: Occupational ALARA Planning and Controls

Procedures

RP-AA-401, Operational ALARA Planning and Control, Revision 13

CY-AP-120-1030, Estimating RCS Crud Release for Refueling Outage, Revision 1

S1. CH-IO.ZZ-111(Z), Salem Unit 1 Shutdown Chemistry Plan, Revision 8

Other Documents

Refueling Outage Radiological Performance Report

ALARA Plans (various)

Radiation Protection Job Guides (7 through 14)

ALARA Work In-process Reviews

Outage Chemistry Control Plan

1R24 Hard Gamma Projection

Corrective Action Documents (various Notifications)

Section 2RS3: In-plant Airborne Radioactivity Control and Mitigation

Procedures

RP-SA-103, Radiological Control of Reactor Cavity and Spent Fuel Pool Operations, Revision 1

RP-AA-220, Annual Bioassay Review, Revision 9

RP-AA-301, Radiological Air Sampling Program, Revision 6

RP-AA-401, Operational ALARA Planning and Control, Revision 13

NF-AA-430, Failed Fuel Action Plan, Revision 8

A-12

Other Documents

Radiological Source Term Data - 10 CFR 61 waste stream report

Airborne Radioactivity Sampling Results (various)

Corrective Action Documents (various Notifications)

Section 2RS4: Occupational Dose Assessment

Procedures

RP-AA-401, Operational ALARA Planning and Control, Revision 13

Other Documents

Radiation Protection Job Guides (7 through 14)

General Source Term Data (various)

Corrective Action Documents (various notifications)

Section 2RS5: Radiation Monitoring Instrumentation

Procedures

RP-AA-301, Radiological Air Sampling Program, Revision 6

RP-AA-504, Routine Operation of the Radiation Protection Gross Counting facility

Other Documents

Instrument Source Check and Operability data (various)

Corrective Action Documents (various notifications)

Section 2RS7: Radiological Environmental Monitoring Program

Procedures

RP-AA-228, 10 CFR 50.75(g0 and 10 CFR 50.72.30(d) Documentation, Revision 3

EN-AA-170-500, Meteorological Monitoring System Calibration and Maintenance (Metrological

Tower), Revision 1

EN-AA-170-1000, Radiological Environmental Monitoring Program (REMP) and Meteorological

Program (MET) Implementation, Revision 1

EN-AA-1001, REMP Vendor Dosimetry and Laboratory QA Program

EN-AA-170-4000, Radiological Ground water Protection program Implementation, Revision 0

EN-AA-170-4160, Station RGPP Controlled sample Points, Revision 0

EN-AA-170-4200, Disposal of Water from Excavation projects, Revision 0

EN-AA0170-4300, Investigative Process for Evaluation of Anomalous Tritium Data from On-site

Wells, Revision 1

CY-AA-170-400, Radiological Ground water protection program, Revision 4

AD-LTS-10, Laboratory and Testing Service (LTS) Quality Assurance Program, Revision 4

Instruction NASSV-1.2.2NS, Service of Low Volume Sampler, Revision 19

Instruction MLKSA-1.1.2, Collection of Raw Milk samples, Revision 12

Instruction VGTSA-1.1.7, Collection of Vegetable, Vegetation and Fodder Crops, Revision 8

Instruction 1.1.9, Collection of Potable Water Samples, Revision 3

Instruction TLDSV-1.2.1, Installation of Area Monitoring Dosimeters in the Field, Revision 16

Instruction AQUACOLL-1.1.10, Collection of Aquatic samples, Revision 11

Instruction GMSA -1.1.11, Collection of Game samples, Revision 3

Instruction VEGECEN-0.3.2, Salem/Hope Creek Vegetable Garden Census, Revision 6

A-13

Instruction NRESCEN, Salem/Hope Creek Nearest Resident Census, Revision 5

Instruction MLKCEN 0.3.1, Salem/Hope Creek Census of Milk Animals, Revision 6

Instruction H2OSA-1.1.1, Collection of Water Samples, Revision 13

Instruction SOLSA -1.1.3, Collection of Soil Samples, Revision 8

Instruction ESS-1.1.5, Collection of Sediment Samples, Revision 9

Instruction ESFCH -1.1.6, Pickup of Fish and Crab Samples, Revision 7

Other Documents

Salem and Hope Creek Offsite Dose Calculation Manuals (ODCM)

UFSAR Section 11.6, Offsite Radiological Monitoring Program

Hope Creek Nuclear Station Buried and Underground Piping Asset Management Plan,

Revision 0

Salem and Hope Creek 2015 Annual Effluent Releases Reports

NEI-07-07, Structure, System, Component (SCC) Review for Turbine Roof Structure (Hope

Creek)

Salem and Hope Creek Annual Radiological Environmental Monitoring Reports

Salem/Hope Creek Meteorological Program Status Report (2014, 2015)

Salem/Hope Creek Metrological Tower Updated Vegetation Review, June 3, 2016

Comparison of 2015 Atmospheric Dispersion Factors for Salem and Hope Creek, dated

March 28, 2016

Chemistry, Radwaste, Effluent and Environmental Monitoring Audit Report, NOSA-SLM-16-04,

May 11, 2016

2016 Self-Assessment REMP Program Inspection

Teledyne Brown Environmental Service Annual Quality Assurance Report

GEL 2015 - Annual Quality Assurance Report (REMP)

Residential Survey, dated December 22, 2015

Milk Animal Survey dated December 2015

Vegetable garden Survey dated August 2015

Calibration Data (Dry Gas Meters 61182898, 14522708, 2424590)

Calibration Data (Laminar Flow Element 16300942)

Global Solutions Annual Testing, dated May 26, 2015

Passive Environmental Dosimetry Calibration data

Ground Water Monitoring Data and RGPP Data

Salem/Hope Creek Part 61 Analysis Review, dated April 27, 2016

Salem Remedial Action Plan Progress Reports

Corrective Action Documents (various Notifications)

Ground Water Monitoring Data

Corrective Action Documents (various Notifications)

Section 4OA2: Problem Identification and Resolution

Condition Reports

20724198

20726264

20727538

20727590

20728329

20732892

20731786

20725142

20736630

Maintenance Orders/Work Orders

70136205

70140618

70154315

70168067

70168874

70180750

70182469

70182519

70183590

70183629

A-14

Miscellaneous

Westinghouse LTR-RIDA-16-125, Rev. 2, Salem Unit 1 Baffle Bolting One Cycle Replacement

Pattern Summary Letter, dated May 31, 2016

Westinghouse LTR-RIDA-16-125, Rev. 3, Salem Unit 1 Baffle Bolting One Cycle Replacement

Pattern Summary Letter, dated July 11, 2016

WCAP-18058-P, Determination of Acceptable Baffle-Former Bolting for Salem Units 1 and 2,

Revision 0

Non-Proprietary Safety Evaluation of WCAP-17096-NP, Revision 2, Reactor Internals

Acceptance Criteria Methodology and Data Requirements (TAC No. ME4200). (ADAMS

Accession No. ML16061A243), dated May 3, 2016

Westinghouse Calculation Note, CN-RIDA-15-34, Rev. 4, Units 1 and 2 Acceptable Baffle-

Former LOCA and Seismic Analysis, dated May 16, 2016

Westinghouse Calculation Note CN-RIDA-15-64, Rev. 2, Salem Units 1 and 2 Acceptable

Baffle-Former Bolting Pattern Fuel Grid Impact Analysis, dated May 16, 2016

Event Notification 51902, Anomalies Identified during Visual Inspection of Reactor Vessel

Internals, dated May 3, 2016

80117136, Salem Unit 1 Baffle to Former Bolt Replacement, Revision 0

80117136, Salem Unit 1 Baffle to Former Bolt Replacement, Revision 1

S2016-156, 50.50 Screen: DCP 80117136, Salem Unit 1 Baffle to Former Bolt Replacement,

Revision 0

S2016-156, 50.50 Screen: DCP 80117136, Salem Unit 1 Baffle to Former Bolt Replacement,

Revision 1

80117136 SUP01, Map of Degraded Bolt Locations, Revision 0

Westinghouse LTR-RIDA-16-112, Rev. 0, Summary of Salem Unit 1 Baffle-Former Bolt Real-

time Analysis Results, dated May 11, 2016

WCAP-18058-P, Determination of Acceptable Baffle-Former Bolting for Salem Units 1 and 2,

Revision 0

Westinghouse LTR-RIAM-16-38 Rev. 0, Salem Unit 1 Real-Time Analysis Results for

LOCA/Seismic Dynamic Analysis and Fuel Grid Impact Analysis, dated May 3, 2016

Westinghouse LTR-RIAM-16-39 Rev. 0, Transmittal of Westinghouse Specification 70041 EB to

Public Service Enterprise Group, dated May 4, 2016

Information Notice 98-11, Cracking of Reactor Vessel Internal Baffle-former Bolts in Foreign

Plants, dated March 24, 1998

Eval-16-19, Westinghouse Electric Company 10 CFR 50.59 Applicability Determination, Salem

Unit 1 Baffle-former Bole Replacement 1R24, Revision 0

MRP-228, Materials Reliability Program: Inspection Standard for PWR Internals - 2012 Update,

Revision 1

Unit 1 and 2 Technical Specifications, Revision 28

ACM 16-011, Baffle Plates Monitoring, dated June 17, 2016

ACM 16-011, Baffle Plates Monitoring, dated July 25, 2016

WCAP-15030-NP-A, Westinghouse Methodology for Evaluating the Acceptability of Baffle-

Former-Barrel Bolting Distributions Under Faulted Load Conditions, dated January 1999

NRC Safety Evaluation of Topical Report wCAP-25029, Westinghouse Methodology for

Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted

Load Conditions (TAC No. MA1152), dated November 16, 1998

NRC Letter, Leak Before Break Evaluation of Primary Loop Piping, Salem Nuclear Generating

Station, Units 1 and 2 (TAC NOS. M85799 and M85800), dated May 25, 1994

51-92566526, Technical Justification for Internal Hex Head E Baffle to Former Bolts Volumetric

Examination at Westinghouse 4-Loop Reactors, dated April 28 2016

A-15

54-ISI-364-00, IVVI Inspection Data Sheet Salem 1R14, dated May 8, 2001

Areva Letter, Completion and Status of Octants 1, 2, 3, 4, 5, 6, 7, and 8 (i.e., 1-8), dated May 5,

2016

OTDM 16-005, Salem Unit 2 Baffle to Former Bolting of Reactor Vessel Internals, dated June

16, 2016

WCAP-18144-P, Generic Replacement Type 316 Cold-Worked Baffle-Former Bolt Qualification

for 4-Loop Downflow Plants, Revision 0

Westinghouse LTR-LIS-11-381, LOCA Assessment of Core Coolable Geometry for Grid

Deformation in Peripheral Fuel Assemblies, dated June 27, 2011

Event Notification 51902, Anomalies Identified during Visual Inspection of Reactor Vessel

Internals, dated May 3, 2016

70187161, Extent of Condition Review for Salem Unit 2 Susceptibility to Baffle Bolt Failure,

Revision 0

70187161, Extent of Condition Review for Salem Unit 2 Susceptibility to Baffle Bolt Failure,

Revision 0

Op Eval 2016-015, Potentially Degraded Baffle-Former Bolts in Salem Unit 2, Revision 0

VEN-16-041, Remote Visual Examination Baffle-former Bolts (Core Side), dated July 27, 2016

Procedures

ER-AA-2003, System Performance Monitoring and Analysis, Revision 10

54-ISI-364-00, Remote Underwater In-Vessel Visual Inspection of Reactor Pressure Vessels,

Vessel Internals, and Components in Pressurized Water Reactors, dated August 22,

2000

54-UT-108-001, Ultrasonic Inspection of Internal Hex Head Baffle Bolts, dated April 28, 2016

Notifications

20704666

20706027

20709417

20710340*

20710947

20711723

20711796

20715617

20716352

20716358

20716401

20716402

20716404

20716754

20721375

20726684

20728492*

20730946

20734279*

20734280*

20734281*

20734284*

20734286*

20734856*

Other Documents

S2.OP-ST.SSP-0011(Q), Engineered Safety Features Response Time Testing performed

October 18, 2015

NRC Event Notification 51663

Exelon PowerLabs Report PSE-65422, 07/01/13

Exelon PowerLabs Report PSE-82817, 11/13/13

Exelon PowerLabs Report PSE-00915, 03/18/14

Exelon PowerLabs Report PSE-19717, 10/22/15

Exelon PowerLabs Report PSE-88030, Draft

Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion

Notifications

20733919*

A-16

LIST OF ACRONYMS

10 CFR

Title 10 of the Code of Federal Regulations

AC

alternating current

ACE

apparent cause evaluation

ADAMS

Agencywide Documents Access and Management System

AFW

auxiliary feedwater

ALARA

as low as is reasonably achievable

ASME

American Society of Mechanical Engineers

AVR

automatic voltage regulator

CAP

Corrective Action Program

CCE

common cause evaluation

CFCU

containment fan cooling unit

CFR

Code of Federal Regulations

CS

containment spray

DC

direct current

DCP

design change package

EC

eddy current

ECAC

emergency compressed air compressor

ECCS

Emergency Core Cooling System

ECT

eddy current testing

EDG

emergency diesel generator

EFPY

effective full power years

EPD

electronic personal dosimeter

EPRI

Electric Power Research Institute

ESFAS

engineered safety feature actuation system

FIN

finding

FOSAR

foreign object search and retrieval

GPI

Groundwater Protection Initiative

HRA

high radiation area

HSS

high safety significant systems

HX

heat exchanger

IMC

Inspection Manual Chapter

IOD

immediate operability determination

IR

inspection report

ISI

In-service inspection

IASCC

Irradiation Assisted Stress Corrosion Cracking

kV

kilovolt

LCO

limiting conditions for operation

LER

licensee event report

LHRA

locked high radiation area

LLRT

local leak rate test

LTS

Laboratory and Testing Services

MPFF

maintenance preventable functional failure(s)

MR

maintenance rule

MRC

Management Review Committee

NCV

non-cited violation

NDE

nondestructive examination

NEI

Nuclear Energy Institute

A-17

NOS

Nuclear Oversight

NOTF

notification(s)

NRC

Nuclear Regulatory Commission

NVLAP

National Voluntary Laboratory Accreditation Program

ODCM

Offsite Dose Calculation Manual

PC

performance criteria

PI

performance indicator(s)

PM

preventive maintenance

PRA

probabilistic risk assessment

PSEG

Public Service Enterprise Group Nuclear LLC

QHPI

Quick Human Performance Investigation

RCE

root cause evaluation

RCS

reactor coolant system

REMP

Radiological Environmental Monitoring Program

RFO

refueling outage

RG

regulatory guide

RHR

residual heat removal

RP

radiation protection

RTO

relay test order

RWP

radiation work permit(s)

SBO

station blackout

SDP

significance determination process

SF

spent fuel

SG

steam generator

SI

safety injection

SOC

Station Oversight Committee

SSC

structure, system, and component

SW

service water

TS

technical specification(s)

UFSAR

Updated Final Safety Analysis Report

URI

unresolved item

UT

ultrasonically testing

V/Hz

volt/hertz

VHRA

very high radiation areas

WGE

work group evaluation

WOs

work order(s)