IR 05000387/1998002

Insp Repts 50-387/98-02 & 50-388/98-02 on 980317-0427. No Violations Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support

ML17159A330
Person / Time
Site:	Susquehanna
Issue date:	05/14/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17159A329	List: IR 05000387/1998002 ML17159A328
References
50-387-98-02, 50-387-98-2, 50-388-98-02, 50-388-98-2, NUDOCS 9805220236
Download: ML17159A330 (46)

Text

U. S. NUCLEAR REGULATORYCOMMISSION

REGION I

Docket Nos:

License Nos:

50-387, 50-388 NPF-14, NPF-22 Report No.

50-387/98-02, 50-388/98-02 Licensee:

Pennsylvania Power and 'Light Company 2 North Ninth Street Allentown, Pennsylvania 19101 Facility:

Susquehanna Steam Electric Station Location:

P.O. Box 35 Berwick, PA 18603-0035 Dates:

March 17 through April 27, 1998 Inspectors:

Approved by:

K. Jenison, Senior Resident Inspector B. McDermott, Resident Inspector J. Richmond, Resident Inspector C, Cahill, Reactor Engineer G. Morris, Reactor Engineer J. McFadden, Radiation Specialist Clifford Anderson, Chief Projects Branch 4 Division of Reactor Projects 9805220236 9805i4 PDR ADOCK 05000387

PDR

EXECUTIVE SUMMARY Susquehanna Steam Electric Station (SSES), Units 1 5 2 NRC Inspection Report 50-387/98-02,50-388/98-02 This integrated inspection included aspects of Pennsylvania Power and Light Company's (PPSL's) operations, engineering, maintenance, and plant support at SSES.

The report covers an 6-week period of resident inspection; in addition, it includes the results of an announced engineering inspection and a radiological control inspection.

~Oerations

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Licensed and non-licensed operator activities were well performed and communicated.

Shift turnovers were observed to be detailed and complete.

(section 01.1)

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The licensee conducted plant operations in accordance with SSES procedures, and established effective equipment alignment and operability.

(section 01.2)

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Operators were observed to respond well to control room alarmed conditions and an infrequently occurring condition (power coast down). Appropriate SSES procedures were adhered to, operability and impact on plant equipment were controlled, and actions were adequately announced and documented.

(section 01.3)

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Safety permits (tagouts) authorized by the control room were properly prepared.

However, due to a weak Work Control review process, the control room operators identified and corrected several errors in the permits, prior to the permit application in the field. (section 01A)

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Operator response to the discovery of a misaligned valve was in accordance with the PPS.L Equipment Status Control Event procedure. Operators responded well and conservatively.

There was no safety impact associated with the misaligned system.

(section 02.1)

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A sample of operator log entries was determined to be complete and accurate.

A specific series of operator log entries was compared to condition report data and determined to be consistent with the data in condition reports.

(section 02.2)

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The two Plant Operations Review Committee (PORC) meetings observed demonstrated that PORC conducted in-depth and conservative reviews and demonstrated a conservative and safe approach.

(section 07.1)

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The licensee identified that portions of the Residual Heat Removal System, designated as closed loop systems which function as redundant containment isolation barriers, had never been leak rate tested as required by the technical specifications.

This was treated as a non-cited violation. (section 08.1)

Executive Summary (cont'd)

Maintenance The planned maintenance activities observed/reviewed were found to be appropriately conducted and controlled.

Procedural control was general in nature.

Interviews with maintenance personnel showed the individuals were knowledgeable, appropriately qualified, and capable of explaining their activities.

(section M1.1)

The surveillance activities observed/reviewed were adequately performed and appropriately controlled.

The surveillance activities were accomplished by qualified and trained personnel.

(section M1.2)

The present material condition and general housekeeping at SSES were determined to be good.

Several minor housekeeping and material condition items that did not affect the system operability were communicated to the licensee for its review.

(section M2.1)

~En ineerin

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The licensee implemented a detailed analysis, qualification, and testing program to address the issue of electrical isolation.

(section E1.2)

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The licensee's Final Safety Analysis Report, for the refuel platform and refueling interlocks, had not been revised following a 1996 modification.

In response to NRC questions, the licensee has issued a CR to review the current design control process which allows partial modification closeouts.

(section E2.1)

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The PPRL methodology, which used a test pressure less than the integrated leak rate test pressure, to accomplish closed loop system integrity verification for the suction lines on the high pressure coolant injection and the reactor core isolation cooling systems was consistent with the NRC expectations for leakage testing of these lines. (section E3.1)

Plant Su ort Performance in radiological controls for individual external and internal exposures for 1997 and for 1998 up to April 28 was fullyeffective.

(section R1.1)

Overall, effective performance in the area of radiological controls for radioactive materials, contamination, surveys, and monitoring was evident.

The licensee identified that they had failed to post high radiation areas during work that involved changing exposure conditions.

This was treated as a non-cited violation.

(section R1.2)

Radiological controls for the As-Low-As-Reasonably-Achievable (ALARA)program were performed in an effective manner.

The selection and qualification of contracted radiological control technicians was proceduralized, conducted,

Executive Summary (cont'd)

administered, and documented in a detailed and thorough manner.

The combination of audits, surveillances, corporate assessments, self-assessments, and the problem identification process resulted in a high volume of deficiencies and improvement opportunities being identified and in a low threshold for such identification.

(sections R1.3, R5, and R7)

• TABLEOF CONTENTS EXECUTIVE SUMMARY TABLE OF CONTENTS

. v Summary of Plant Status I. Operations

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01.1 Operator Shift Activities and Turnover

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01.2 Operational Safety System Alignment 01.3 Operator's Response to Alarmed, Unexpected and Infrequently Performed Situations 01.4 Safety Tagouts - Permits Operational Status of Facilities and Equipment 02.1 Equipment Status Control Event - Valve Misalignment.'. ~...

02.2 Operator Logs............

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Operator Knowledge and Performance.....

04.1 Operability Determinations and Condition Report Action Items Quality Assurance in Operations 07.1 Plant Operations Review Committee Activities Miscellaneous Operations Issues..........................

08.1 Licensee Event Report Review.......................

08.2 Followup of Open Items.......... ~......... ~...

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M1.1 Preplanned Maintenance ActivityReview

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M1.2 Surveillance Test ActivitySample Reviews.... ~......

Maintenance and Material Condition of Facilities and Equipment

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M2.1 Equipment Material Condition and Housekeeping.......

M2.2 NRC Information Notice 96-67, Vulnerability of Emergency Generators to Fuel Oil/Lubricating Oil Incompatibility....

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Diesel

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E1 E2 E3 EB Conduct of Engineering

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~ 1 Seismic Strapping for Lead Shielding E1.2 Electrical Isolation Between Class 1E and non-Class C'fcults

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Engineering Support of Facilities and Equipment E2.1 Unit 1 Refuel Platform Engineering Procedures and Documentation E3.1 Leak Rate Testing of Closed Loop Systems...

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Miscellaneous Engineering Issues..............'....

E8.1 Followup of Open Items............ ~......

1E Electrical II. Engineenng...................................................

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Table of Contents (cont'd)

V. Plant Support I

R1 Radiological Protection and Chemistry (RPRC) Controls R1.1 Radiological Controls-External and Internal Exposure R1.2 Radiological Controls-Radioactive Materials, Contamination, Surveys, and Monitoring................. ~.......

R1,3 Radiological Controls, As-Low-As-Reasonably-Achievable..

R5 Staff Training and Qualification in RPSC...................

R7 Quality Assurance in RPSC Activities....

R8 Miscellaneous RPRC Issues............................

F8 Miscellaneous fire Protection Issues F8.1 Followup of Open Items.......

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.. 21 V. Management Meetings............... ~....,.................

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X1 Exit Meeting Summary..............,................

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.. 21 ATTACHMENT Attachment 1 - Inspection Procedures Used

- Items Opened, Closed, and Discussed

- List of Acronyms Used

Re ort Details Summa of Plant Status Susquehanna Steam Electric Station (SSES) Unit 1 operated at 100% power during the inspection period until commencing a power coast down, plant shutdown, and-commencement of a refuel outage.

Power decreased from 100% on March 29, 1998 to 93% on April 13, 1998, when the unit was shutdown for refueling.

SSES Unit 2 operated at 100% power during the inspection period, except for an 8 day forced outage, to perform repairs for a stator water cooling leak.

Unit 2 commenced power reduction on April 3, 1998, and returned to 100% power on April 14, 1998.

I. 0 erations

Conduct of Operations

'1.1 0 erator Shift Activities and Turnover ao Ins ection Sco e 71707 b.

Routine activities of plant control operators (PCOs), nuclear plant operators (NPOs)

and unit supervisors (USs) were observed throughout the inspection period.

Observations and Findin s Routine operator activities were prescribed, concisely communicated, and performed in accordance with SSES operations department procedures.

Shift turnovers were observed to be detailed and complete.

The inspectors discussed plant conditions with oncoming PCOs and USs following shift turnovers and determined that sufficient information and status were transferred to the oncoming. shift to ensure the safe operation of the units.

In general, communication between PCOs and NPOs was observed to be of good quality.

C.

Conclusions Licensed and non-licensed operator activities were well performed and communicated..Shift turnovers were observed to be detailed and complete.

rToplcal headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor Inspection report outline, Individual reports are not expected to address all outline topic.2 0 erational Safet S stem Ali nment aO Ins ection Sco e 71707 The inspectors observed plant operation and verified that the facilitywas operated safely and in accordance with procedures and regulatory requirements.

Inspectors conducted walkdown inspections of selected safety related systems, and observed equipment alignment and operability.

b.

Observations and Findin s The licensee conducted plant operations in accordance with procedures, and effective controls were implemented for safe plant operation.

Overall equipment operability, material condition, and housekeeping conditions were good.

During plant tours, the alignment/operability of selected safety related systems, engineered safety features, and on-site power sources were verified. A partial walkdown of the following systems was performed:

Unit 1 Containment Instrument Gas System Unit 1 5 2 Hydraulic Control Units Unit 1 4kv Safeguard Switchgear Unit 2 High Pressure Coolant Injection System Unit 2 Reactor Core Isolation Cooling System Unit 2 Core Spray The inspectors identified several minor housekeeping and material condition items, that did not affect system operability, and communicated the items to the licensee for its review (see section M2.1 of this report).

C.

Conclusions The licensee conducted plant operations in accordance with SSES procedures, and established effective equipment alignment and operability.

01.3 0 erator's Res onse to Alarmed Unex ected and lnfre uentl Performed Situations a.

Ins ection Sco e 71707 During'control room observations, the inspectors observed/reviewed PCO and US response to alarmed, unexpected, and infrequently performed situations to determine compliance with Technical Specification (TS) and SSES operating procedure b.

Observations and Findin s Operator responses to the following alarmed conditions were observed to be aggressive'and in accordance with TSs and SSES operating procedures.

The control room staff responded conservatively to several minor plant transients, including a Unit 2 shutdown for a stator water cooling leak repair.

Event 2-98-01 ON-255-001 ON-1 58-001 AR-106-F1 5 AR-204-D1 6 April 3-14, 1998 April 11-12, 1998 March 25, 1998 March 25, 1998 March 20, 1998 Plant Shutdown, Stator Water Cooling Leak I

Rod Control Recirculation Pump Cooling Radiological Waste Dew Point Reactor Building HVAC Pressure Operators responded well to control end of core life recirculation flow and rod manipulation activities to support the coast down of Unit 1. The power maximization and coast down period occurred between March 29, 1998, from 100% power, to April 13, 1998, with power at 93%, immediately prior to the unit shutdown.

C.

Conclusions Operators were observed to respond well to control room alarmed conditions and an infrequently occurring condition (power coast down)'. Appropriate SSES procedures were adhered to, operability and impact on plant equipment were controlled, 'and actions were adequately announced and documented.

01.4 Safet Ta outs - Permits (71707)

The inspectors conducted a review of six safety permits (tagouts) that were in effect and/or being applied to plant systems.

It was determined that the permits authorized by the control room were properly prepared and authorized.

PCOs identified several errors in the subject permits, prior to their issuance, and either resolved the errors or returned the permits to Work Control. The errors identified by the PCOs were insightful and were determined, by the inspectors, to have resulted from weak Work Control permit reviews.

Because the permits were corrected prior to their application in the field, no impact on the safe operation of the units was identified, and no violations of NRC requirements were identified.

Operational Status of Facilities and Equipment 02.1 E ui ment Status Control Event - Valve Misali nment a e Ins ection Sco e 71707 The inspectors observed/reviewed Shift Supervisor (SS), US, and AuxiliarySystem Operator (ASO) response to a licensee identified misalignment of a chemical dispersant system valv b.

Observations and Findin s Operator response, to the discovery of a valve misalignment, was in accordance with the Equipment Status Control Event procedure, NDAP-QA-302. The system was realigned in accordance with OP-111-003 and a condition report (CR) was initiated to determine root cause.

The inspectors determined the licensee responded to the misalignment conservatively and in accordance with established SSES procedures.

There was no safety impact associated with this specific misaligned system.

The inspector reviewed licensee performance data and determined there were 26 misalignment events in 1996, 55 events in 1997, and 5 events thus far in 1998.

c.

Conclusions Operator response to the discovery of a misaligned valve was in accordance with the PP&L Equipment Status Control Event procedure. Operators responded well and conservatively.

There was no safety impact. associated with the misaligned system.

a.

'Ins ection Sco e 71707 During routine control room tours, the inspectors made detailed reviews of PCO, US, Limiting Condition for Operations (LCO), Bypass, Work Around, and Equipment Status logs.

b.

Observations and Findin s Overall, operator logs were determined to adequately document and discuss plant conditions and events.

The inspectors questioned the age and necessity of some work arounds, status controlled plant conditions, and bypasses, but no safety impacts or violations of NRC requirements were identified.

During a review of operator log entries addressing the April 11 - 12, 1998,'Unit 2 startup, the inspectors determined there was a potential performance problem in the control rod drive system.

The unit experienced nine double notch rod movements, one stuck rod, and one uncoupled rod during the startup.

Upon further review the inspectors verified each of the occasions was appropriately addressed by the operators through the implementation of an off-normal procedure, and each event was documented in a CR.

CRs 98-1116, -1122, -1124, -1125, -1129, -1131, and-1123 were reviewed by the inspectors and determined to adequately document the individual conditions.

Following NRC questions, the licensee reviewed the CRs as a group, and determined that there was no impact on the safe operation on the unit.

The inspectors had no further question c.

Conclusions A sample of operator log entries was determined to be complete and accurate.

A specific series of operator log entries was compared to condition report data and determined to be consistent with the data in condition reports.

Operator Knowledge and Performance 04.1 0 erabilit Determinations and Condition Re ort Action Items a.

Ins ection Sco e 71707 The'inspectors reviewed a sample of operability determinations (ODs) and CR action items to determine if degraded conditions were identified, initiallyresolved with conservatism, and long term corrective actions completed.

b.

Observations and Findin s The initial ODs and corrective actions for the following CRs were reviewed:

98-0936

'8-0948 98-0938 98-0890 98-0908 98-0891

"A" Emergency Diesel Generator (EDG) Pole Piece

"A" EDG AirStart System Blow Down Valve

"A" EDG AirStart System Blow Down Valve

"D" EDG Foreign Material in the Intercooler

"D" EDG Oil Cooler Support 4kv Circuit Breakers Isolator IVlodifications C.

Conclusions Six initial operability determinations (ODs) were reviewed, in detail, and were determined to be adequate.

Quality Assurance in Operations 07.1 Plant 0 erations Review Committee Activities a.

Ins ection Sco e 71707 Portions of two Plant Operations Review Committee (PORC) meetings, required by TS, were observed to determine if the activities were aggressive in identifying areas needing improvement and were overall conservativ b.

Observations and Findin s TS 6.5.1 establishes the requirements for PORC.

The following PORC activities were reviewed/observed:

Meeting Number Major Issues Observed 98-3-19

- NDAP-QA-0401, EDG Reliability Program

- 18 Month RHR System Logic Functional Test

- Five Year Post Maintenance Test of the "D" EDG 98-4-07

- Unit 2 Startup

- Suppression Chamber Purge Valve Relay Failure

- SGTS Heater Failure

- Drywell/Suppression Chamber Vacuum Breakers did not open at ) 0.5 psid The inspectors determined, in general, that PORC conducted in-depth reviews and demonstrated a conservative and safe approach to power operation.

PP&L management made conservative presentations to the committees and received, when appropriate, recommendations to improve safety and compliance.

c.

Conclusions The two Plant Operations Review Committee (PORC) meetings observed demonstrated that PORC conducted in-depth and conservative reviews and demonstrated a conservative and safe approach.

Miscellaneous Operations Issues 08.1 Licensee Event Re ort Review (92700)

Closed LER 50-387 97-04-00 Reactor Core Thermal Power Exceeded 102%

On February 1, 1997, the Unit 1 "A" recirculation pump increased speed to its high speed stop, resulting in a reactor power increase to approximately 103.5%.

The operators responded appropriately to the alarmed transient condition and limited the time that the licensed full power limitwas exceeded, to approximately one minute.

The licensee determined that the cause of the transient was a failed fuse in a Bailey controller and replaced the fuse and additional suspect components.

The inspector reviewed the licensee's corrective actions, the analyzed transients described in the Final Safety Analysis Report (FSAR), the operations off-normal response procedure, documented operator actions, and the licensee's efforts to determine the root cause of the transient, The inspector was not able,to definitively conclude the root cause, because the licensee discarded one of the failed components (a fuse) prior to performance of a failure analysis.

Based on a review

of licensee records, operator logs, industry experience review data, and SSES procurement documentation, the inspector determined the probable cause was a failed fuse.

The licensed power was only exceeded for a short period of time, operator response was good, and the failure has not reoccurred.

The peak power, during the transient, was less than the maximum analyzed in the FSAR. Although the root cause of this self-revealing event was obvious, the licensee demonstrated initiative in evaluating the event.

Based on the identified root cause, the licensee did not have a prior opportunity to identify the problem or prevent the event.

Therefore, this non-repetitive event is considered a violation of minor significance and is being treated as a non-cited violation, consistent with Section VII.B.1 of the NRC Enforcement Policy. This LER is closed.

(NCV 50-387/98-0241)

Closed LER 50-387 97-05-01 Closed System Integrity Not Verified by Testing On February 4, 1997, with Unit 1 and Unit 2 at 100/o power, the licensee determined that closed loop systems, functioning as redundant containment isolation barriers, for,the Residual Heat Removal (RHR) full flowtest line and the RHR suppression pool spray line were determined to never have been leak rate tested as required by TS.

The inspector reviewed data from the completed surveillances, design drawings, system alignment checklists, and additional licensee corrective actions.

The inspector determined the corrective actions were adequate.

A similar NRC identified issue was reviewed, in section E3.1 of this report. The normal alignment of the containment isolation valves, separating the closed systems from the containment, was determined to be shut and locked.

These containment isolation valves were determined to have been properly leak rate tested.

The failure to test the closed loop portions of the RHR lines had little safety impact, therefore, this licensee identified and corrected event is considered a violation of minor significance and is being treated as a non-cited violation, consistent with Section VII.B.1 of the NRC Enforcement Policy. This LER is closed.

(NCV 50-387/98-02-02)

Followu of 0 en Items (92901)

Closed VIO 387 388 97-04-03 Failure to Get Prior Approval of a QA Program Change In February 1995, the licensee made changes in the accepted Quality Assurance program description, without prior NRC approval, which reduced the commitment for the span of control on the manager Quality Assurance, previously accepted by the NRC.

The licensee responded to this violation by PPS.L letter PLA-4666, dated September 4, 1997. The letter described organizational changes and reviews that were performed in response to this violation. The response concluded the organizational structure that was established, following receipt of the violation, was not a reduction in the QA program commitment The inspector reviewed; the licensee's response, the implemented PPSL organizational changes, and corrective actions.

Because of the nature of this violation, there was no need for field inspection.

The licensee's corrective actions were determined to be adequate and this violation is closed.

II. Maintenance M1 Conduct of Maintenance M1.1 Pre lanned Maintenance Activit Review a.

Ins ection Sco e 62707 The inspectors observed/reviewers selected portions of preplanned maintenance activities, to determine whether the activities were conducted in accordance with NRC requirements and SSES procedures.

b.

Observations and Findin s Maintenance activities authorized by the following WAs were observed/reviewed during, this inspection.

In addition, selected personnel qualifications, equipment permits (tagouts), procedures, drawings, and/or vendor technical manuals associated with the maintenance activities were also reviewed.

P74433 P70350 P74549 C60660 S71 519 P70900 S81 731 M81 347 P72395 Standby Liquid Control (SLC) Bladder

"D" Emergency Diesel Generator ESS Transformer 1A204041D-Bus Isolator Modification Reactor Cavity Refueling Platform Rigid Pole Hoist Reactor Hydraulic Snubber VOTES High Pressure Coolant Injection Valve Testing Interviews with maintenance personnel showed the individuals involved in the maintenance activities to be knowledgeable and capable of explaining their function.

Field supervision was noted for the observed activities. Although the procedural guidance was general in nature, it routinely incorporated, by reference, a number of other sources of information that was available to workers in the field. No violations of NRC requirements were identified.

C.

Conclusions The planned maintenance activities observed/reviewed were, found to be appropriately conducted and controlled.

Procedural control was general in natur Interviews with maintenance personnel showed the individuals were knowledgeable, appropriately qualified, and capable of explaining their activities.

M1.2 Surveillance Tes Activit Sam le Reviews a.

Ins ection Sco e 61726 The inspectors observed/reviewed selected portions of surveillance activities, to determine whether the activities conformed to TS requirements and SSES

,procedures.

b.

Observations and Findin s Portions of the following surveillance activities were observed/reviewed:

SC-116-102 3/30/98 SO-181-001 4/21/98 SE-259-1 1 6 4/5/98 SE-259-119 4/5/98 SE-259-1 22 4/5/98 SE-259-1 28 4/5/98

"A" RHR SW Weekly Refueling Platform Grapple Operability Local Leak Rate Test (LLRT) on PS-E11-2N010C(Pen.

X-3B)

LLRT on PS-E11-2N011B (Pen. X-32A)

LLRT on PSH-C72-2N002A (Pen. X-3B)

LLRT on PSHL-C72-2N004 (Pen. X-3B)

The observed/reviewed surveillance activities were determined to conform to the requirements of TS and met PPSL administrative requirements (i.e., approvals, personnel qualifications, scheduling, and permits).

Components were properly removed from service and, when appropriate; the TS LCOs were documented and met. The surveillance activities were determined to have been accomplished by qualified and trained personnel.

No violations of NRC requirements were identified.

C.

Conclusions The surveillance activities observed/reviewed were adequately performed and appropriately controlled.

The surveillance activities were accomplished by qualified and trained personnel.

M2 Maintenance and Material Condition of Facilities and Equipment M2.1 E ui ment Material Condition and Housekee in a.

Ins ection Sco e 62707 71707 The inspectors toured the SSES control structure (CS), reactor buildings (RB),

essential service water building, central alarm station, circulating water building, and radiological waste building to establish the material condition, appearance, and accessability of plant equipmen b.

Observations and Findin s Minor findings, with no apparent immediate safety impact, were communicated to the licensee at the completion of each portion of the tour.

In part, these findings included:

- Coaxial radio antenna cable's, not routed in conduit/tray, in CS and RB.

- Standing water on floor/floor drain in Unit 2 RHR heat exchanger room.

- Oscillations on high pressure coolant injection inlet steam pressure gauge.

c.

Conclusions The present material condition and general housekeeping at SSES were determined to be good.

Several minor housekeeping and material condition items that did not affect the system operability were communicated to the licensee for its review.

M2.2 NRC Information Notice 96-67 Vulnerabiiit of Emer enc Diesel Generators to Fuel Oil Lubricatin Oil Incom a ibilit a.

Ins ection Sco e 62707 The inspectors toured portions of the EDG buildings, observed the bulk oil sumps on the "D" EDG, and observed open EDG storage tanks, and storage tank access areas to establish the material operability condition, appearance and accessability of plant equipment.

In addition, because the licensee had a past history of EDG piston scuffing, a review of the licensee Industry Experience Review Program (IERP)

response to NRC Information Notice (IN) 96-67 was conducted, b.

Observations and Findin s The licensee performed a review of IN 96-67 with IERP item 97-017. The licensee determined it used a different type of oil and machine design than described by IN 9667 and, based on the results of its current EDG overhauls, concluded the type of deposits discussed in the NRC IN had not been identified.

The licensee's resolution of the IN issues, specific to SSES EDGs, was acceptable.

c.

Conclusions PPRL's resolution of issues identified by NRC Information Notice 96-67, specific to SSES emergency diesel generator fuel oil quality, were acceptabl III. En ineerin E1 Conduct of Engineering E1.1 Seismic Stra in for Lead Shieldin a.

Ins ection Sco e 37550 Qualification calculations, for strapping used to support lead shielding blankets above RHR pumps and heat exchangers, were reviewed by the inspectors.

b.

Observations and Findin s PPSL calculation PLS-9116, Evaluation of Lead Shielding, was reviewed to determine if adequate controls and design support existed to ensure that lead shielding would not become missile hazards in a seismic event.

The inspectors determined that the licensee adequately justified and qualified the use of heavy duty lashing ties, self locking cable ties and stainless steel banding.

c.

Conclusions PPS.L's application of shielding, on and around safety related equipment, while the SSES units are in condition 1 was adequately supported by engineering analysis, and acceptable.

E1.2 Electrical Isolation Between Class 1E and non-Class 1E Electrical Circuits a.

Ins ection Sco e 37550 The inspector reviewed engineering documents to assess the licensee's analysis, qualification, testing, and surveillance of class 1E isolators.

b, Observations and Findin s On May 22, 1986, at SSES, an open circuit in the "D" Emergency Diesel Generator (EDG) field circuit produced an inductive surge across an instrument current shunt which migrated through the common input card in the plant computer and caused a

common mode ground detection trip on the remaining diesel generators.

These trips are automatically bypassed on an emergency start signal.

A number of Plant Monitoring System (PMS) computer points were also simultaneously lost.

Upon investigation. a small fire was found in the unit 1 computer input/output cabinet and was extinguished.

This event raised the concern of a single fault in non-Class 1E circuits affecting redundant Class 1E circuits. The licensee performed an evaluation of all such circuits and removed the computer inputs associated with the EDG field. These circuits were the only computer inputs developed directly from a current shunt.

Additionally, the licensee submitted several analysis, test results, and proposed

modifications as documented in the NRC Safety Evaluation Report (SER), dated June 28, 1991.

The inspector reviewed the licensee's modification to verify adequate isolation between Class 1E and non-Class 1,E circuits. The defense-in-depth plan that the licensee executed included:

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Requiring all potential high energy inputs into the computer to go through, transducers.

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Installation of circuit protectors (Thyrites) across the 40,000/5A current transformers.

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Installation of Class 1E analog isolators in the class 1E current transformer transducer circuits. These isolators are electrically connected between the output side of the existing transducer and the associated computer inputs.

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Redesigned the computer such that separate computer ch'assis are provided.

for redundant safety circuit inputs to the computer.

The inspector reviewed report SEA-EE-181-R1, Evaluation of Open Current Transformer (CT) Secondary Voltages on the Computer Input Circuits. The inspector found that an analysis assumption made in section 5.1 of that evaluation stated that an open CT circuit is expected to be discovered within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as part of shift surveillance and/or by receipt of erroneous data from the computer.

The inspector interviewed the Operations Supervisor concerning the implementation of this assumption.

The inspector found that there was a.general inspection for broken gauges and instruments conducted every six hours as part of the NPO Plant Log for the Unit 1 and 2 Reactor Buildings. However, operations personnel were not trained to evaluate the indications for open CTs and therefore would not readily identify an open CT and take action in the time prescribed in section 5.1 of SEA-EE-181-R1

~ Also, the inspector found that an open CT would provide a zero value to the computer and not cause any loss of signal alarm.

These computer inputs are only procedurally checked during shift turnover which is every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Therefore, unless the operator was specifically instructed to look at the computer monitor for this condition, it would be doubtful that the open circuit CT condition would be recognized within the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prescribed in section 5.1 of SEA-EE-181-R1.

In response to the inspector's concerns, the licensee conducted shift briefings addressing the design assumption to check for open CT. The licensee issued Hot Box 98-35 to operations personnel addressing class 1E isolator design concerns and issued Condition Report 98-0891 to track the closure of this issue associated with engineering study SEA-EE-181-R1.

This is a violation of 10 CFR 50, Appendix B, Criterion III, in that the design basis was not correctly translated into procedures and instructions.

The inspector determined that this violation was of minor significance because adequate isolation of the class 1E and non-class 1E circuits was demonstrated by the high reliability of

CTs, multiple layers of circuit isolation, and the re-evaluation of the isolation devices to sustain peak voltage.

Therefore, this failure constitutes a violation of minor significance and is being treated as a non-cited violation, consistent with Section IV of the NRC Enforcement Policy.

(NCV 50-387,388/98-02-03)

C.

Conclusion The inspector concluded that the licensee implemented a detailed analysis, qualification, and testing program to address the issue of electrical isolation.

Although the licensee had failed to implement the assumptions of the evaluation of open current transformers into the operating procedures or instructions, the inspector concluded that this was a minor violation.

E2 Engineering Support of Facilities and Equipment E2.1 Unit 1 Refuel Platform aO Ins ection Sco e 37551 The inspectors reviewed the TS surveillance requirements, TS basis, and the FSAR design description associated with the refuel platform refueling interlocks.

b.

Observations and Findin s Between February and March 1996, SSES performed modification 95-3006A to upgrade the Unit 1 refuel platform. The upgrade included installation of a new style General Electric refuel mast and a PLC control system to automate platform operations and fuel movement.

The PLC controller also replaced the original relay logic for the TS related reactor mode switch refueling interlocks.

Post modification acceptance tests were completed in August 1996 and the refuel platform was placed in service and used during the September 1996 Unit 1 refueling outage.

The Unit 1 refuel platform was subsequently used for the March 1997 Unit 2 refueling outage, and is currently in service for refuel operations during the April 1998 Unit 1 refuel outage.

The inspectors determined the FSAR had not been updated, following completion of the modification work, in 1996.

In addition, the vendor manual for the refuel platform, referenced by vendor procedure number in the TS surveillance procedures, was not approved and issued by SSES Document Control until April 16, 1998.

SSES design control requirements do not require FSAR revisions and vendor supplied documents (i.e., new vendor manuals) to be issued until after modification closure.

PPRL was issued an exemption to 10 CFR 50.71(e)(4), in May 1997, which changed the periodicity of FSAR updates from six months after each refueling outage, to once per refueling cycle, based on SSES Unit 2 refuel outage schedule, not to exceed 24 months.

Based on this exemption, no violations of NRC requirements were identified. The licensee has issued a CR to review the current design control process which allows partial modification closeout C.

Conclusions The licensee's Final Safety Analysis Report, for.the refuel platform and refueling interlocks, had not been revised following a 1996 modification.

In response to NRC questions, the licensee has issued a CR to review the current design control process which allows partial modification closeouts.

E3 Engineering Procedures and Documentation E3.1 Leak Rate Testin of Closed Loo S s ems (37551)

TS Table 3.6.3-1 identifies containment isolation valves which do not have redundant isolation valves and instead have a closed system which provides the redundant isolation barrier for the penetration.

Footnote (c) of the TS table states that the closed system integrity is verified by Type A test.

The inspectors reviewed the PP&L testing methodology for the suction lines on the high pressure coolant injection (HPCI) and the reactor core isolation cooling (RCIC) systems.

The HPCI and RCIC suction lines are normally aligned to the condensate storage tank and would not be exposed to the peak containment pressure following a design basis loss of coolant accident (LOCA). Because of this alignment, PP&L pressurizes these lines to 19 psig for leakage testing, the maximum pressure the lines are expected to experience under accident conditions.

SSES engineering work request EWR-16898 concluded that 19 psig was the appropriate test pressure,to be used for the HPCI and RCIC suction lines during an integrated leak rate test, based upon engineering calculation EC-059-1022.

The inspectors questioned the licensee'.s testing methodology, because this pressure is less than the normal Type A test pressure.

On April 7, 1998, a conference call between NRC Region I, NRR and PP&L discussed the licensee's practice.

Subsequently, the NRC determined that PP&L's

'ractice was consistent with their expectations for testing of HPCI and RCIC suction lines, given their normal alignment.

Based on NRC's review, PP&L's testing methodology was acceptable, and the TS Table 3.6.3-1 footnote (c) was met.

PP&L has submitted a TS change as part of the Improved Technical Specifications that willclarify the expected testing methodology.

The inspectors had no further questions.

E8 Miscellaneous Engineering Issues E8.1 Followu of 0 en Items (37551,92903)

U dated URI 50-387 388 98-01-09 Primary Containment Penetration Leak Rate Testing - Notice of Enforcement Discretion On February 2, 1998, SSES received a Notice of Enforcement Discretion (NOED) for containment penetration leak rate tests that were not performed when required.

During the Unit 2 forced outage, penetrations X-3B and X-32A were tested by surveillance procedures SE-259-114through -128. The inspectors reviewed

portions of the 15 individual leak rate tests for the two Unit 2 penetrations, and several Unit 1 penetration test results.

The test results were satisfactory.

This completed all penetration leak rate testing required for both Units.

Pending additional information from the licensee regarding the root cause of the issue of the tests not being performed when required, this unresolved item willremain open.

IV. Plant Su o

R1 Radiological Protection and Chemistry (RP8cC) Controls R1.1 Radiolo ical Controls-External and Internal Ex osure a.

Ins ection Sco e 83750-02 A selective review of radiation work permits (RWPs) and the controls in place for these outage work tasks, individual dosimetry results for 1997 and for 1998 (up to mid April), the results for in vivo blind spike testing of whole body counters (WBCs),

the laboratory accreditation for the thermoluminescent dosimeters (TLDs) used as personnel dosimeters, the use of personal alarming dosimeters (PADs) and the automated access control system for the radiologically controlled area (RCA),

access controls to locked high radiation areas (HRAs), and posting and labeling practices was performed.

Information was gathered through observation of activities, tours of the RCA, discussions with cognizant personnel, and review and.

evaluation of procedures and documents.

b.

Observations and Findin s Performance in radiological controls for individual external and internal exposures for 1997 and for 1998 up to April 28 was fullyeffective.

The RWPs at the Unit 1 main health physics control point contained appropriate and detailed information including task descriptions, radiation survey data, required personal protective equipment, PAD settings, engineering controls, and special instructions for workers and for health physics (HP) technicians.

The controls for the following specific RWPs were observed and were found to be implemented effectively, and supervisory oversight was evident:

~

'8-1001 General entry, work, and tool control on refueling bay,

~

98-1113 Diving activities: suppression pool clean-out/inspection and strainer installation and removal, and

~

98-1306 General entry/work in drywell.

For 1997, the licensee managed individual external exposure accumulation so that no individual received 2 rem or greater total effective dose equivalent (TEDE). Other individual external exposure results for lens of the eye dose equivalent (LDE), for shallow-dose equivalent to the skin (SDE WB), and for any extremity (SDE ME) were

below regulatory requirements.

The maximum individual committed effective dose equivalent (CEDE) and cumulative CEDE were low, 6 millirem (mrem) and 161 mrem, respectively.

For 1998, the current individual exposure results (1794 mrem maximum for TEDE and 13 mrem maximum for CEDE) were well below regulatory limits.

An ~i vivo blind spike test of the two WBCs by Corporate Radiological Services demonstrated agreement with the bias and precision criteria in HPS-N13.30 (Performance Criteria for Radiobioassay) for the recommended radionuclides.

Regarding record personal dosimetry, the laboratory accreditation for the TLD program was current and included ANSI-N13.11 categories I through Vill. PADs were required for RCA access and were used to activate the access control computer terminals and the turnstiles at the main HP control points.

Electronic verification of having read and understood a RWP involved bar scanning one'

identification badge and the RWP.

Access controls for locked HRAs, radiological postings and labels, and radiological surveys available at HP control points met regulatory requirements and were appropriate.

Areas reviewed included the refuel floor, the lower levels of the drywell, and the suppression pool.

The radiation protection staff and technicians were knowledgeable of radiological issues and conditions, well-experienced, and displayed a professional and helpful approach in dealings with the radiation workers.

The number of inside HP control points and the amount of HP coverage provided good radiological control of work, Each RWP required that the worker contact HP for a radiological briefing prior to signing onto the RWP and that the worker explain the work plan and location in detail.

Conclusions Performance in radiological controls for individual external and internal exposures for.

1997 and for 1998 up to April 28 was fullyeffective.

Radiolo ical Controls-Radioactive Materials Contamination Surve s and Monitorin Ins ection Sco e 83750-02 A selective review of the licensee's control of radioactive materials, contamination, surveys, and monitoring, including adequacy of surveys, calibration status of survey and monitoring equipment, the proper use of personal contamination monitors and friskers, adequacy of surveys necessary to control occupational dose received during work that involves changing exposure conditions, tracking of personal contamination events and goals, and the percentage of contaminated area and related goals was performed.

Information was gathered through observation of activities, tours of the RCA, discussions with cognizant personnel, and review and evaluation of procedures and document )

b, Observations and Findin s Overall,.effective performance in the area of radiological controls for radioactive materials, contamination, surveys, and monitoring was evident; however, there was a non-cited violation involving failure to post HRAs. Survey records contained

.

appropriate radiological information for the radiation worker. Selected friskers and radiation survey meters had calibration stickers showing valid current calibration.

Radiation workers were observed appropriately using whole body contamination monitors and frisking hand-carried items using small article monitors at the RCA exit.

Regarding the adequacy of surveys necessary to control occupational dose received during work that involves changing exposure conditions, two licensee-identified events showed a deficiency in this area.

A Condition Report with an event date of April 10, 1998 identified an un-posted HRA. While a radiological control technician was checking a penetration for contamination in the Unit 1 condensate demineralizer pipe way, a HRA was found near a pipe connecting the reactor water cleanup backwash receiving tank (RWCU BWRT) to the radwaste phase separator.

The technician measured 600 mrem per hour at 30 centimeters near the end of the letdown operation.

This pipe had no field identification and was misidentified in plant drawings.

This physical piping configuration had existed since the start of commercial operation.

Based on licensee data, the RWCU BWRT is presently letdown every 14 to 21 days, and the letdown duration is approximately 30 minutes.

Upon identification, the area was properly posted, and a procedure change was initiated to add this area to the list for access and key control during a RWCU BWRT letdown in Procedure No. HP-Hl-073, Notifications of Plant Evolutions and Expected HP Actions.

The licensee reported that there was no history of unexplained personnel exposures connected with this area of the plant. This condition does not apply to the Unit 2 condensate demineralizer pipe way because these lines remain embedded in the wall until entering the turbine building pipe tunnel.

A Condition Report with an event date of April 23, 1998 identified another un-posted HRA. The Unit 1 "A"condensate demineralizer was surveyed and properly posted as a radiation area prior to the outage.

During the outage, the cavity letdown and the residue from draining of the hotwell went to this demineralizer (selected as the sacrificial demineralizer).

A routine weekly survey on April 16, 1998 showed that it was still appropriate to post it as a radiation area.

Then, it was found to be a high radiation area during the subsequent weekly survey on April 23, 1998. The survey recorded 200 mrem per hour at 30 centimeters.

Upon identification, the area was properly posted, and two procedure changes were initiated which involved a change in the survey program procedure and a change in the procedure for plant shutdown and startup so that the sacrificial demineralizer room will be posted and un-posted appropriately.

The licensee reported that there was no

I

~

I'ecord of unexplained personnel exposures connected with this area of the plant during the period from April 16 to April 23, 1998 The two above-described events are two examples of failure to comply with 10 CFR 20.1902(b), Posting of High Radiation Areas.

This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy.

(NCV 50-387/98-02-04)

Goals for personal contamination report (PCR) rate and for percent recoverable contaminated area were established and were being tracked.

C.

Conclusions Overall, effective performance in the area of radiological controls for radioactive materials, contamination, surveys, and monitoring was evident.

The licensee identified that they had failed to post high radiation areas during work that involved changing exposure conditions.

This was treated as a non-cited violation.

R1,3 Radiolo ical Controls As-Low-As-Reasonabl-Achievable a0 Ins ection Sco e 83750-01 A selective review of the licensee's pre-job As-Low-As-Reasonably-Achievable (ALARA)reviews, use of temporary shielding, and radiological goals, projections, and results was performed.

b., Observations and Findin s Radiological controls for ALARAwere being performed in an effective manner.

Numerous ALARApre-job review packages were being developed, and the foilowing completed packages which included the RWP and review were found to be detailed and thorough.

~

1113 Diving Activities: Suppression pool clean-out/inspection and strainer removal and installation

~

1120 DCP 97-3011; Check valve seat mods

~

1315 Drywell Temporary Shielding

~

1319 Insulation Work in the Drywell At a meeting of the Station'ALARA Committee on April 16, 1998, a reduction in both the outage (250 to 242) and annual (480 to 440) person-rem goals for 1998 was approved.

These reductions were based on deferred or eliminated work from the planned work scope for 1998. At the time of this inspection, the outage was

behind schedule by approximately one day, and the current cumulative dose results were tracking well below goal projections C.

Conclusions Radiological controls for As-Low-As-Reasonably-Achievable were being performed in an effective manner.

Staff Training and Qualification in RP&C a0 Ins ection Sco e 83750-02 E

A selective review of the licensee's selection, training, and qualification program for the contracted radiological control technicians hired for the current outage was performed.

Information was gathered through discussions with cognizant personnel and review and evaluation of procedures and docume'nts.

b.

Observations and Findin s The selection and qualification of contracted radiological control technicians was proceduralized, conducted, administered, and documented in a detailed and thorough manner.

Selection was conducted in accordance with the technical specification requirement for two years appropriate experience and with the licensee's procedure which provided instructions for evaluating experience.

Based on the inspector's review of the experience of several contracted radiological control technicians working the outage, the licensee was conservative in their evaluation of past experience to meet the technical specification experience requirement and their internal requirement of three years of experience..

By a detailed procedure, junior and senior contracted radiological control technicians were required to be qualified in a set of license practices and procedures, and seniors were required to be qualified in an additional set of procedures.

In addition to using a detailed procedure for selection and training, the licensee administered and documented these activities. in a detailed and thorough manner.

C.

Conclusions The selection and qualification of contracted radiological control technicians was proceduralized, conducted, administered, and documented in a detailed and thorough manner.

Quality Assurance in RP&C Activities a. 'ns ection Sco e 83750-02 A selective review of the licensee's audits, surveillances, corporate assessments, and self-assessments was performed.

Information was gathered through discussions with cognizant personnel and review and evaluation of procedures and document I

~

)

b.

Observations and Findin s In the radiological control area, the combination of audits, surveillances, corporate assessments, self-assessments, and the problem identification process resulted in a high volume of deficiencies and improvement opportunities being identified and in a low threshold for such identification. The inspector reviewed the most recent audit

.

(1) and surveillance reports (5) since the last NRC inspection in this area.

These were performed by Nuclear Assessment Services.

The audit was programmatic and resulted in numerous findings and recommendations.

The surveillances were focused in scope and highly detailed and also resulted in findings. Three corporate asse'ssments by corporate'adiological Services were reviewed.

Independent

,outside consultants were used in addition to licensee personnel.

They provided a detailed assessment of contamination control, records, and radiation work permits

.

and provided numerous findings and recommendations.

The last two quarterly self-assessments by the Health Physics Group concentrated on radiological controls in the field and identified numerous deficiencies.

Documentation demonstrated that corrective actions were being identified and tracked for completion.

Additionally, a review of the listing of Condition Reports, for 1998 up to April 20; 1998 which were related to radiological controls, included a number of low threshold items among the total of approximately fifty-eight Condition Reports.

c.

Conclusions In the radiological control area, the combination of audits, surveillances, corporate assessments, self-assessments, and the problem identification process resulted in a high volume of deficiencies and improvement opportunities being identified and in a low threshold for such identification.

R8 Miscellaneous RP&C Issues (83750)

A recent discovery of a licensee operating their facility in a manner contrary to the FSAR description highlighted the need for a special focused review that compares plant practices, procedures, and/or parameters to the FSAR descriptions.

While performing the inspections discussed in this report, the applicable portions of the FSAR that related to the areas inspected were reviewed.

It was verified that the FSAR wording was consistent with observed plant practices, procedures, and/or parameter F8 Miscellaneous Fire Protection Issues F8.1 Followu of 0 en Items (92903,92904)

Closed Violation 50-387 94-16-01 Simplex Fire Protection System Failures This violation addressed the failure to establish continuous fire watches within one hour when the fire detection and suppression systems were inoperable.

An inspection of this violation, see NRC Inspection Report (IR) 50-387,388/95-18, concluded that the licensee had adequately resolved the requirements to establish continuous fire watches within one hour when the fire detection and suppression systems are inoperable.

However, the licensee had not completed all of the comprehensive corrective actions to prevent recurrence of the Simplex Fire System failure, as determined by the event review team (ERT). The licensee committed, in IR 50-387, 388/95-18, to send NRC, Region I, a letter presenting the necessary elements for closure of Violation 50-387/94-16-01,following completion of all'ctions.

Specifically, those actions included work items associated with Design Change Package (DCP) No. 95-9003 for removing the installed spare transponder cards, installation of additional surge suppressors, and resolution of ongoing evaluations pertaining to baud rate changes and the associated system effects.

The licensee submitted a letter to the NRC entitled "Completion of Actions for Closure of Violation 50-378/94-16-01; 50-388/94-16-01, Simplex Fire System Failure" dated April 8, 1997. The inspector interviewed the Fire Protection Senior Engineer and reviewed the associated DCP work packages for Surge Suppression Installation (S-47173), Baud Rate Modification (S-56188), and Spare Transponder Card Removal (S-56188).

Based on the in-office letter review, review of the associated DCPs, interviews with the Fire Protection Senior Engineer, and the previous inspection observations, the inspector concluded that, the licensee's corrective actions were appropriate.

V. Mana ement Meetin s X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at an exit meeting on April 27, 1998. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.

No proprietary information was identifie g

~

IP 37550 IP 37551 IP 61726 IP 62707 IP 71707 IP 83750 IP 92700 IP 92901 IP 92902 IP 92903 IP 92904 ATTACHMENT1 INSPECTION PROCEDURES USED Engineering Onsite Engineering Observations Surveillance Observations Maintenance Observations Plan't Operations Occupational Radiation Exposure On Site Followup of Reports Followup Plant Operations Followup Maintenance Followup Engineering Followup Plant Support Qgened None.

ITEMS OPENED, CLOSED, AND DISCUSSED 50-387,388/98%1%9 URI Primary Containment Penetration Leak Rate Testing-Notice of Enforcement Discretion (section E8.1)

Closed 50-387/98-02-01 50-387/9842-02 NCV Reactor Core Thermal Power Exceeded 102%

'(section 08.1)

NCV Closed System Integrity Not Verified by Testing (section 08.1)

50-387,388/98-02-03 NCV Electrical Isolation Between Class 1E and non-Class 1E Electrical Circuits (section E1.2)

50-387/98-02-04 50-387/97-04-00 50-387/97%51 NCV Posting of High Radiation Areas (section R1.2)

LER Reactor Core Thermal Power Exceeded 102%

(section 08.1)

LER Closed System Integrity Not Verified by Testing (section 08.1)

Attachment

50-387,388/97-04-03 50-387/94-1 6-01 VIO Failure to Get Prior Approval of a QA Program Change (section 08.2)

VIO Simplex Fire Protection System Failures (section F8.1)

LIST OF ACRONYMS USED ASO ALARA ANSI BWRT CEDE CFR CR CS DCP EDG ESS FSAR HP HPCI HRA IERP IN IR kv LCO LER LLRT mrem

'NCV NOED NPO NRC NRR OD PAD PCO PCR PORC Psld Pslg QA RB RCA RCIC RHR AuxiliarySystems Operator As-Low-As-Reasonably-Achievable A'merican National Standards Institute Backwash Receiver Tank Committed Effective Dose Equivalent Code of Federal Regulations Condition Report Control Structure Design Change Package Emergency Diesel Generator Engineered Safeguard System Final Safety Analysis Report Health Physics High Pressure Coolant Injection High Radiation Area Industry Experience Review Program

[NRC) Information Notice (NRC] Inspection Report Kilovolts Limiting Condition for Operation Licensee Event Report Local Leak Rate Test millirem Non-Cited Violation Notice of Enforcement Discretion Nuclear Plant Operator Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Operability Determination Personnel Alarming Dosimeter Plant Control Operator Personnel Contamination Report Plant Operations Review Committee Pounds per Square Inch Differential Pounds per Square Inch Gauge Quality Assurance Reactor Building Radiologically Controlled Area Reactor Core Isolation Cooling Residual Heat Removal

Attachment

3 RPRC RWCU RWP SER SLC SS SSES TEDE TLD TS URI US VIO WA WBC Radiological Protection and Chemistry Reactor Water Cleanup Radiation Work Permit Safety Evaluation Report Standby Liquid Control System Shift Supervisor Susquehanna Steam Electric Station Total Effective Dose Equivalent Thermoluminescent Dosimeter Technical Specification

[NRC] Unresolved item Unit Supervisor Violation Work Authorization Whole Body Count

',t

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I,-

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