ML20196C337

Insp Repts 50-352/98-08 & 50-353/98-08 on 980901-1017.No Violations Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support.Security & Safeguards Activities Were Conducted in Manner That Protected Public Health

ML20196C337
Person / Time
Site:	Limerick
Issue date:	11/23/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20196C336	List: ML20196C331 ML20196C337
References
50-352-98-08, 50-352-98-8, 50-353-98-08, 50-353-98-8, NUDOCS 9812020035
Download: ML20196C337 (37)

See also: IR 05000352/1998008

Text

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket Nos. 50-352

50-353 ,

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' License ~ Nos. NPF-39

NPF-85

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. Report Nos. 98-08 j

98-08

Licensee: PECO Energy

Correspondence Control Desk l

P.O. Box 195

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Wayne, PA 19087-0195

Facilities: Limerick Generating Station, Units 1 and 2

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Location: Wayne, PA 19087-0195 - i

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Dates: September 1,1998 through October 17,1998

Inspectors: A. Burritt, Senior Resident inspector

F. Bonnett, Resident inspector

S. Hansell, Resident inspector

S. Barr, Resident inspector

B. Welling, Peach Bottom Resident inspector

J. Noggle, DRS, Sr. Radiation Specialist

G. Smith, DRS, Sr. Physical Security Specialist  !

S. Dennis, DRS, Operations Engineer

Approved by: Clifford Anderson, Chief

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Projects Branch 4

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Division of Reactor Projects

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~9812O20035 981123 I

PDR ADOCK 05000352 }

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EXECUTIVE SUMMARY

Limerick Generating Station, Units 1 & 2

NRC Inspection Report 50-352/98-08,50-353/98-08

This integrated inspection included aspects of PECO Energy operations, engineering,

maintenance, and plant support. The report covers a 7-week period of resident inspection

and region-based inspection in the security, radwaste transportation, and Senior Reactor

Operator Limited to Fuel Handling (LSRO) requalification program areas.

Ooerations

e in structure, the LSRO program was good overall. The program guidelines and

examinations were comprehensive and well maintained by the program coordinator

and LSRO license maintenance was well documented. The inspector also

determined that the areas of exam security, remediation, operator feedback, and

medical records were acceptable. (Section 05.1 )

e LER 50-353/2-98-003 described a condition prohibited by Technical Specifications

in that the main condenser offgas pre-treatment radiation monitor was inoperable

and would not have alarmed as required during a high radiation condition due to an

procedural deficiency. This licensee identified issue is being treated as a Non-Cited

Violation. (Section 08.2)

Maintenance

e The expert panel performed its assigned function well and ensured the consistent

implementation of the maintenance rule in accordance with the program

requirements. (Section M1.3)

e Operator recognition and response for the Unit 2 transformer failure was excellent

resulting in minimalimpact and the timely restoration of the plant to a normal

condition. The transformer replacement, testing and restoration were well

coordinated and performed without error. (Section M2.1)

e Station personnelimplemented the preventive maintenance program consistent with

administrative procedures. Safety related preventive maintenance tasks were

typically performed at the frequencies established by the program guidelines.

Although one UFSAR discrepancy was identified, the licensee was already aware of

and in the process of resolving the inconsistency. (Section M3.1)

e LER 50-353/2-98-006 described a condition prohibited by Technical Specifications

involving the failure to perform an emergency bus undervoltage channel calibration

within period specified. The duc date for this monthly surveillance test was missed

primarily as a result of a personnel error involving l&C's failure to notify the control

room staff that the end of the grace period for this test was approaching. This

licensee identified issue is being treated as a Non-Cited Violation (Section M8.1)

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Executive Summary (cont'd)

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l Enaineerina '

e Engineering personnel took prompt and effective corrective actions following their

identification of a potential suppression chamber bypass path between the drywell

and suppression pool air spaces due to postulated cable failures. This issue was an

apparent violation of 10 CFR 50 Appendix B, Criterion til, " Design Control."

However, in accordance with the NRC Enforcement Policy, Section Vil.B.3,

Violations involving Old Design issues, the NRC is exercising enforcement discretion

and not citing this violation. (Section E2.1)

e PECO personnel responded well to quickly detect and suppress a fuelleak at Unit 1. l

The multi-disciplined fuel monitoring task force developed a strategy to locate and

suppress the fuelleak prior to the initiation of further failure. (Section E2.2)

• LER 50-352; 353/1-98-013 described the failure to meet the requirements for

maximum travel distance limitation for portable fire extinguishers. The discrepancy

was a result of PECO and Bechtel not adhering to the National Fire Protection

Association 101975 code when the fire extinguishers were distributed during plant

construction. Additionally, subsequent audits of the fire protection program had

failed to identify the discrepancy. This licensee identified issue is being treated as a

Non-Cited Violation. (Section E8.4)

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• LER 50-352; 353/1-98-015 described a condition prohibited by Technical

Specifications due to an error in calibration of core spray line break differential

pressure instruments. The error was a result of a faulty assumption in the setpoint

calculation which did not account for the differential pressure between the two

trains under normal conditions. This licensee identified issue is being treated as a

Non-Cited Violation. (Section E8.5)

e LER 50-352; 353/1-98-017 described a condition involving the inability of hatchway

fire protection flow control valves to remain open when actuated. Additionally, the

surveillance testing of these valves and previous corrective actions for other fire

protection flow conteol valves were inadequate. This licensee identified violation of

10 CFR 50 Appendix B, Criterion lil, Design Control is being treated as a Non-Cited

Violation. (Section E8.6)

• LER 50-353/2-98-C01 described a condition prohibited by Technical Specifications

involving three inoperable Barksdale model C9622-3-B differential pressure

switches. This result in two independent trains of a single safety system being

inoperable from a common cause. This event occurred as a result of inadequate

margins to account for setpoint drift over a 24-month fuel cycle. This licensee

identified issue is being treated as a Non-Cited Violation. (Section E8.7)

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Executive Summary (cont'd)

Plant Suncort

• The licensee was conducting security and safeguards activities in a manner that

protected public health and safety in the areas of alarm stations, communications,

protected area access control of personnel and packages. This portion of the

program as implemented, met the licensee's commitments and NRC requirements.

(Section S1)

• The licensee's security facilities and equipment in the areas of protected area

assessment aids, protected area detection aids, and personnel search equipment i

were detera ined to be well maintained and reliable and were able to meet the

licensee's commitments and NRC requirements. (Section S2)

• Security and saniguards procedures and documentation were being properly

implemented. Event Logs were being properly maintained and effectively used to

analyze, track, and resolve safeguards events. (Section S3)

• The security force members adequately demonstrated that they had the requisite

knowledge necessary to effectively implement the duties and responsibilities

associated with their position. (Section S4)

• Limerick solid radioactive wastes were effectively sampled, packaged, and

dewatered with respect to requirements. The radwaste staff is pursuing an

enhancement to the program to more accurately quantify the condensate filtrate

waste volumes. (Section R1)

• Radioactive material shipments were prepared in an expeditious manner and met all

regulatory requirements. Shipping records were properly prepared with no

deficiencies identified. (Section R1)

• The licensee has effectively minimized the amount of contaminated equipment and

radioactive wastes stored onsite. (Section R1)

• Monitoring of material exiting the radiological controlled area was not always

conducted at the low sensitivities specified by station procedure. (Section R1)

• Limerick radioactive waste processing and radioactive material shipping procedures

were of good quality and effectively implemented regulatory requirements. (Section

R3)

• All authorized radioactive material shipment personnel have met the applicable DOT

and NRC training requirements. (Section RS)

• Quality assurance oversight of the radioactive material shipment program was

effective through performance of an independent program assessment and

surveillances and through radwaste staff shipment verifications. (Section R7)

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TABLE OF CONTENTS

EX ECUT (V E SU M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii

TAB LE O F C O NT ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v

Summary of Plant Status ............................................1

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1. O p e ra tio n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 '

01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . ............ 1

01.1 General Comments (71707) ...........................1

04 . Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 2

04.1 Control RM Mispositioning . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

05 Operator Trairdng and Qualification ...........................2

05.1 Limited Senior Reactor Operator (LSRO) Requalification Program . . 2 -

08 Miscellaneous Operations issues (92700,92702) . . . . . . . . . . . . . . . . . 4

08.1 (Closed) LER 50-352; 353/1-98-016: Manual MCR Ventilation

isolation and CREFAS Initiation due to Small Freon Leak ....... 4  :

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08.2 (Closed) LER 50-353/2-98-003: Condition Prohibited by Technical

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Specification in that the Main Condenser Offgas Pre-treatment

Radiation Monitor was inoperable and the Action was not met due

to an incorrect Procedure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

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ll . M ainte n anc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 j

M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

M1.1 General Comments on Maintenance Activities (62707) ........ 5

M1.2 General Comments on Surveillance Activities (61726) . . . . . . . . . 6

M1.3 Maintenance Rule Program Observations ..................6

M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . 7

M2.1 Load Center Transformer Failure . . . . . . . . . . . . . . . . . . . . . . . . 7

M3 Maintenance Procedures and Documentation ....................8

M3.1 Preventive Maintenance Program Review . . . . . . . . . . . . . . . . . . 8

M8 Miscellaneous Maintenance issues (92902) .....................9

- M8.1 (Closed) LER 50-353/2-98-006: Failure to Meet Undervoltage Channel

Calibration Technical Specification Surveillance Requirement .... 9

111. E ng i n e e ri ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10

E2 Engineering Support of Facilities and Equipment ................. 10

E2.1 (Closed) URI 50-352;353/98-05-05and (Closed) LER 50-352;

353/97-010: Potential Containment Bypass Path Resulting in a

Condition Outside the Design Basis . . . . . . . . . . . . . . . . . . . . . 10

E2.2 Fuel Failure at Unit 1 ...............................11

E8 Miscellaneous Engineering issues (92903,92700) . . . . . . . . . . . . . . . . 13

E8.1 (Closed) LER 50-353/2-98-004: Secondary Containment isolation,

Standby Gas Treatment System (SGTS) and Reactor Enclosure

Recirculation System (RERS) Initiation ...................13

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Table of Contents (cont'd)

E8.2 (Closed) IFl 50-352/97-07-02: Reactor Water Cleanup (RWCU)

Isolations and LER 50-352; 353/1-98-014:ESF Actuation Due to

RWCU System Isolations ............................13

E8.3 (Closed) VIO 50-352; 353/98-04-04: Failure to Submit Licensee

Eve nt R e port . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 l

E8.4 (Closed) LER 50-352;353/1-98-013: Failure to Meet the Maximum

Travel Distance Limitation for Portable Fire Extinguishers . . . . . . 14

E8.5 (Closed) LER 50-352; 353/1-98-015: Condition Prohibited by i

Technical Specifications Due to an Error in Calibration of Core Spray I

Line Break Differential Pressure Instruments. .............. 15 l

E8.6 (Closed) LER 50-352; 353/1-98-017: Failure of Hatchway Fire l

Protection Flow Control Valves to Actuate . . . . . . . . . . . . . . . . 15

E8.7. (Closed) LER 50-353/2-98-001: Three Inoperable Barksdale Model

C9622-3-B Differential Pressure Switches Resuit in Two or More ,

independent Trains of a Single Safety System Being inoperable j

From a Common Cause . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 16 i

IV. Pl a nt S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 7

R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 17

R1.1 Solid Radwaste Processing . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

R1.2 Radioactive Material Shipping . . . . . . . . . . . . . . . . . . . . . . . . . 18

R1.3 Solid Radioactive Waste Storage . . . . . . . . . . . . . . . . . . . . . . . 18

R1.4 Radiological Controlled Area Material Monitoring . . . . . . . . . . . . 19

R3 RP&C Procedures and Documentation ........................19  ;

R3.1 Radioactive Material Shipment Procedures . . . . . . . . . . . . . . . . 19 )

R5 Staff Training and Qualification in RP&C . . . . . . . . . . . . . . . . . . . . . . . 20 '

R5.1 Radioactive Material Shipment Training ..................20

R7 Quality Assurance in Radiological Protection and Chemistry Activities . . 21

R7.1 Radioactive Material Shipping QA Oversight . . . . . . . . . . . . . . . 21 J

S1 Conduct of Security and Safeguards Activities ..................21  ;

S2 Status of Security Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . 23 i

S3 Security and Safeguards Procedures and Documentation . . . . . . . . . . . 23

S4 Security and Safeguards Staff Knowledge and Performance . . . . . . . . . 24

SS Security and Safeguards Staff Training and Qualification . . . . . . . . . . . 24

S6 Security Organization and Administration . . . . . . . . . . . . . . . . . . ... 25

V. M anag eme nt Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5

X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5

ATTACHMENT

l Attachment 1 - Inspection Procedures Used

- Partial List of Persons Contacted

-Items Opened, Closed, and Discussed

- List of Acronyms Used

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Report Details

Summary of Plant Status

Unit 1 began this inspection period operating at 100% power. The unit remained at full

power throughout the inspection period with minor exceptions for testing, rod pattern

adjustments, and the following plant events,

o September 30 An operator noted a 30 millirem /hr step increase in the offgas

radiation monitor levels, indicative of a potential fuel leak.

o October 7 Operators reduced reactor power to 60% per GP-5, Power

Operations, to establish conditions to perform power

suppression (flux-tilt) testing.

o October 12 Operators commenced increasing reactor power from 60% per

GP-5, after completing power suppression (flux tilt) testing.

Unit 1 reached 100% power on October 13 and remained at

full power for the remainder of the period.

Unit 2 began this inspection period operating at 100% power. The unit remainod at full

power throughout the inspection period with minor exceptions for testing, rod pattern

adjustments, and the fc! lowing plant event.

o October 17 Operators reduced reactor power to 60% per GP-5, to perform

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a deep / shallow control rod exchange, scram time testing, and j

l condenser 2A waterbox cleaning. Power was returned to  !

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100% on October 18. ,

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1. Operations

01 Conduct of Operations 1

01.1 General Comments (7170_7) i

Using inspection Procedure 71707,the inspectors conducted frequent reviews of

l ongoing plant operations. PECO Energy's (PECO) conduct of activities at Limerick

Units 1 and 2 was generally characterized by safe and conservative operations and

decision making. Operators' response to the Unit 2 load center transformer failure

and early detection of the Unit 1 fuel failure were excellent. Management's

proactive response to the fuel failure demonstrated a concerted effort to minimize

the effects of the leak.

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1 ' Topical headings such as o1, M8, etc., are used in accordance with the NRC standardized reactor inspection report

outhne. Indiv' dual reports are not expected to address all outline topics.

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04 Operator Knowledge and Performance

04.1 Control Rod Mispositionina

On October,4,1998, a single control rod on the Unit 2 reactor was inadvertently 1

inserted one notch. The error occurred when a reactor operator (RO) attempted to

reset a control room alarm associated with the rod block monitor (RBM). Instead of

depressing the overhead annunciator reset push-button the RO inadvertently

depressed the single notch insert push-button on the rod control panel. The

annunciator reset and single notch insert push-buttons are located on the same

main control panel and are approximately 15 inches apart. The RO immediately

recognized the error and informed the control room supervisor (CRS). The operators

entered the appropriate off-normal procedure and moved the control rod out one l

notch to the proper position. After the rod insertion, the RO immediately checked

the reactor core thermallimits report. The computer printout verified that

conditions remained normal after the rod movement.

The inspector revicwed the issue and interviewed the control personnel. The error

was attributed to less than adequate self-checking by the RO. A performance

enhancement program (PEP) report was written to document the issue and ensure

the implementation of appropriate corrective actions. The corrective actions were

thorough and timely. In addition, the issue was discussed with all operators in

detail to reinforce the importance of proper self checking.

05 Operator Training and Qualification

05.1 Limited Senior Reactor Ooerator (LSRO) Reaualification Proaram

a. Insoection Scope (71001)

The inspector evaluated the dual site, Limerick / Peach Bottom, PECO Nuclear

(PECON) LSRO requalification training program to verify it's compliance with

10 CFR 55 requirements. The inspector used NRC inspection Procedure 71001,

Licensed Operator Requalification Program Evaluation, and NUREG-1021 Interim

Rev.8 - ES-702 for the evaluation.

The inspector evaluated the following program areas:

• Program guidelines

• Operating and written examinations

• Exam security

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• Management oversight -license activation and maintenance of records,

remediation, training, attendance, feedback system, and medical records

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PECON procedures and documents associated with the LSRO training program and

its implementation were also reviewed.

Since the annual operating exam was not administered during this inspection period,

no insigi'ts could be obtained on operator performance,

b. Observations and Findinas

Proaram Guidelines

The inspector determined that PECON procedures LSRO-9500,"LSRO Course Plan",

and LSRO-0000," Multi-Site Fuel Handling Director", acceptably described a

program which met 10 CFR 55 requirements and previous written commitments by

PECON to the NRC. Additionally, the inspector found the content of the LSRO

program subject index and selected LSRO classroom and practical job performance

lesson plans to be comprehensive and well maintained by the program coordinator.

Operatina and Written Examinations

The inspector determined that three written biennial examinations and two annual

operating exams acceptably sampled the items specified in 10 CFR 55. The

inspector also found that the exams adequately assessed knowledge level in the

area of abnormal and emergency procedures. Additionally, it was noted that a large

percentage of the questions in the exams were of the more challenging, higher

order, analytical type.

The inspector determined that job performance measures (JPMs) met the qualitative

guidelines of the inspection procedure and the PECON program. The JPMs reviewed

included those for normal, emergency, and abnormal conditions.

Exam Security

The inspector determined that the security measures and programmatic controls

taken by the facility for exam development and administration were satisf actory,

with no indications of exam compromise.

Activation and Maintenance of Operator Licenses

The inspector found acceptable PECONs programmatic controls for maintaining

an active license and for reactivating a license while meeting the requirements of

10 CFR 55.53. The inspector reviewed various training attendance records,

including missed training make-up sessions or exams, and determined that controls

for maintenance and reactivation of operator licenses were good.

Remedial Trainin Proaram

The inspector found that the remediation records for two individuals who had failed

the biennial written exams were good. The remediation p: % iges developed by the

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training coordinator were appropriate for the weaknesses demonstrated and were l

properly documented in accordance with PECON procedures.  !

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Ooerator Feedback

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The inspector found that management's review and disposition of feedback records j

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for the past three years was timely.

Medical Records

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The inspector also reviewed all LSRO medical files to ensure that medical exams

were being conducted biennially in accordance with 10 CFR 55.21 and determined

that requirements were met. )

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c. Conclusions '

In structure, the LSRO program was good overall. The program guidelines and

examinations were comprehensive and well maintained by the program coordinator

and LSRO license maintenance was well documented. The inspector also

determined that the areas of exam security, remediation, operator feedback, and

medical records were acceptable.

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08 Miscellaneous Operations issues (92700,92702)

08.1 (Closed) LER 50-352: 353/1-98-016: Manual MCR VrJ!ation Isolation and CREFAS

Initiation due to Small Freon Leak

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On July 28,1998, technicians identified'a small Freon leak on the Unit 1 A drywell

chiller unit. The CRS directed the operators to manually initiate a main control room

(MCR) chlorine mode isolation in anticipation of a possible toxic gas analyzer alarm

in response to the Freon. Plant procedures required additional operator actions in

response to the alarm if MCR ventilation was not isolated. The control room i

emergency fresh air system (CREFAS) initiated as designed. PECO stated the cause

of the manualisolation and CREFAS initiation to be the CRS's conscious,

conservative decision to manually control the event with the least impact on plant

operation. The Freon leak resulted from lack of preventive maintenance on the

chiller motor lead packing gland. The Freon leak was repaired. Planned corrective

actions include inspecting the other plant chillers at both units for proper packing

gland torque values, establishing a PM task, and evaluating manufacturer

information for possible chiller modifications and /or work practice changes. The

inspector determined, during the in-office review of the LER and PEP 10008741,that

PECO's actions were appropriate and there was no violation of NRC requirements.

This LER is closed.

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08.2 (Closed) LER 50-353/2-98-003: Condition Prohibited by Technical Specification in

l that the Main Condenser Offacs Pre-treatment Radiation Monitor was Inocerable

! and the Action was not met due to an incorrect Procedure.

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! This LER documented an event that occurred on June 22,1998, where a system

l manager discovered the flow-rate through the main condenser offgas pre-treatment

i radiation monitor was inadvertently throttled to a flow-rate lower than required.

This resulted in a condition where the radiation monitor would not have alarmed

l_ during a high radiation condition in the off gas system at the required setpoint of

l 1.5 times normal full power background. This was a condition prohibited by TS 3.3.7.12.

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PECO performed an adequate review of the event which is documented in the LER

and PEP 10008589. PECO attributed the primary cause of the event to be an

l incorrect system operating procedure and implemented corrective actions involving:

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1) procedure revisions to system procedure S26.1.G, Placing the Air Ejector /Offgas

l Monitor in Service, and to chemistry procedure, CH1005A, Sampling and Analysis ,

of Offgas from Recombiner Aftercondenser Discharge; 2) management emphasizing l

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performance expectations; and 3) verifying that the similar valves were l

appropriately aligned at other radiation monitor skids. The inspector reviewed, in i

office, the PEP and procedure revisions and discussed the corrective actions with j

a radiological technician. The radiation monitor being inoperable for seven days is  !

a violation of Technical Specifications. This licensee-identified, non-repetitive and

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corrected violation is being treated as a Non-Cited Violation consistent with

l Section Vll.B.1 of the NRC Enforcement Poliev. (NCV 50-352; 353/98-08-01) This l

LER is closed.

11. Maintenance

M1 Conduct of Maintenance

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M1.1 General Comments on Maintenance Activities (62707)

The inspectors observed selected maintenance activities to determine whether

approved procedures were in use, details were adequate, technical specifications

were satisfied, maintenance was performed by knowledgeable personnel, and post-

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maintenance testing was appropriately completed.

The inspectors observed portions of the following work activities:

! * Unit 2 - D-24 Diesel Generator Auxiliary Lube Oil Pump Seal Replacement,

I September 17;

e Unit 1 - HPCI Pump Discharge (1-HV-F007) MOV Replacement,

September 22-24;

* Unit 1 - 1 A2125 VDC Safeguard Battery Cell Replacement,

j September 17-18,27-29;

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Observed maintenance activities were conducted well using approved procedures,

and were completed with satisfactory results. Communications between the j

various work and support groups were good, and supervisor oversight was good.

M1.2 General Comments on Surveillance Activities (61726)

The inspectors observed selected surveillance tests to determine whether approved

procedures were in use, details were adequate, test instrumentation was properly

calibrated and used, technical specifications were satisfied, testing was performed

by knowledgeable personnel, and test results satisfied acceptance criteria or were

properly dispositioned.

The inspectors observed portions of the following surveillance activities:  ;

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e Unit 1 - ST-6-092-31-1,"D-11 Diesel Generator Monthly Slow Start Test,"

- September 1;

e Unit 2 - RT-6-092-312-2,"D-22 Diesel Generator Run-in Test,"

- September 8;

e Unit 2 - ST-6-049-230-2,"RCIC Pump and Turbine Performance Data Test,"

- September 10;

e Unit 1 - ST-6-092-3141,"D-14 Diesel Generator Monthly Slow Start Test,"

- September 29;

e Unit 2 - ST-6-071-307-2," Channel B1 and B2 RPS Manual Scam Channel

Functional Test," - September 29;

e Unit 1 - ST-6-051-233-1,"C RHR Pump, Valve & Flow Tests," -

September 17;

e Unit 1 - ST-6-092-314-2,"D-24 Diesel Generator Monthly Slow Start Test,"

September 16;

e Unit 1 - S74.0.A, " Operation of Transversing in-Core Probe System," -

September 16;

e Unit 1 - ST-6-076-250-1,"SGTS and RERS Flow Test,"- October 15;

• Unit 1 - ST-6-076-200-1," Reactor Enclosure Secondary Containment Auto

Isolation Valve Timing Test,"- October 15

Observed surveillance tests were conducted well using approved procedures, and

were completed with satisfactory results. Communications between the various

work and support groups were good, and supervisor oversight veas good.

M1.3 Maintenance Rule Proaram Observations

a. Inspection Scoce (61726)

The inspectors reviewed PECO procedure AG-CG-28.1, " Maintenance Rule

implementation Program," which detailed the responsibilities of the expert panel.

The inspectors attended the expert panel meeting on September 24,1998.

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b. Observations __are d Findinas

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The expert panel consisted of members with experience in plant operations,

maintenance, engineering, and probabilistic risk assessment. The expert panel

reviewed and concurred with the status of (a)(1) systems, the addition of safety  !

related coatings to the maintenance rule program, the decision for the Unit 1

electrohydraulic control (EHC) system to remain in category (a)(2), the revised

action plans for the standby gas treatment (SGTS) and reactor enclosure

recirculation systems (RERS), and evaluation of recent equipment functional failures

and maintenance preventable functional failures.

All panel members discussed each topic in depth. The panel conclusions were

supported by well researched information and written documentation. The panel  !

provided enhancements to the non-safety related reactor enclosure ventilation and j

the SGTS/RERSimprovement plans. The changes included specific time frames to

l determine when improvement goals should be achieved and also ensured

maintenance rule program consistency.  ;

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c. Conclusions

The expert panel performed its assigned function well and ensured the consistent

implementation of the maintenance rule in accordance with the program

requirements.

l M2 Maintenance and Material Condition of Facilities and Equipment

M2.1 Load Center Transformer Failure

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a. Inspection Scoce (62707)

The inspectors reviewed the Unit 2 load center (LC) 224B transformer failure,

operator response to the event, and subsequent transformer replacement.

b. Observations and Findinas

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On September 24,1998, the Unit 2 load center LC-2248 electrical supply breaker

tripped open. At the time instrumentation and control (l&C) technicians were

l recording temperature measurements (thermography) of the 480 Volt load center.

The breaker tripped when a technician attempted to close the LC transformer door.

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Simultaneously, a number of alarms annunciated in the Unit 2 control room due to

the electrical power interruption. Control room operators recognized immediately

that the LC normal supply breaker had tripped open. The event resulted in the loss

of power to both reactor water cleanup pumps, the operating drywell chiller, a loss

of cooling to the reactor recirculation pumps, and both recirculation pump scoop

tubes locked.

As a result, drywell(DW) primary containment pressure and temperature began to

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increase. Operators entered the appropriate abnormal procedures and started the

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backup DW chiller to restore normal cooling to the DW. DW pressure increased to

approximately 0.6 psig, well below the trip setpoint of 1.68 psig, before cooling

was restored. The inspector observed good control room recognition and response

to the LC power loss. The excellent operator response resulted in minimalimpact s

and timely restoration of the plant to a normal condition. In addition, good operator l

procedure adherence, proper supervisory oversight and conservative decision

making were noted.

The initial investigation of the transformer indicated that a wire came in contact

with the B phase transformer coil and shorted it to ground while the LC door was

being closed. The wire was the cable which connects the B phase coil to a

temperature indication on the LC door. No one was injured as a result of the event. 1

As a safety precaution, all thermography work on electrical LCs was stopped and '

the LC doors were tagged closed for both Units until the problem is resolved. After

inspection of the 480 Volt side of the LC for damage, equipment power by the LC

was transferred to a backup electrical power supply until the damaged transformer

was replaced.

PECO electricians replaced and tested the transformer within a week as a result of a

well coordinated effort to remove the damaged transformer and install the new

replacement. The work was planned, scheduled, and performed without a problem

in a minimal amount of time. After testing, the LC power supply was returned to

the normal alignment by plant operators.

c. Conclusions

Operator recognition and response for the Unit 2 transformer failure was excellent

resulting in minimalimpact and the' timely restoration of the plant to a normal

condition. The transformer replacement, testing, and restoration were well

coordinated and performed without error.

M3 Maintenance Procedures and Documentation i

M3.1 Preventive Maintenance Proaram Review

a. Inspection Scoce (62707)

The inspectors reviewed selected aspects of the implementation of the preventive

maintenance (PM) program as described in administrative procedure A-C-28,

" Preventive Maintenance Program." The inspectors also examined the scheduled

frequencies of several safety related PM tasks and compared them to vendor

recommendations and licensing commitments in the Updated Final Safety Analysis

Report (UFSAR) and Licensee Event Reports (LERs),

b. Observations and Findinas

PM tasks were scheduled consistent with the established frequencies. The tasks

were usually performed by the assigned due dates, although some PMs were

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allowed to be completed in the " grace period," which was defined similar to that

used for surveillance testing. The PM coordinator was actively managing the

number of PMs in the grace period and had reduced this number over the past I

several months. Few PM tasks had exceeded the grace period.

The PM frequencies were typically determined by system managers according to the

PM program guidance. Changes to the frequencies were usually evaluated by i

engineering personnel, j

Exceptions to the specified/ committed PM frequencies were identified for the high  !

pressure coolant injection (HPCl) system. The inspectors noted that the UFSAR,

Section 6.3, stated that periodic inspections and maintenance of the system are

conducted in accordance with manufacturers' instructions. The HPCI Turbine

Vendor Manual, E41-C002-K001, specified one-year and five-year intervals for HPCI

minor and major maintenance inspections, respectively. However, the inspections

were actually being performed at two and eight-year intervals. The inspectors I

discussed this discrepancy with engineering personnel and learned that these )

intervals were based on an industry maintenance and troubleshooting guide. 1

Engineering personnel also stated that they had recently identified the inconsistency

between the UFSAR statement and the specified intervals. The system manager

documented that the UFSAR will be revised, through the engineering change

request process, to indicate that the inspections will be based on industry /

manufacturers' guidelines. The inspectors identified no concerns with this

approach.

c. Conclusions

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Station personnel implemented the preventive maintenance program consistent with

administrative procedures. Safety related preventive maintenance tasks were

typically performed at the frequencies established by the program guidelines.

Although one UFSAR discrepancy was identified, the licensee was already aware of

and in the process of resolving the inconsistency.  !

M8 Miscellaneous Maintenance issues (92902)

M8.1 (Closed) LER 50-353/2-98-006: Failure to Meet Undervoltaae Channel Calibration

Technical Soecification Surveillance Reauirement

The LER documents an event that occurred on July 27,1998, where an l&C

technician discovered that the monthly surveillance test ST-2-092-324-2,D-24 4kV

Emergency Bus Undervoltage Channel Calibration / Functional Test, exceeded its due

date. The failure to perform the surveillance test prior to the due date resulted in a

non-compliance with TS 4.0.2 and Table 4.3.3.1-1 Item 5b. The overhaul of the I

associated diesel generator was in progress the week the test was scheduled to be

performed and the l&C manager failed to notify the control room staff that the end

of its grace period for the surveillance test was July 26.

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PECO attributed the primary cause of the event to be personnel error. Corrective

actions implemented included l&C management reinforcing with l&C supervisors

their accountability for the surveillance test program and the briefing of all I&C staff

personnel to reinforce the need to notify supervision if a surveillance cannot be

performed. Lastly, approaching overdue surveillance tests are discussed at the

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afternoon Work Coordination meeting. The inspector reviewed, in office, the

circumstances of this event and the licensee's analysis of and response to it. The

inspector also observed discussions during the afternoon Work Coordination

meeting. This licensee-identified, non-repetitive and corrected violation is being

treated as a Non-Cited Violation consistent with Section Vll.B.1 of the NRC

Enforcement Poliev." (NCV 50-352:353/98-08-02) This LER is closed.

Ill. Engineering

E2 Engineering Support of Facilities and Equipment

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E2.1 (Closed) URI 50-352:353/98-05-05and (Closed) LER 50-352:353/97-010:Fotential

Containment Bvoass Path Resultina in a Condition Outside the Desian Basis

a. Inspection Scope (92903)

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The inspectors concluded a review of licensee actions taken in response to the

identification of a potential suppression chamber bypass path between the drywell

and suppression pool air spaces.

b. Observations and Findinas

NRC Inspection report 50-352:353/98-05 documented a review of PECO's interim

corrective actions for a potential containment bypass condition through six-inch

containment purge nitrogen supply piping. The inspectors noted that PECO

personnel had discovered that " hot shorts" or a control cabinet f ailure could

potentially cause both the drywell and suppression pool inboard nitrogen supply

isolation valves to open, interconnecting both areas. If this condition occurred

during a loss of coolant accident (LOCA), the design pressure of the containment

could be exceeded. The inspectors concluded that the interim corrective actions,

which included disabling one of the two isolation valves and revising procedures,

were acceptable. This item was left unresolved pending NRC review of PECO's

event evaluation and determination of perme unt resolution of the issue.

The inspectors reviewed non-conformance reports, a PEP report, and other

engineering documentation for this issue. The inspectors also conducted

discussions with engineering personnel in order to determine the causes,

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evaluations, and proposed final resolution.

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Engineering personnel attributed the cause to an original design deficiency, in that

the design requirements for lines which connect the drywell airspace to the

suppression pool airspace were not adequately specified. Single failure and

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electrical independence design criteria were not originally applied to the drywell and

suppression pool inboard nitrogen supply valves.

Engineering evaluations of postulated LOCA events with the bypass condition

indicated that, under some scenarios without operator action, the design pressure of

the containment would be exceeded. Engineering also noted that operator actions

to initiate suppression pool spray would mitigate the pressure increase under small-

break LOCA conditions. An evaluation of other possible bypass leakage paths was

completed in October 1998, and identified no additional credible paths. Engineering

personnel concluded that a modification was necessary to provide a permanent

resolution. Analyses of various modification alternatives were in-progress, with a

final determination planned for December 1998. The inspectors concluded that

engineering had made adequate progress on evaluating and permanently resolving

the issue.

The inspectors determined that this issue was an apparent violation of 10 CFR 50

Appendix B, Criterion 111, " Design Control." However, the inspectors noted that it

was licensee identified as a result of reviews of industry operating experience and

General Electric 10 CFR Part 21 notification No. SC97-04 dated October 15,1997. l

In addition, the inspectors concluded that station personnel took prompt and i

effective interim corrective actions, and this issue was not likely to be identified l

through routine efforts, in accordance with the NRC Enforcement Policy,

Section Vll.B.3, Violations involving Old Design issues, the NRC is exercising

enforcement discretion and not citing this violation as noted in a separate

correspondence issued on November 23,1998. (NCV 50-352: 353/98-08-03)

c. Conclusions j

Engineering personnel took prompt and effective corrective actions following their

identification of a potential suppression chamber bypass path between the drywell

and suppression pool air spaces due to postulated cable failures. This issue was an

apparent violation of 10 CFR 50 Appendix B, Criterion lil, " Design Control."

However, in accordance with the NRC Enforcement Policy, Section Vll.B.3,

Violations involving Old Design issues, the NRC is exercising enforcement discretion

and not citing this violation.

E2.2 Fuel Failure at Unit 1

a. Inspection Scoce (37551)  ;

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On September 30,1998, PECO personnel detected a fuel leak on Unit 1. The l

inspectors attended several fuel monitoring task force meetings, observed portions

of the power suppression testing, and discussed PECO's corrective actions with

various members of PECO management.

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b. Observations and Findinas

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l A reactor operator identified that the main condenser radiation monitor had spiked

l up about 20 mrem /hr and then remained constant. The control room staff

i implemented actions as per off-normal procedure ON-102, " Air Ejector Discharge

High Radiation", and general procedure GP-5, " Normal Operations.

l Chemistry initially confirmed an activity increase from 1800pci/sec to 3100 ci/sec

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in the steam jet air ejector discharge to off-gas system. On-going chemistry

samples confirmed the source of activity was a fuelleak. Chemistry results

l indicated a steady increase in Neptunium-239, Strontium-92, lodine-131, " Sum of

l Six" (Krypton-85, 86, 87 and Xenon-133,135, and 138), and other isotopes

characteristic of a fuelleak.

A multi-disciplined fuel monitoring task force (FMTF) was formed to provide a

comprehensive evaluation of the failure. The FMTF developed recommendations for

l continued plant operation in accordance with the failed fuel action plan detailed in

l section 7.3 of procedure FM-C-3, " Fuel Reliability." The FMTF reviewed the uriit's

l power history prior to the event, contacted the fuel vendor, obtained industry

support, and planned a strategy to suppress the leak as per procedure RE-C-30,

" Fuel integrity Monitoring and Response."

PECO conducted flux-tilt testing of all 185 control rods between October 8 and 11,

l to determine the location and magnitude of the leak. The leak was located in a

I second cycle fuel bundle in control cell 41-40 and was estimated from the data

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characteristics to be about eight inches long. Five control rods were fully inserted

to suppress local power in the vicinity of the leak. Reducing local power minimizes

fission products released int'o the coolant. As a result of the power suppression

I chemistry levels have remained relatively constant with only a slight increase in

I activity.

PECO intends to remove the leaking fuel bundle during a planned outage starting

December 4,1998. PECO willinspect the fuel for indications of possible failure

mechanisms and implement required corrective measures at that time.

c. Conclusions

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PECO personnel responded well to quickly detect and suppress a fuel leak at Unit 1.

l The multi-disciplined fuel monitoring task force developed a strategy to locate and

suppress the fuel leak prior to the initiation of further failure.

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E8 Miscellaneous Engineering issues (92903,92700)

E8.1 (Closed) LER 50-353/2-98-OO4:Secondarv Containment isolation. Standbv Gas

Treatment System (SGTS) and Reactor Enclosure Recirculation System (RERS)

Initiation

On June 26,1998, manual actions were taken by plant operators to perform a

secondary containment isStation in conjunction with a SGTS and RERS initiation.

The cause of the event was the inability of the normal reactor enclosure (RE)

ventilation system to maintain a negative pressure in the secondary containment

during severe weather conditions. Corrective actions included: 1) the immediate

initiation of the SGTS and RERS systems to restore RE pressure to normal; 2) an

evaluation of the RE ventilatic , system flow balance and capabilities; 3) an

enhancement of the system operating procedure guidance; and 4) the SGTS and

RERS systems were added to the maintenance rule (a)(1) category to address the

repetitive equipment problems.

A review of the corrective actions, by the inspector, was performed in the plant.

Control room alarm response procedure, " Reactor Enclosure Low Delta P/ Loss of

Power /INOP," was revised to provide clearer guidance to operators if a positive

pressure occurred in the RE. The system manager and maintenance rule expert

panel have documented the necessary corrective actions to improve system

performance and reduce the number of challenges to plant operators. No violation

of NRC requirements were identified and this LER is closed.

E8.2 (Closed) IFl 50-352/97-07-02: Reactor Water Cleanuo (RWCU) Isolations and LER

50-352: 353/1-98-014:ES_F Actuation Due to RWCU System isolations

, The inspector opened this inspection follow-up issue (IFI) to address maintenance

rule implications and common causes for several Unit 2 RWCU system isolations.

The system experienced several isolations due to high differential flow conditions

while restoring a filter demineralizer to service. The RWCU system was reviewed

during the maintenance rule team inspection, NRC Inspection Report 50-352;

353/98-06. The team concluded that the RWCU system was properly classified

and monitored based on system performance. A system walkdown determined that

the plant equipment conditions were satisfactory. This IFl item is closed.

LER 1-98-014 addressed three similar RWCU system isolation events. The system

manager has implemented hardware and operating procedure changes to improve

the system reliability and reduce the number of operator challenges. Also, the

RWCU pumps will be replaced with a seal-less pump beginning in February 1999.

The inspector conducted an in-field review and determined that the licensee's

corrective actions were appropriate. No violations of NRC requirements associated

with the RWCU isolations were identified and this LER is closed. The late reporting

of this LER was reviewed and documented in NRC Inspection Report 50-352;

353/98-05, section E7.1.

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E8.3 (Closed) VIO 50-352: 353/98-04-04: Failure to Submit Licensee Event Report

in February 1998, the licensee identified 20 safety-related valves that had not been

adequately tested as per TS 4.6.3.2 and this condition was not reported in an LER

within the required time. The inadequate testing of the valves was identified during

a generic implications review of a PEP involving similar testing deficiencies. As

corrective action all PEP investigation review leaders were instructed to notify  ;

station Experience Assessment personnel for reportability determinations when new

issues or concerns were identified during PEP reviews. PECO also corrected

weakness identified in the governing procedure LR-C-10, PEP, which included

adding requirements for initiating a new PEP evaluation when additional problems

are identified. The inspector found these corrective actions to be adequate. This

item is closed.

E8.4 (Closed) LER 50-352:353/1-98-013: Failure to Meet the Maximum Travel Distance

Limitation for Portable Fire Extinauishers.

This LER documented the June 3,1998, determination by PECO's Fire Protection

Group that the distribution of fire extinguishers in the Limerick power block did not

meet the maximum travel distance limitation or the guidance for replacement of

those extinguishers with hose stations as identified in the National Fire Protection

Association (NFPA) 10-1975 code. The failure to meet the NFPA requirements

constituted a failure to maintain the provisions of the Limerick fire protection

program as described in tae UFSAR and was, therefore, a violation of the Limerick

Operating License. PECO deterrrined that this discrepancy was a result of PECO

and Bechtel not adhering to the NFPA code when the fire extinguishers were

distributed during plant construction. Additionally, subsequent audits of the fire

protection program had failed to identify the disc'epancy.

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NRC Generic Letter 86-10, " Implementation of Fire Protection Requirements,"

permitted licensees to deviate from the requirements of the NFPA code, provided

the deviations were evaluated as not adversely affecting the approved fire

protection program. PECO Engineering's evaluation of this discrepancy concluded

that the deviations from NFPA 10-1975 did not reduce the effectiveness of the

Limerick fire protection program and were acceptable. While that evaluation and

conclusion were pending, the licensee had implemented interim corrective actions,

including a shift night order briefing to the operations fire brigade of the situation

and the placement of additional fire extinguishers in the fire brigade locker.

The inspectors conducted an on-site tour of the power block following the initial

discovery of the discrepancy, verified the licensee's determination, and confirmed

the implementation of the interim corrective actions. The inspectors later reviewed

the 10 CFR 50.59 determination and engineering evaluation which dispositioned the

NFPA code deviation. The inspector concluded that the corrective actions taken to

resolve the issue were adequate. This licensee-identified, non-repetitive and

corrected violation of the Operating License is being treated as a Non-Cited

Violation consistent with Section Vll.B.1 of the NRC Enforcement Poliev." This LER

is closed. (NCV 50-352;353/98-08-04)

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E8.5 (Closed) LER 50-352: 353/1-98-015: Condition Prohibited by Technical

Specifications Due to an Error in Calibration of Core Sorav Line Break Differential

Pressure Instruments.

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On July 11,1998, PECO Engineering identified that an error in the calibration of the

Unit 2 core spray (CS)internalline break detection differential pressure j

instrumentation had resulted in a setpoint that was outside the band required by

Technical Specifications. The effected instrumentation is used to detect an

abnormal differential pressure between the piping of the two redundant CS systems,

thereby detecting a break in the piping of one of those systems. Due to

configuration differences in the piping, a normal differential pressure exists between

the two CS systems. Technical Specifications prescribes a value, above and below

that normal differential pressure value, at which the detection instrumentation must

alarm to warn operators of a break in the system piping. The July 11, discovery

was due to the fact that, since power uprates had been implemented at Unit 1 in i

February 1996 and at Unit 2 in February 1995, the detection instruments had been

calibrated assuming that the normal differential pressure between the two CS

systems was O psid. The actual differential pressure between the two systems

during normal rated power conditions is -2.5 psid. This value is approximately the

same at both units, and because it was not properly considered during the

calibration of the detectors both units had not been in compliance with the

differential pressure band specified in their Technical Specifications since the time

of their power uprate.

PECO's corrective actions included the proper recalibration of the CS line break

detection instrumentation at both units and the review of the calibration process for

similar instrumentation which confirmed that the power uprate had not had

adverse'ly affected any other setpoints. The inspector conducted an on-site review

and concluded that the licensee's analysis of, and corrective actions for, the event

were adequate. This licensee-identified, non-repetitive and corrected violation of

Technical Specifications is being treated as a Non-Cited Violation consistent with

Section Vll.B.1 of the NRC Enforcement Poliev." This LER is closed. (NCV 50-352:

, 353/98-08-05)

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E8.6 (Closed) LER 50-352: 353/1-98-017: Failure of Hatchway Fire Protection Flow

Control Valves to Actuate.

On July 28,1998, PECO Engineering determined that six fire protection system

hatchway valves (three on each unit) may have been incapable of performing their

! design function during a postulated fire event. The problem was discovered at

Unit 1 while troubleshooting activities were being performed to fix a leaking block

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valve in the system. PECO determined that the timer switch settings in the control

l panel for the flow control valves would de-energize the solenoid valve after

approximately five seconds, closing the flow control valve sooner than expected.

PECO initially suspected that all six flow control valves were similarly affected, but

later learned that the Unit 2 valves had been corrected in 1989 after the design

deficiency was first identified. The similar proposed design change to correct the

Unit 1 valves was canceled, apparently due to the licensee's belief that interim

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corrective actions were adequate to resolve the issue. The July 1998 discovery

revealed the cancellation to have been in error. The licensee determined the ,

surveillance test for these valves had been inadequate in that the procedure only l

verified flow was established upon actuation, not that it would be sustained for the

required time.  ;

PECO corrective actions initially consisted of performing a firewatch for the affected

valves, to ensure proper manual actuation if required, while a design change for the l

Unit 1 valves was implemented. The inspectors performed a field walk down of the

hatchway fire protection system, verified implementation of the compensatory fire

watches, and observed portions of the design change including the post

modification testing. The inspectors concluded that PECO's corrective actions were

adequate and satisfactorily implemented. This licensee identified, non repetitive,

and corrected violation of 10 CFR 50 Appendix B, Criterion Ill, Design Control is

being treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC

Enforcement Policy. This LER is closed. (NCV 50-352/98-08-06)

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E8.7 (Closed) LER 50-353/2-98-001: Three Inoperable Barksdale Model C9622-3-B

pifferential Pressure Switches Result in Two or More Independent Trains of a Sinale

Safety System Beina Inocerable From a Common Cause.

On June 3,1998, during the implementation of a setpoint change of the Barksdale

differential pressure switches in the relayed emergency trip system (RETS), PECO

found that three of the four switches had fallen below the allowable setpoint value

which is prohibited by Technical Specification. The function of these RETS pressure

switches is to provide an anticipatory trip signal to the end-of-cycle reactor

recirculation pump trip system and to the reactor protection system for a main

turbine trip. The setpoint change was being implemented to accommo'date

additional setpoint drift to address a similar problem with Barton pressure switches

used to provide the same function on the Unit 1 RETS system.

During an in-office review, the inspector determined that PECO had previously

evaluated the impact of instrument drift for the RETS pressure switches in

conjunction with the 24-month fuel cycle review. The study evaluated the impact

of a 200 psig instrument drift and found that this drift would have delayed the trip

actuation by only 3 milliseconds and that such delay would have had minimal

impact on the overall TS-required response time of the trip function. Further, based

on the most recent response time test data overall response times remained within

the bounding values of the transient analysis. The inspector concluded that

although previously evaluated, this event was a result of inadequate margins to

account for setpoint drift over a 24-month fuel cycle. The inspector also concluded

the corrective actions implemented to resolve this issue including recalibration of

the pressure switches and raising the setpoints an additional 100 psig (total of 200

psig change) to ensure adequate margin to the TS allowable value is maintained

were appropriate.

17

This licensee identified, non repetitive, and corrected violation of Technical

Specifications surveillance requirements is being treated as a non-cited violation,

consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-353/98-

08-07) This LER is closed.

IV. Plant Support

R1 Radiological Protection and Chemistry (RP&C) Controls

R1.1 Solid Radwaste Processina

a. Inspection Scoce (86750)

Plant tours were conducted to review the solid radwaste processing activities with

respect to Updated Final Safety Analysis Report (UFSAR) descriptions and radwaste

sampling, characterization, and waste classification requirements.

b. Observations and Findinas

Limerick radwaste liquids were processed through powdered and bead resins as

described in the UFSAR. Condensate liquids were filtered using a precoatless filter

and the backwash filtrate represented a second waste stream. RWCU powdered

resin represented a third waste stream. Contaminated trash represented the final

waste stream. Representative samples of each waste stream were taken and

analyzed on an annual basis. Quantification of resin and contaminated trash (dry

active waste) waste streams utilized accepted methodologies. Quantification of

condensate filtrate wastes were generally estimated without an established

measurement methodology.' During the inspection, an acceptable approach was'

developed by the licensee and entered into the corrective action program for

resolution. Due to the relatively low volumes and radioactivity of the condensate

filtrate wastes, no difference in waste classification would have resulted from the

observed inaccuracies in volume estimates.

Resin / condensate filtrates were dewatered to less than 1 % free standing water

utilizing an NRC approved process control program as required.

c. Conclusions

Limerick solid radioactive wastes were effectively sampled, packaged, and

dewatered with respect to requirements. The radwaste staff is pursuing an

enhancement to the program to more accurately quantify the condensate filtrate

waste volumes.

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R1.2 Radioactive Material Shinoino

a. Insoection Scope (86750)

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Two radioactive material outgoing shipments were observed and selected 1998 l

shipping records were reviewed with respect to 10 CFR 20,61,71, and l

49CFR171-179 requirements.

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b. Observations and Findinos

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A seavan bulk shipment of dry active wastes and a cask shipment of spent resin

were properly packaged, marked, and placarded for shipmeat. The shipment

preparation crew worked well together in an expeditious manner with no

deficiencies observed. All shipping papers were in accordance with regulatory

requirements.

c. Conclusions

Radioactive material shipments were prepared in an expeditious manner and met all

regulatory requirements. Shipping records were properly prepared with no

deficiencies identified.

R1.3 Solid Radioactive Waste Storace

a. Insoection Scope (86750)

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Limerick plant areas were toured to observe the condition of radioactive material

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storage areas. The Limerick Low Level Radioactive Waste Storage Facility

(LLRWSF) condition was also reviewed.

b. Observations and Findinos

Limited amounts of stored contaminated equipment were properly maintained and

controlled. Located within the radwaste building, there was an inventory of

3,200 ft' polyethylene liners of filters, four liners of spent resin and one liner of

spent reactor water cleanup resin. This was considered a normal backlog and well

within the design of the radwaste high level storage area.

The LLRWSF did not contain any stored radioactive wastes. The adjacent area

contained approximately 17 seavans of reusable outage equipment that was

properly posted and inventoried.

c. Conclusions

The licensee has effectively minimized the amount of contaminated equipment and

radioactive wastes stored onsite.

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R1.4 Radioloaical Controlled Area Material Monitorina

a. Insoection Scope (86750)

Radwaste staff monitoring of material to be released from the radiological controlled

area (RCA) was observed and the applicable procedure was reviewed.

b. Observations and Findinas

Radwaste personnel release " green is clean" material collected inside the

radiological controlled area (RCA) utilizing a small article monitor (SAM), monitoring

a bag full of material at a time. The individualitems were not smeared or direct

frisked. The licensee indicated that plant practice dictated that only items 6xiting a

posted contamination area were monitored individually with both smears enu direct

frisk surveys. With the help of a umerick fully qualified radiation protection

technician, the inspector determined the capability for the SAM monitor to detect

contamination on a single item located in the center of the detector cavity. Based

on a frisker efficiency of 5%, the SAM monitor did not alarm in 5 out of 5 counts

until approximately 10,000 dpm of activity was accumulated.

Procedure HP-C-810, Rev. 3, " Radioactive Material (RAM) Control", Section 7.5

specifies that all material shall be monitored prior to release from the RCA; and that

material to be released from the RCA shall meet the following conditions:

smearable < 1000 dpm/100cm2 and total (smearable and fixed)

< 5000 dpm/100cm 2. NRC Circular 81-07 also indicates that licensees are

expected to monitor to at least the sensitivity as stated in the Limerick procedure.

The plant practice of monitoring the " green is clean" materials was not in

accordance with procedure, but were being monitored with assurance that no

radioactive material greater than 10,000 dpm was released. The licensee indicated

that this area would be reviewed and evaluated. Due to the minor safety

significance of this practice, this is considered a violation of minor significance that

is not subject to formal enforcement action.

c. Conclusions

Monitoring of material exiting the radiological controlled area was not always

conducted at the low sensitivities specified by station procedure.

R3 RP&C Procedures and Documentation

R3.1 Radioactive Material Shioment Procedures

a. Insoection Scope (86750)

The following procedures were reviewed with respect to DOT and NRC radioactive

material transportation regulations.

RW-C-100, Rev. 4, " Solid Radwaste System Process Control Program"

. _ _

,

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RW-429, Rev. 2, " External Processing Station Resin Transfer and Dewatering for

. Rapid Dewetering, using Vendor Compression Dewatering System"

RW-C-242, Rev. 4, " Packaging Radioactive Material"

RW-C-244, Rev. 5, " Shipping Radioactive Material"

.

RW-C-255, Rev.1, " Characterizing and Classifying Packages"

RW-226, Rev.11, "Radwaste and Radioactive Materialinspection and Loading

Operations"

RW-C-110, Rev. 2, "10CFR61 Compliance Program"

RW-C-201, Rev.1, "Quantification and Classification of Radioactive Material"

b. Observations and Findinas

The radwaste and radioactive material transportation procedures reviewed were of

good quality and accurately reflected regulatory requirements.

c. - Conclusions

Limerick radioactive waste processing and radioactive material shipping procedures

were of good quality and effectively implemented regulatory requirements.

,

R5- S' taff Training and Qualification in RP&C

R5.1 Radioactive Material Shioment Trainina

a. Insoection Scope (86750)

- Radioactive material shipping lesson plans and training attendance documentation

were' reviewed, and interviews with cognizant licensee individuals were conducted

with respect to 49CFR172 Subpart H and NRC IE Bulletin No. 79-19 requirements.

b. Observations and Findinas

For Limerick Station, radioactive material shipments were accomplished by four

authorized shippers who also provided shipment verification prior to departure from

the plant. Training records were verified to be current with annual training provided

for all four individuals. The licensee's in-house training program was of good

quality, reflecting current NRC and DOT regulations.

A

c. Conclusions

All authorized radioactive material shipment personnel have met the applicable DOT

and NRC training requirements.

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R7 Quality Assurance in Radiological Protection and Chemistry Activities l

R7.1 Radioactive Material Shionina OA Oversiaht

a. hsoection Scope

A quality assurance (QA) assessment of radioactive material shipping activities,

dated May 1,1997 was reviewed as well as ten QA surveillances of the program

area conducted during 1997 through the date of this inspection. In addition, I

radioactive waste processing and transport vendor audits were reviewed in I

accordance with IE Bulletin 79-19 requirements.

l

b. _ Observations and Findinas I

le Quality Assurance assessment conducted March 25,1997 through May 1,  ;

1997, was a sufficiently broad and detailed review of the solid radwaste and l

radioactive material transport program area and indicated that the program was  !

effectively implemented. In addition, during the past 18 months, there have been

10 QA surveillancec that included: three radwaste shipments, resin dewatering )

activities, burning of contaminated oil, fuel poolinventory, and store room receipt of l

radioactive material. Spot checks of outgoing radioactive material shipments were  !

made and the radwaste authorized shippers provided peer review verifications of

each outgoing shipment. Results have been good, without any non-compliances

identified. Several offsite vendors supply transfer, packaging and transport of l

licensee's radioactive waste and fall within the audit requirements of IE Bulletin

79-19. These include: Molten Metal Technology, Frank Hake, GTS Duratek, ATG,

U.S. Ecology, and Chem Nuclear Systems, Inc. Vendor audits were only available  !

for Molten Metal' Technology and Chem Nuclear Systems, Inc., althou'g h the other

vendor licensees were verified to be on the Nuclear Utilities Procurement issues

Council (NUPIC) list. The licensee stated that the other radioactive material

processing vendor audits would be obtained and reviewed on a regular basis.

c. Conclusions

Quality assurance oversight of the radioactive material shipment program was

effective through performance of an independent program assessment and

surveillances and through radwaste staff shipment verifications. )

S1 Conduct of Security and Safeguards Activities

a. Inspection Scope (81700)

Determine whether the conduct of security and safeguards activities met the

! licensee's commitments in the NRC-approved security plan (the Plan) and NRC

regulatory requirements. The security program was inspected during the period of

l September 21-24,1998. Areas inspected included: access authorization program;

altsrm stations; communications; protected area access control of personnel and

packages.

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b. Observations and Findinas

l Access Authorization Pronram. The inspectors reviewed the Access Authorization

(AA) program to verify implementation was in accordance with applicable regulatory

requirements and Plan commitmentt. The review included an evaluation of the

effectiveness of the AA procedures, as implemented, and an examination of AA l

records for 15 individuals. Records reviewed included both persons who had been

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granted and had been denied access. The AA program, as implemented, provided

assurance that persons granted unescorted access did not constitute an

'

unreasonable risk to the health and safety of the public. Additionally, the inspectors

l reviewed access denial records and applicable procedures to verify that appropriate

actions were taken when individuals were denied access or had their access

I terminated.

Alarm Stations. The inspectors observed operations of the Central Alarm Station

(CAS) and the Secondary Alarm Station (SAS) and verified that the alarm stations

were equipped with appropriate alarms, surveillance and communications

capabilities. Interviews with the alarm station operators found them knowledgeable

of their duties and responsibilities. The inspectors also verified, through

observations and interviews, that the alarm stations were continuously manned,

independent and diverse so that no single act could remove the plant's capability for

detecting a threat and calling for assistance and the alarm stations did not contain

l any operational activities that could interfere with the execution of the detection,

! assessment and response functions.

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Communications. The inspectors verified, by document reviews and discussions

with alarm station operators, that the alarm stations were capable of maintaining

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continuous intercommunications, continuous communications with each security '

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force member (SFM) on duty, and alarm station operators were testing

l communication capabilities with the local law enforcement agencies as committed

to in the Plan.

Protected Area (PA) Access Control of Personnel and Hand-Carried Packaaes. On

September 23 and 24,1998, during peak activity periods, the inspectors observed

personnel and package search activities at the personnel access portal. The

inspectors determined, by observations, that positive controls were in place to

ensure only authorized individuals were granted access to the PA and that all

personnel and hand-carried items entering the PA were properly searched.

c. Conclusions

The licensee was conducting its security and safeguards activities in a manner that

protected public health and safety and that this portion of the program, as

implemented, met the licensee's commitments and NRC requirements.

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S2 Status of Security Facilities and Equipment

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, a. insoection Scoce (81700)  !

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Areas inspected were: PA assessment aids; PA detection aids and personnel search

equipment.

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i b. Observations and Findinas

!

Assessment Aids. On September 22,1998, the inspectors evaluated the '

,

effectiveness of the assessment aids, by observing on closed circuit television

l (CCTV), a SFM conducting a walkdown of the PA. The assessment aids had good

! picture quality and excellent zone overlap. Additionally, to ensure Plan

! commitments are satisfied, the licensee has procedures in place requiring the

'

implementation of compensatory measures in the event the alarm station operator is

unable to properly assess the cause of an alarm.

PA Detection Aids. On September 22,1998, the inspectors observed testing of l

selected intrusion detection zones in the plant protected area. The inspectors

'

determined, by observations and by reviewing the testing documentation associated

with the equipment repairs, that repairs were made in a timely manner and that the

equipment was functional and effective, and met the commitments in the Plan.

Personnel and Packaae Search Eauiomen.1. On September 24,1998, the inspectors

observed both the routine use and the weekly performance testing of the licensee's

personnel and package search equipment. Personnel search eqi.ipment was being

tested and maintained in accordance with licensee procedures and the Plan and '

personnel and packages were being prop'erly searche f pdor to PA access.

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The inspectors determined, by observations and procedural reviews, that the search

equipment performed in accordance with licensee procedures and Plan  ;

commitments. l

c. Conclusions

The licensee's security facilities and equipment were determined to be well

maintained and reliable and were able to meet the licensee's commitments and NRC

requirements.  ;

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S3 Security and Safeguards Procedures and Documentation

j a. Inspection Scope (81700)  :

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. Areas inspected were: implementing procedures and security event logs. l

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l b. Observations and Findinas

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!- ' Security Proaram Procedures. The inspectors verified that the procedures were

j' consistent with the Plan commitments, and were properly implemented. The

l

verification was accomplished by reviewing selected implementing procedures

associated with PA access control of personnel, testing and maintenance of

personnel search equipment and the vehicle barrier system.

Security Event Loas. The inspectors reviewed the Security Event Log for the

previous six months. Based on this review, and discussion with security

management, it was determined that the licensee appropriately analyzed, tracked,

I

resolved and documented safeguards events that the licensee determined did not

require a report to the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

! c. Conclusions

l

Security and safeguards procedures and documentation were being properly

implemented. Event Logs were being properly maintained and effectively used to

analyze, track, and resolve safeguards events.

S4 Security and Safeguards Staff Knowledge and P.erformance

a. inspection Scope (81700)

i

Area inspected was security staff requisite knowledge.

b. Observations and Findinas

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Security Force Reauisite Knowledae. The inspectors observed a number of SFM's  !

in the performance of their routine duties. These observations included alarm j

l station operations, personnel and package searches, and exterior patrol alarm

l response. Additionally, the inspectors interviewed SFMs and based on the

responses to the inspector's questioning, determined that the SFMs were

knowledgeable of their responsibilities and duties, and could effectively carry out l

their assignments.

c. Conclusions

, The SFMs adequately demonstrated that they have the requisite knowledge

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necessary to effectively implement the duties and responsibilities associated with

l their position.

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l S5 Security and Safeguards Staff Training and Qualification .

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a. Insoection Scope (81700)

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Areas inspected were security training and qualifications and training records.

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b. Observations and Findinos

1

Security Trainina and Qualifications. On September 23,1998, the inspectors 1

selected and reviewed T&O records of 7 SFMs. The results of the review indicated i

that the security force was being trained in accordance with the approved T&Q l

plan.

Trainina Records. The inspectors were able to verify, by reviewing training records,

that the records were properly maintained, accurate and reflected the current

qualifications of the SFMs.

c. Conclusions l

l

Security force personnel were being trained in accordance with the requirements of  !

the T&O Plan. Training documentation was properly maintained and accurate and

.

j

the training provided by the training staff was effective. l

S6 Security Organization and Administration

a. Inspection Scope (81700)

Areas inspected were management support and staffing levels.

b. Observations and Findinas

Manaaement Suocort. The inspectors reviewed various program enhancements

made:sinco the last program inspection, which was conducted in March 1998.

These enhancements included upgrades to the alarm assessment' systems and

firearms. training facilities.

Staffina Lovels. The inspectors verified that the total number of trained SFMs

immediately available on shift met the requirements specified in the Plan. 1

c. Conclusions

The level of management support was adequate to ensure effective implementation

of the security program, and was evidenced by the allocation of resources to

support programmatic needs.

V. Management Meetings

X1 Exit Meeting Summary

l The inspectors presented the inspection results to members of plar.t management at

( the conclusion of the inspection on October 23,1998. The plant manager

acknowledged the inspectors' findings. The inspectors asked whether any materials

examined during the inspection should be considered proprietary. No proprietary

j information was identified.

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The inspector met with licensee representatives at the conclusion of the radwaste

transportation and security inspections on September 18 and September 24,1998,

respectively. At that time,' the purpose and scope of the inspection were reviewed,.

and the preliminary findings were presented. - The licensee acknowledged the

preliminary inspection findings .

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ATTACHMENT 1,

INSPECTION PROCEDURES USED'

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IP 61726: Surveillance Observation 'I

~lP 62707: Maintenance Observation

IP 71001: . Licensed Operator Requalification Program Evaluation I

~ IP -71707: . Plant Operations .

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. lP 81700: Physical Security Program for Power Reactors

LIP 86750: . Occupational Radiation Exposure . j

IP 90712: In-office Review of Written Reports

L IP 92700: ' On-site Follow-up of Written Reports - 1

L IP 92702: - Follow-up on Corrective Actions  !

iP 92902: Follow-up Maintenance i

IP 92903: Follow -up Engineering l

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l PARTIAL LIST OF PERSONS CONTACTED )

LICENSEE PERSONNE.L,

j

M. Gallagher, Plant Manager ,

M. ,Karney, Security / Emergency Planning Manager i

D. LeQuia, Director, Site Support Services  ;

R. Bixler, Corporate Security investigation i

R. Eickhardl NOA Assessor - l

H. McNeill, Manager, Industrial Risk ')

,

J. Spaniel, Security Systems Manager

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' B. Whitman, Security Supervisor , ,

f. C. Coimbach, Security Supervisor

J. Lotz, Security Supervisor.

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ITEMS OPENED, CLOSED, AND DISCUSSED k

e

l4 QgirJed/ Closed

l- NCV 50-353/98-08-01 Condition Prohibited by Technical Specification in that

the Main Condenser Offgas Pre-treatment Radiation

Monitor was Inoperable and the Action was not met

r due to an incorrect Procedure. (Section 08.2)

. NCV 50-353/98-08-02 - Failure to meet undervoltage channel calibration

.-technical specification surveillance requirement.

- (Section M8.1)

NCV 50-352; 353/98-08-03 Potential containment bypass path resulting in a

. condition outside the design basis. (Section E2.1)

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Attachment 1 2

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NOV 50-352; 353/98-08-04

'

Failure to Meet the Maximum Travel Distance Limitation

for Portable Fire Extinguishers. (Section E8.4)

NCV 50-352; 353/98-08-05 Condition Prohibited by Technical Specifications Due to l

an Error in Calibration of Core Spray Line Break i

Differential Pressure Instruments. (Section E8,5)

NCV 50-352; 353/98-08-06 Failure of Hatchway Fire Protection Flow Control Valves

to Actuate. (Section E8.6)

i

NCV 50-353/98-08-07 Three Inoperable Barksdale Model C9622-3-B l

Differential Pressure Switches Result in Two or More

Independent Trains of a Single Safety System Being

inoperable From a Common Cause. (Section E8.7)

Closed

LER 50-352; 353/1-98-013 Failure to Meet the Maximum Travel Distance Limitation

for Portable Fire Extinguishers. (Section E8.4)

LER 50-352; 353/1-98-014 ESF actuation due to reactor water cleanup system

isolations. (Section E8.2)

LER 50-352; 353/1-98-015 Condition Prohibited by Technical Specifications Due to ;

an Error in Calibration of Core Spray Line Break

Differential Pressure Instruments. (Section E8.5)

LeR 50-352; 353/1-98-016 Manual MCR ventilation isolation and CREFAS initiation

due to small Freon leak. (Section 08.1)

LER 50-352; 353/1-98-017 Failure of Hatchway Fire Protection Flow Control Valves

to Actuate. (Section E8.6)

LER 50-353/2-98-001 Three Inoperable Barksdale Model C9622-3-B

Differential Pressure Switches Result in Two or More

independent Trains of a Single Safety System Being

inoperable From a Common Cause. (Section E8.7)

LER 50-353/2-98-003 Condition Prohibited by Technical Specification in that

the Main Condenser Offgas Pre-treatment Radiation

Monitor was inoperable and the Action was not met

i due to an incorrect Procedure. (Section 08.2)

LER 50-353/2-98-004 Secondary containment isolation, standby gas treatment

system and reactor enclosure recirculation system

initiation. (Section E8.1)

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Attachment 1 3

LER 50-353/2-98-006 Failure to meet undervoltage channel calibration

l

technical specification surveillance requirement.

(Section M8.1)

l IFl 50-352/97-07-02 Reactor Water Cleanup (RWCU) Isolations. (Section

E8.2)

VIO 50-352,353/98-04-04 Failure to Submit Licensee Event Report (Section E8.3)

'

! URI 50-352; 353/98-05-05 Potential containment bypass path resulting in a

condition outside the design basis. (Section E2.1) i

!

LER 50-352: 353/1-97-010 Potential containment bypass path resulting in a  ;

condition outside the design basis. (Section E2.1) ,

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Attachment 1 4

-

. LIST OF ACRONYMS USED

AA Access Authorization -

, ,

.CAS: -

Central Alarm Station

CCTV. Closed Circuit Television-

L'* iCFR' , Code of Federal Regulations -

CREFAS: Control Room Engineering Fresh Air System

i

CRSL> ' Control Room Supervisor

DOT-- U. S. Department of Transportation I

c DW: Drywell : .

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EHC: Electrohydraulic control

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-EPRI Electric Power Research Institute  !

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. ESF Engineered Safety Feature

FMTF- Fuel Monitoring Task Force

GP=

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. General Procedure .

HPCI- . High Pressure Coolant Injection

1&C ' Instrumentation and Control

IFl Inspection Follow-up item i

IR" _ . Inspection Report

JPM Job Performance Measures

l~ LC . Load Center i

p- -LER Licensee Event Report

LGS . . Limerick Generating Station .

l LLRWSF . Low Level Radioactive Waste Storage Facility _

! LOCA Loss Of Coolant Accident 'I

' LSRO - Limited Senior Reactor Operator

MCP Main Control Room '

NCV Non-Cited Violation

NQA Nuclear Quality Assurance '

NRB ~ ' Nuclear Review Board  ;

NRC' Nuclear Regulatory Commission

< NUPlc - Nuclear Utilities Procurement issues Committee l

r ON- Off-Normal  !

PA Protected Area

PCP Process Control Program

PECO PECO Energy

L PECON ' PECO Nuclear

PEP..- Performance Enhancement Program

- the Plan-~ NRC-Approved Physical Security Plan

.

-OA- Quality Assurance .

[ RAM -- Radioactive Material

L RBM. Rod Block Monitor

L. RCA. . Radiological Controlled Area

' RCIC Reactor Cora Isolation Cooling

RERS - Reactor Enclosure Recirculation System '

- RHR Residual Heat Removal

RO , Reactor Operator

, RP&C. Radiological Protection and Chemistry

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Attachment 1 5 '

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RWCU Reactor Water Clean-up

SAM Small Article Monitor

S A.S - Secondary Alarm System

SFM Security Force Member - i

SGTS. _ Standby Gas Treatment System j

SJAE Steam Jet-Air Ejector

SSC Systems, Structures, & Components

-ST Surveillance Test

SUN Shift Update Notice

TS Technical Specification

T&Q Training and Qualification j

UFSAR Updated Final Safety Analysis Report )

URI Unresolved item

VIO Violation

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