ML20135F571

Insp Rept 50-309/96-12 on 960915-1026.No Violations Noted. Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20135F571
Person / Time
Site:	Maine Yankee
Issue date:	12/09/1996
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20135F544	List: ML20135F541 ML20135F553 ML20135F571
References
50-309-96-12, NUDOCS 9612130130
Download: ML20135F571 (87)

See also: IR 05000309/1996012

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U. S. NUCLEAR REGULATORY COMMISSION

REGION i

Docket No:

50-309

License No:

DPR-36

Report No:

50-309/96-12

Licensee:

Maine Yankee Atomic Power Company (MYAPC)

Facility:

Maine Yankee Atomic Power Station

Location:

Bailey Point

Wiscasset, Maine

Dates:

September 15 through October 26,1996

Inspectors:

Jimi Yerokun, Senior Resident inspector

Division of Reactor Projects

William Olsen, Resident inspector

Division of Reactor Projects

{

John Lusher, Emergency Prepardness Specialist

Division of Reactor Safety

Lonny Eckert, Radiological Controls Specialist

Division of Reactor Safety

Jason Jang, Radiological Controls Specialist

Division of Reactor Safety

Approved by:

Richard Conte, Chief, Reactor Projects Branch No. 5

Division of Reactor Projects

9612130130 961209

PDR

ADOCK 05000309

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EXECUTIVE SUMMARY

Maine Yankee Atomic Power Company

NRC Inspection Report 50-309/96-12

This integrated inspection included aspects of licensee operations, engineering,

maintenance, and plant support. The report covers a 6-week period of resident inspection;

in addition, it includes the results of announced inspections by regional inspectors in the

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areas of Radiological Controls and Emergency Prepardness.

Ooerations

The operators performed all observed tasks in accordance with station procedures with a

very good safety perspective in evidence. The decision to reduce station power to

investigate the bearing cooling water problem with the circulating water pump was well

controlled. Operators responded well to the unplanned reactor trip and properly completed

the steps of emergency operating procedure E-0, Reactor Trip or Safety injection; along

with other challenges due to the material condition problems. The Shift Operating

Supervisor displayed excellent command and control of the operating crew during the

period of observation. The Post Trip Review Team provided a very detailed report with an

excellent basis for the conclusions that were drawn and the recommended corrective

actions (short and long term). Station management provided good oversight and direction

to ensure that all the necessary corrective actions were completed prior to plant restart.

A violation of the station Off-Site Dose Calculation Manual and T.S. 5.8.a.3 was identified

when the Primary Vent Stack Air Particulate and Gas Monitoring System filters were not

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installed during maintenance and apparently due to procedure inadequacy and/or personnel

error (VIO 50-309/96-12-01, Section O2.1). Prompt immediate corrective action was

)

taken, however, root cause and long term corrective actions are yet to be determined.

Instances of operator performance weakness were noted, such as the unplanned power

excursion during a RCS delithiation. While operators performed well to restore the reactor

to a steady power level of 2440 MWt, performance weaknesses contributed to the power

excursion occurring in the first case. The licensee's root cause analysis was still on-going

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and the appropriateness of implemented corrective actions still needs to be verified

(URI 50-309/96-12-02, Section 04.1).

PORC meetings were conducted well and safety focused. PORC continued to demonstrate

good technical and safety perspective in addressing plam~ issues.

Maintenance

l&C personnel provided good support to the plant and conducted troubleshooting activities

in a safe and controlled manner. In the case of the RPS trip breakers the decision to

replace all the circuit components was very conservative and displayed an excellent safety

perspective. Station maintenance management properly provided guidance to ensure that

all possible avenues of repair were exercised prior to completion of repairs.

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Maine Yankee promptly accessed the operability of the containment high range radiation

monitor and conservatively declared the channel inoperable until troubleshooting and

calibration activities were complete.

Electrical maintenance personnel demonstrated excellent support to operations by timely

identifying and resolving the problem with control rod (malfunctions), thereby preventing

an unnecessary transient to the plant. However, based on problems experienced in the

past, it appeared that this was indicative of a weakness in the licensee's process. What

appeared to be appropriate actions were planned and are being taken to control the

problem. During and following troubleshooting activities, plant personnel showed good

safety perspective and ensured that the plant was always maintained safely. However, the

problems with the Control Element Assemblies (CEAs) were similar to those that occurred

during reactor startup the previous month and it appears that the right solution of the

previously identified problems had not been made. While the problems concern a non-

safety related system, they produced unnecessary burden on control room operators during

their operation of the plant.

l&C personnel responded timely and performed well by safely and quickly diagnosing the

problem (failed RPS channel C trips) and correcting it. Plant personnel showed good safety

focus and ensured that the plant continued to be operated safely while the troubleshooting

and repair efforts were on-going.

Enaineerina

Engineering personnel demonstrated good efforts at addressing the turbine building design

issues affecting cold weather operations. The activities of the corporate engineering and

licensing department personnel appeared technically sound and detailed. There was good

management attention and involvement in addressing these issues.

Plant Sucoort

Maine Yankee radiation protection department properly responded to unplanned exposure

events. The analysis of the data was thorough with proper review by independent

expertise. Most of the conclusions were appropriate. However, the inspector noted that

one of the conclusions for the Duratek event did not properly take into consideration the

experience and training of the two individuals that were involved in the event. While no

personnel exposures exceeded regulatory limits, the contamination and exposure controls

at Maine Yankee were weak.

The licensee established, implemented, and maintained effective radioactive liquid and

gaseous effluent control programs. The Radiation Monitoring Systems (RMSs) for effluent

processes were well maintained. Licensee completion of several RMS upgrades and the

RMS manual demonstrated the licensee's commitment to a strong Radiological

Environmental Monitoring Program (REMP).

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Overall, maintenance and surveillance of the ventilation systems was adequate. The

licensee had not documented the plant air balance surveillance results to demonstrate

consistency with Chapter 9.13 of the UFSAR assumptions. Section 9.13.2.4 of the

USFAR had not been updated. These items will be reviewed during a subsequent

inspection (IFl 50-309/96-12-03, Section R2.3).

The licensee's Off-Site Dose Calculation Manual (ODCM) contained sufficient specification,

information, and instruction to implement and maintain the radioactive liquid and gaseous

effluent control programs. The content of the Annual Report was very good and met the

TS reporting requirements. Oversight of the REMP was improved. No degradation was

noted as a result of the reorganization.

In the area of Emergency Preparedness both the CATS and the self-assessment program

appeared to be an effective licensee control. Facility inventories were complete, radiation

survey instrumentation was within the calibration requirements and the emergency

response facilities were found to be in a state of operational readiness. However, an

instance was identified where maintenance activity might have negatively affected the

Emergency Response Facility (ERF) ventilation system (URI 50-309/96-12-04, Section

P8.2).

The training program was being effectively implemented and the Emergency Response

Organization (ERO) is adequately staffed. The licensee maintained on-shift dose

assessment and adequate back-up capabilities to ensure that on-shift dose assessments

could be performed.

A violation for failure to properly control Safeguards Information SI was identified

(VIO 50-309/96-12-05, Section S1.1). This violation is notable in that there have been

severalinstances of failure to properly control safeguards information in the last three

years. It appears that previous corrective actions have been too narrow in scope to

prevent recurrence.

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Maine Yankee, with assistance from off site security specialists, properly performed the

required annual audit of the security program. The audit was comprehensive with the

proper involvement of the station quality programs department. Corrective actions were

appropriate to resolve the observations. However, the observation concerning the control

of Si material appears to require more extensive corrective actions to resolve the problem

of continued failures in this area.

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TABLE OF CONTENTS

TA B LE O F C O NTE NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v

l . O p e ra t io n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

01

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

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

01.2 Plant Trip During Surveillance Testing . . . . . . . . . . . . . . . . . . . . . 1

02

Operational Status of Facilities and Equipment

3

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02.1 Primary Vent Stack Sampling Filters Not Installed

(VIO 50-309/96-12-01V)

3

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02.2 Loss of Plant Computer . . . . . . .

4

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02.3 Spurious Trip of Reactor Protection System, Channel C . . . . . . . . 5

04

Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 6

04.1 Unplanned Reactor Power increase During Delithiation

(URI 50-309/9 6- 12-0 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

08

Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

08.1 Plant Operations Review Committee . . . . . . . . . . . . . . . . . . . . . . 8

11. Maintenance . . . . . . . . . . .

9

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M1

Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

M 1.1 G eneral Activitie s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

M1.2 Troubleshooting of Reactor Protection System Matrix

Trip Circ uit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

M1.3 Troubleshooting and Repair of Containment Radiation Monitor . . 10

M2

Maintenance and Material Condition of Facilities and Equipment . . . . . . 11

M2.1 Broken Handwheel on Valve LSI-M-11

11

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M2.2 Operability of Containment High Radiation Monitor

12

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M2.3 Inoperable Pressurizer Proportional Heater Trains . . . . . . . . . . . . 13

M2.4 Control Element Assernbly Problems During Reactor Startup . . . . 14

111. Engineering

16

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E2

Engineering Support of Facilities and Equipment

16

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E2.1 Design issues Affecting Cold Weather Operations

in the Turbine Building (update URI 50-309/96-08-02 and

50-309/96-08-04)

16

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I V. Pl a nt S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

R1

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

R 1.1 Multiple Contamination Events and Unplanned Exposures

17

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R1.2 Implementation of Radioactive Liquid and Gaseous Effluent

Co ntrol Prog ra m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

R2

Status of RP&C Facilities and Equipment

20

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R2.1 Effluent / Process Radiation Monitoring Systems . . . . . . . . . . . . . 20

R2.2 Air Clea ning Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

R2.3 Ventilation Systems Air Balance (IFl 50-309/96-12-03)

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R3

RP&C Procedures and Documentation

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R3.1 Off Site Dose Calculation Manual Review . . . . . . . . . . . . . . . . . 23

R6

RP&C Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 24

R6.1 Radioactive Liquid and Gaseous Effluent Program Review

24

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R7

Quality Assurance in RP&C Activities . . . . . . . . . . . . . .

25

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R7.1 Quality Assurance Audit Report Review . . . . . . . . . . . . . . . . . . 25

R8

Miscellaneous RP& C issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26

R8.1 Review of Updated Final Safety Analysis Report (UFSAR)

Com mitm e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26

P1

Conduct of Emergency Preparedness (EP) Activities

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P1.1 Emergency Planning Self-Assessment Program and Corrective

Action Tracking System

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P2

Status of EP Facilities, Equipment, and Resources . . . . . . . . . . . . . . . . 27

P2.1 Emergency Planning Equipment Inventories and Surveillance

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P3

EP Procedures and Documentation

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P3.1 Review of Emergency Response Plan Changes (Closed, URI 50-

309/9 6-007-01 ) . . . . . . . . . . . . . . . . . . . . . . . . . .

28

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P5

Staff Training and Qualification in EP . . . . . . . . . . . . . . . . . . . . . . . . . 29

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PS.1 Emergency Planning Training Program Evaluation

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P6

EP Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . . . 30

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P6.1 Emergency Planning Staffing and Management Changes . . . . . . 30

P7

Quality Assurance (QA) in EP Activities

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P7.1 Review of Annual Emergency Planning Program Audit Reports . . 30

P8

Miscellaneous EP Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31

P8.1 Updated Final Safety Analysis Report (UFSAR) Inconsistencies . . . 31

P8.2 Emergency Response Facilities Ventilation System

(U RI 50-3 0 9/9 6-12-04) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32

S1

Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . 33

S1.1 Control of Safeguards Information (VIO 50-309/96-12-05)

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S7

Quality Assurance in Security and Safeguards Activities . . . . . . . . . . . 34

S7.1 Review of Annual Security Program Audit . . . . . . . . . . . . . . . . . 34

V. M a n a g e m e n t M e e ting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35

X1

Exit Meeting Summ a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35

X 1.1 Routine Resident inspection Exit Meeting . . . . . . . . . . . . . . . . . 35

X1.2 Radiological Control Inspection Exit Meeting . . . . . . . . . . . . . . . 35

X1.3 Emergency Preparedness inspection Exit Meeting

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X3

Management Meeting Summary

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X3.1 Independent Safety Assessment Team Exit Meeting

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PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36

INSPECTION PRO CED URES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37

LIST OF ACRONYMS USED

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Report Details

Summary of Plant Status

Maine Yankee began this inspection period at 82% power. This was due to a power

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reduction on September 10,1996, when a problem with main circulating water pump,

P-6D, bearing cooling water flow was identified. The plant returned to 90.3% power on

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September 17, after completion of repairs to the pump. On October 9,1996, the plant

tripped from 90.3% power during reactor trip breaker surveillance testing. The reactor was

re-started on October 12,1996, and the plant returned to 90.3% power on October 14,

1996, and remained there for the remainder of the inspection period.

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I. Operations

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Conduct of Operations

01.1 General Comments (71707)

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Using Inspection procedure 71707, the inspectors conducted reviews of ongoing

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plant operations. The operators performed all observed tasks in accordance with

station procedures with a very good safety perspective in evidence. The reduction in

station power to investigate the bearing cooling water problem with the circulating

water pump was well controlled. In addition, station operators responded very well

to the unplanned reactor trip during reactor trip breaker surveillance testing.

Operator performame, when challenged by material condition problems, was good.

In particular, Plant Shift Superintendent (PSS) questioning attitude led to the

identification of a material condition problem (Section M2.3).

01.2 Plant Trio Durina Surveillance Testina

a.

Inspection Scope (92901)

The inspector observed the on-shift operations crew during the recovery phase of a

plant trip in the control room. In addition, the inspector reviewed the Post Trip

Review Team report which documented the event and attended the station plant

operations review committee (PORC) meeting when the report was reviewed. The

inspector also observed the instrument and controls (l&C) technicians during the

retest of the reactor trip breakers after troubleshooting and repairs were completed.

b.

Observations and Findinas

On October 9,1996, The reactor automatically tripped off line from 90% power. At

the time of the trip, station l&C personnel were conducting routine monthly

surveillance testing of the reactor trip breakers in accordance with station procedure

3-6.2.2.11, " Logic Relays Trip Test". The trip was uncomplicated and all control

rods fully inserted into the reactor core. Both emergency feedwater pumps

automatically started on low steam generator level and one of the two non-safety

electric motor driven feedwater pumps automatically started on low feedwater

header pressure (the other feedwater pump was isolated from service as expected).

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When steam generator water level stabilized, both emergency feedwater pumps were

stopped. The inspector observed operator performance and equipment response

from the control room. Operator performed well and maintained the plant safely. All

safety related systems responded as expected. There was no need for a safety

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injection actuation and none occurred.

The computer sequence of events log was printed out for the event and is listed

below:

• Reactor Trip Breakers Open

• Reactor Trip Computer Alarm (PTID 3809)

• Relay Actuation for Turbine Trip (20 ET)

• Relay Actuation for Backup Turbine Trip (20 AST)

• Turbine Valves Shut

• Reactor Trip (loss of load, Thermal Margin / Low Pressure,

• Low Steam Generator Level)

The normal sequence of events for a reactor trip is listed below:

• Reactor Trip (Due to a normal initiating signal)

• Reactor Trip Breakers Open

• Reactor Trip Computer Alarm (PTID 3809)

• Relay Actuation for Turbine Trip (20 ET)

• Relay Actuation for Backup Turbine Trip (20 AST)

• Turbine Valves Shut

The above information indicated that the reactor trip initiating event was an

inadvertent opening of the reactor trip breake"s without a reactor trip signal.

In response to the reactor trip, Maine Yankee station management directed that a

Post Trip Review Team be assembled to evaluate the event and provide

recommendations for corrective actions. The inspector noted that the team

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demonstrated a good questioning attitude and a good safety focus. The team

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determined that the most probable cause for the trip was a loss of test power voltage

during performance of reactor protection system (RPS) logic trip relay testing.

The inspector also observed maintenance activities being conducted to determine the

cause of the trip. Troubleshooting by I&C personnel did not identify any failed or

degraded components. Simulation of a failed test power supply gave the identical

indications that were observed during the reactor trip. As a precautionary measure

l&C technicians replaced the RPS trip matrix test power supply and the test hold

push button which were part of the circuit undergoing testing at the time of the

event. Also, prior to performance of the next routine monthly surveillance test,

Maine Yankee plans to replace the matrix relay trip select switch. During the next

monthly surveillance test, the test power supply voltage will be constantly monitored

to ensure that the voltage remains stable during testing or the testing will be

terminated.

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On October 12,1996, the plant was restarted after plant management reviewed and

approved the findings and conclusions of the Post Trip Review Team report. This

included a review of the completed maintenance and surveillance testing on the

equipment that was undergoing testing at the time of the plant trip.

c.

Conclusions

Operators responded well to the unplanned reactor trip and properly completed the

steps of emergency operating procedure E-0, Reactor Trip or Safety injection. The

Shift Operating Supervisor displayed excellent command and control of the operating

crew during the period of observation. The Post Trip Review Team provided a very

detailed report with an excellent basis for the conclusions that were drawn and the

recommended corrective actions (short and long term). Station management

provided good oversight and direction to ensure that all the necessary corrective

actions were completed prior to plant restart.

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Operational Status of Facilities and Equipment

02.1 Primarv Vent Stack Samolina Filters Not Installed (VIO 50-309/96-12-01 V)

a.

Inspection Scope (71707)

The inspector reviewed the event involving the primary vent stack filters, the short

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term corrective actions and discussed the chronology of the event with operations

department personnel. This inspection was to assess the licensee's actions to

determine the appropriateness and regulatory compliance.

b.

Observations and findinas

On October 1,1996, the primary vent stack (PVS) high range particulate, gas,

halogen, and particulate filters were taken out of service to allow station instrument

and controls technicians to repair the PVS air particulate detector (APD). Alternate

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sampling was invoked to compensate for the loss of halogen and particulate filters.

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This consisted of removing the normal halogen and particulate filters and operating

an alternate filter cartridge. The normal station practice is to remove the filters from

the normal cartridge and reinstall them in the alternate cartridge,

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The normal sampling system was returned to service on October 4,1996. The

station radiochemist, upon being informed that the normal sampling system was back

in service, performed an independent verification of the filter arrangement and found

that the iodine and particulate filter cartridges were not installed. The filters were

immediately installed in the normal PVS sampling system and the system was

returned to normal operation.

The plant operations department investigated the problem and determined that

station procedure 1-12-8, " Primary Vent Stack Air Particulate and Gas Monitor"

section 6.3, Auxiliary PVS Sampling System, did not specifically direct operators to

ensure that the halogen sampler cartridge and a particulate filter paper disk were

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installed in the normal sampling system prior to operation. A temporary procedure

change (TPC 96-328) was initiated to revise procedure 1-12-8 to resolve this

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problem prior to the system configuration being changed again.

Maine Yankee personnel disabled the Primary Vent Stack Air Particulate and Gas

Monitoring System by not installing the filter cartridge for a period of approximately

one hour due to a deficient procedure and personnel error. The f ailure to install the

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Primary Vent Filter is a violation Station T.S. 5.8.a.3. Technical Specification 5.8.a.3

requires in part that monitoring, sampling and analysis of radioactive gaseous effluent

is in accordance with 10 CFR 20.106 and with the methodology and parameters in

the ODCM (VIO 50-309/96-12-01).

At the completion of the inspection the licensee revised the station procedure

1-12-8, " Primary Vent Stack Air Particulate and Gas Monitor" and trained all

operations personnel as to the requirements of the procedure change.

c.

Conclusion

The inspector identified a violation of the station Off-Site Dose Calculation Manual

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and T.S. 5.8.a.3 when the Primary Vent Stack Air Particulate an Gas Monitoring

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System filters were not installed as required during conduct of maintenance activity.

02.2 Loss of Plant Computer

a.

Insoection Scope (71707)

The inspector reviewad the licensee's actions and processes to deal with the loss of

the plant computer on October 22,1996.

b.

Observations and Findinas

During the daily control room observation on October 22,1996, the inspector

observed operators following a problem with the computer not being available.

Inquiries revealed that the computer had failed and that the loss of the computer also

involved a loss of the Safety Parameter Display System (SPDS) in the Control Room

and the Emergency Centers. Also, the loss involved a loss of the on-line incore

analysis (INCA) program.

The computer system consists of two parallel front end systems (FES) for data

collection which input into two parallel operational support system (POSS) which

then input into two parallel emergency supply systems (ESS). The inputs into the

SPDS and the Emergency Response Data System (ERDS) is provided by the ESS.

The on-line INCA program is provided by the OSS. Three data gathering channels

(CH20, CH22 and CH24) provide data into each of two FES.

The loss of SPDS for over a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was reported to the NRC in accordance

with 10 CFR 50.72(B)(1)(V) for a major loss of emergency assessment capability.

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The Emergency Notification System (ENS) phone to the NRC was available as backup

to the ERDS. Plant personnel continued to perform incore analysis as required by

TS 3.10.C, Power Distribution Limits, without the plant computer. Operators also

began performing Precedure 1-4-1, Plant Operations Without the Plant Computer.

The inspector verified that plant information and parameters continued to be available

to operators who maintained safe operation of the plant. The loss of the computer

had no effect on the safe operation of the plant. The loss would have been a

challenge to operators during emergency response because the data on the computer

would not have been readily available. However, all the necessary information would

still be available on the various indicators and recorders.

The licensee's computer department personnel identified several failed data

acquisition controlled cards and a failed power supply during troubleshooting

activities. These failed components were replaced and subsequently tested

satisfactory.

Also, there appeared to be some inadequacies in the control of parts especially

during the change of vendors in early 1992. Two versions (Revision 5 and 6) of

power supplies for the file in the FES had been used when the update (Revision 5

and 6) had been inadequately communicated to the new vendor. The wrong version

of power supply reflects a communication problem, but it is not clear it led to the

failure of the power supply (UOR No.96-101).

c.

Conclusion

Troubleshooting efforts by computer department personnel were good. Operators

responded well to the problem and completed the required actions to ensure that loss

of the support functions provided by the computer did not affect the safe operation

of the plant. However, the inspector noted some weaknesses in the control of the

plant computer system.

The computer was repaired and restored to service on October 23,1996, and has

remained operational since.

02.3 Sourious Trio of Reactor Protection System, Channel C

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a.

Insoection Scone (71707. 62707)

On September 27,1996, a spurious trip of the reactor protection system (RPS)

Channel C; High Power, Symmetric Offset, and Thermal Margin Low Pressure trips

occurred. The inspector observed control room and troubleshooting and repair

activities following the channel trip.

b.

Observations and Findinas

During a control room tour on September 27, the inspector observed that three

channel C trip bins on the RPS panel were lit. The trips were as stated above. The

inspector verified that the corresponding annunciator panel alarms for these trips

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were illuminated. The annunciators were R3-4, Power Dependant insertion Lo-Lo

Limit Any Group and R3-11, Hi Power Level Channel C. The inspector reviewed the

actions discussed in Alarm Response Procedure (ARP) 2-37.R to be taken in response

to the alarms. The inspector also reviewed the current plant conditions and

parameters; and ascertained, that despite the channel C trips, an actual plant trip

was not required. The plant was operating steadily at 90% power with all plant

parameters normal. Allindications on RPS channels A, B and D remained normal.

Operators responded well and promptly to the failed channel C trips. They entered

the remedial action of TS 3.9.A.1 which required that an inoperable trip channel be

restored to operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Instrument and Controls (l&C) department was

notified of the problem and personnel from that department began troubleshooting.

At the time, the failed channel trips were declared inoperable for troubleshooting

maintenance and operators entered the exception staternent number 1 for TS 3.9.A

and exited the remedial action. The exception allowed a channel to be inoperative

for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for troubleshooting maintenance.

Later on, l&C personnel were able to identify a failed 18 volt power supply in

Channel C Symmetric Offset Trip Calculator. The power supply was replaced, the

trips cleared, and I&C surveillance procedures for Power Range Safety Channels,

Thermal Margin / Low Pressurizer Pressure, and Symmetric Offset were successfully

completed. With the RPS restored to normal, operators exited the exception for

TS 3.9.A on the same day.

The inspector observed and reviewed operators and l&C personnel actions during this

event and determined that they were appropriate. The inspector also per.'ormed a

walkdown of the control room panels including the RPS after the event and verified

that the RPS was in operational condition and that plant parameters were normal. No

discrepancy was identified.

c.

Conclusions

Operators performed well and took safe and appropriate actions in response to the

failed RPS channel C trips. l&C personnel responded timely and performed well by

safely and quickly diagnosing the problem and correcting it. Plant personnel showed

good safety focus and ensured that the plant continued to be operated safely while

the troubleshooting and repair efforts were on-going.

04

Operator Knowledge and Performance

04.1 Unolanned Reactor Power increase Durina Delithiation (URI 50-309/96-12-02)

a.

Inspection Scope (71707)

The inspector reviewed the circumstances leading to an unplanned power increase to

2457 MWth from 2440 MWth during a delithiation of the RCS, using purification

demineralizer,1-28.

.

.

7

b.

Observations and findinas

During a control room observation on September 23,1996, the inspector observed

operators taking actions to maintain the plant in a steady condition. Earlier, they had

been conducting a delithiation of the RCS when a power excursion had occurred.

The delithiation process involves taking the inservice demineralizer ott of service

while placing another fresh demineralizer in service. The demineralizer placed in

service could be previously borated to RCS boron concentration or not. A

predetermined quantity of RCS would then be flushed through the fresh demineralizer

to the primary drain tank (PDT) while a pre-calculated blended makeup to the volume

control tank (VCT) would be provided for making up the inventory in the RCS.

Operators had determined that the delithiation would be 8,000 gallons on this day.

Procedure 1-11-3, CVCS Filter and Demineralizer Operation, contains the instructions

for accomplishing the evolution. On this day, operators had removed 1-2C from

service and placed I-2B in service. Since 1-2B was considered ti

a unborated,

operators were performing section 5.2 (placing an unborated det. .seralizer in service)

of procedure 1-11-3. A power decrease was observed and attempts to restore

power to 2440 MWth resulted in a power increase to greater than 2450 MWth. The

approximate sequence of events was as follows:

initiation of delithiation by placing 1-2B in service and isolating 1-2C.1-2B was

e

being flushed with RCS to the PDT. Blended makeup was being added to the

VCT.

!

Power decrease (to about 2415 MWth) observed.

l

Operators attempted to restore power by adding primary water to the VCT.

Power increase observed and operator attempted to compensate by borating

and later by inserting rods (a total of 8 steps). But a peak power level of

about 2457 MWth was reached before power decrease was attained.

A power oscillation to another low of about 2423 and a high of about 2445

MWth occurred before the steady state 2440 MWth could be attained.

Delithiation was completed successfully.

The inspector observed most of the activities during this event from the control

room. Operators performed wellin trying to recover from the power oscillation. Shift

supervision maintained excellent command and control of activities. The plant shift

superintendent (PSS) and the shift operating supervisor (SOS) discussed and

analyzed the information being provided by the operator calmly and methodically

before the SOS provided directions to the operator. The crew performed very well

during the event.

The inspector's review of recent RCS delithiations showed that such a magnituda of

power increase was unusual and not expected. It appeared to be due to weak

performance by the operators. The inspector independently calculated the ratio of

..

.- . _-

-_

- -

-

- - - . - . _ - . - -

- - -

_ - .

.

Q

.

i

8

!

primary water to boric acid that, based on RCS boron concentration, would have

'

been required for blended makeup per procedure 1-11-3. The calculated ratio

corresponded to that which operators had determined earlier, and so would have

been using the correct blended makeup to the VCT. The licensee is conducting a

!

>

i

root cause of the event.

The inspector reviewed procedure 1-11-3, that was being used for the delithiation,

and portions of Procedure 1-4, Operations at Power, Section 6.3.2 of Procedure 1-4

requires that a power reduction to less than 2440 MWt shall be initiated if the

,

instantaneous power exceeds 2450 MWt. Operators complied with this requirement.

The hourly average power over a period of four hours covering the event were less

than or equal to 2440 MWt.

,

c.

Conclusions

While operators performed well to restore the reactor to a steady power level of

>'

2440 MWt, it appeared that performance weaknesses contributed to the power

excursion occurring in the first case. The licensee's root cause analysis was still

'

on-going and the appropriateness of implemented corrective actions still needs to be

verified. Therefore, this item remains open pending completion of licensee's root

cause analysis and the NRC's review of the results and the appropriateness of

instituted corrective actions (URI 50-309/96-12-02).

,

08

Miscellaneous Operations issues

08.1 Plant Operations Review Committee

a.

Inspection Scope (71707)

On September 20,1996, the inspector attended a Plant Operations Review

Committee (PORC) meeting,

b.

Observations, Findinas and Conclusions

I

The meeting was conducted in accordance with the requirements of TS 5.5.A, Plant

Operations Review Committee. The proper quorum was present. The issues were

well discussed with good safety focus evident, and personnel were available to

provide detaile on the issues and to answer questions raised by the PORC members.

l

Some of the issues discussed were: Design Basis Screen (DBS) No.96-060,

j

Containment Spray Pump NPSH; Event Review Board Report (ERB-013), HPSI Pump

P-14A Auto Start Severed Wire and Design Basis Screen No.96-059, DG-1 A(B)

Integral and Day Tank Capacity. A member of the Event Review Board for the HPSI

l

Pump Severed Wire presented the board's findings at the PORC meeting. The

I

presentation was clear, logical and concise. The inadequacies identified by the board

l

were discussed such that if any immediate corrective actions were needed, they

,

could be implemented.

1

.

.

9

The inspector concluded that the PORC meeting was conducted well and safety

focused. PORC continued to demonstrate good technical and safety perspective in

'

addressing plant issues.

11. Maintenance

M1

Conduct of Maintenance

M 1.1 General Activities

a.

Insoection Scope (62707)

The inspector observed maintenance troubleshooting and repair activities for the RPS

trip breakers channel failure, containment radiation monitors and LS-M-11 valve

repair and selected portions of the following activities.

WO 96-02769-00, Perform PM-26-M-E, Startup Check an J

Lube Oil Sample

WO-96-02761-00, Perform PM-26-SA-B, Check Oil in

Starting Air Lines and Check Lubrication Performance

WO 96-02760-00, Perform PM M-26-O-B, Drain Condensate

From Oil Cleaner and Fill Tank

WO 96-0344-00, Perform EM-26-2A-E DG-2 Panel Meter Testing

WO 96-03605-00, Troubleshooting of RPS System Matrix

Trip Circuit

b.

Observations, Findinas, Conclusions

The performance of these activities was observed to be very professional with

excellent diagnostic skills in evidence. Repairs were timely with excellent interaction

and cooperation between operations, radiation protection, and engineering

departments. The maintenance activities were conducted with good supervisory and

quality controls involvement. Workers followed procedures, had the workorders at

the work site and displayed excellent knowledge of the activities they were

conducting. Specifics of the observed work activities are described below:

M1.2 Troub!echootina of Reactor Protection System Matrix Trio Circuit

a.

Inspection # 70e (62707)

The inssee' r observed portions of the troubleshooting activities to identify the cause

of the

uvertent reactor trip that occurred during routine monthly surveillance

testing of the reactor protection system trip breakers,

b.

Observations and Findinas

The l&C technicians were very thorough and systematically tested all circuit

components to identify the source of the problem. However, they did not identify

l

.

l

.

10

any failed components during this activity. After discussions were held with

maintenance management, it was decided to replace all components of the test

l

circuit that was undergoing testing at the time of the plant trip. After replacement of

l

the components, l&C personnel performed surveillance test 3-6.2.2.11, " Logic

Relays Trip Test". The results were satisfactory with no problems identified.

l

c.

Conclusions

'

The inspector noted that I&C personnel conducted the troubleshooting activities in a

safe and controlled manner and properly tested all components in an attempt to

identify the cause of the failure. The decision to replace all the circuit components

showed a good safety perspective. Station maintenance management provided good

guidance to ensure that all possible avenues of repair were exercised prior to

completion of the repairs.

M1.3 Troubleshootina and Repair of Containment Radiation Monitor

a.

Insoection Scoce (62707)

The inspector reviewed the activities to troubleshoot and repair radiation monitor

]

RI-6113A, Containment High Range Radiation Monitor. (Se also section M2.2).

b.

Observations and Findinas

i

On September 26,1996, operators determined that the containment high range

radiation monitor, RI-6113A, was operating erratically. After comparing the observed

readings to those of the redundant channel (RI-6113B), which was operating

,

normally, the instrument was declared inoperable. A workorder was written for I&C

technicians to investigate and repair the radiation monitor. The l&C technicians

inspected and tested several components in the instrument and performed a

calibration check which was satisfactory. The instrument was returned to service for

trending only. On October 25,1996, the radiation monitor was returned to full

service after no further spiking was observed. In addition, a spare instrument channel

was ordered as a backup to reduce the unavailability time if a problem develops with

the instrument in the future.

c.

Conclusions

The inspector concluded that Maine Yankee promptly accessed the operability of the

radiation monitor and conservatively declared the channel inoperable. The l&C

technicians pro;;erly performed troubleshooting activities in accordance with the

station work control program. Although no particular problem was identified, the

decision to operate the radiation monitor for trending purposes only was appropriate

and after several weeks of operation with no further spiking, the equipment was

demonstrated to be stable. Also the decision to purchase a spare instrument to

reduce radiation monitor down time appeared to be sound.

i

.

l

.

11

M2

Maintenance and Material Condition of Facilities and Equipment

M2.1 Broken Handwheel on Valve LSI-M-11

a.

Inspection Scope (62707)

The inspector inspectud valve LSI-M-11, in the containment spray building while in

the as-found damaged condition, observed the handwheel removal, reviewed the

work package, and plant engineering department Technical Evaluation (TE). The

inspection was to determine if the maintenance department personnel properly

adhered to plant procedures and to verify the quality of the plant engineering

department assessment to resolve the problem.

b.

Observation and Findinas

On September 26,1996, a Maine Yankee Plant Shift Superintendent (PSS), during a

plant inspection tour of the containment spray building, observed that the handwheel

spokes for a motor operated valve LSI-M-11 were broken. This valve is a normally

open containment isolation valve and is in the low pressure safety injection line to

reactor coolant loop one. The PSS and the Shift Technical Advisor (STA) assessed

the problem and determined that the valve was operable because there was no

apparent damage to the valve itself, but only the handwheel. The valve was also

verified to be open which is the required safety function position. Maintenance

personnel were called in to verify the material condition of the valve (internal and

external). After completion of the evaluation, a decision was made to partially stroke

the valve. A partial stroke test manually (from the control room) was performed and

no malfunctions were observed and the valve was then re-opened.

Plant engineering department personnel generated a technical evaluation (TE) to

address the status (operability) of the valve. The inspector reviewed the TE and

determined that it included the proper design reviews, including a 10 CFR 50.59

screening which concluded that no full safety evaluation was needed. The failure of

the valve handwheel was evaluated to determine if there was risk to other

components on or nearby the valve. It was determined that only minor damage

could occur due to valve operation with a broken handwheel and that there were no

seismic concerns. The TE also included a determination that the handwheels on

valves LSI-M-11,21 and 31 were constructed of the same material and susceptible

to the same failure mechanism and should be removed. The failed welds were sent

to an outside testing facility to determine the failurc mechanism and at the close of

the inspection period the data was not available for review and the problems is being

tracked by a licensee Unusual Occurrence Report (UOR No. 96-87).

c.

Coriciusic ns

The inspector determined that the PSS displayed an excellent questioning attitude by

identifying the valve handwheel problem. The location of the valve handwheel was

not conducive to easy problem identification. The work package to perform the work

was properly performed by the maintenance mechanics and radiation protection

-.- .-

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_ - . - - . - - . . - - - . - .--- ... . - .

.

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l

12

1

personnel provided proper support to the field activities. The TE that was written to

l

provide engineering analysis and guidance was well written and comprehensive with

!

j

excellent supporting data.

J

f

M2.2 Operability of Containment High Radiation Monitor

!

a.

Inspection Scope (62707)

l

The operability of the containment high radiation monitor was questioned due to

^

erratic operation. The inspector reviewed Maine Yankee's troubleshooting and repair

l

activities to resolve the problem.-

j

b.

Observations and findinas

4

On September 26,1996, Maine Yankee operations personnel responded to main

I

control board panalarm RH-2-8 (Containment Radiation High). The operators noted

'

that control room indicator, RI-6113A, was reading approximately 10 r/hr but the

instrument needle was fluctuating and dropped to below zero r/hr as well. The

redundant channel (RI-6113B) was steady and reading normal (less than 2 r/hr).

After checking other containment instrumentation which were reading normal, the

operators concluded that Rl-6113A was inoperable and logged into the remedial

action of station technical specification 3.9C. This remedial action required that the

instrument be returned to service within seven days or a letter written to the NRC

j

delineating Maine Yankee's plans to effect repairs and return the instrument to

i

service.

)

A station workorder (WO 96-3456) was written to troubleshoot and repair the

)

instrument. However, the troubleshooting did not identify any failed component in

j

the instrument. Operations department management made a decision to operate the

1

instrument for trending purposes only to determine its reliability. This was performed

until October 25,1996, at which time the instrument was declared operable after no

I

further instances of spiking occurred. As a backup, another spare instrument was

i

ordered from the vendor to minimize the equipment downtime in the event of another

j

equipment problem.

I

Maine Yankee sent the technical specification required letter to the NRC within the

,

required time period. The letter, dated October 3,1996, discussed the plans for

'

resolving the problem. The letter stated in part, that the monitor was removed from

!-

'

service, inspected, cleaned and place back in service for trending purposes only.

j

Plant operations and I&C personnel continued to monitor the instrument closely for

signs of erratic operation to ensure the equipment is reliable for operation.

'

c.

Conclusion

Maine Yankee l&C technicians properly performed the troubleshooting as required by

<

j

the station work controls program. Maine Yankee personnel properly tested and

restored the instrument to service after the period of observations. The procurement

of another instrument as a spare was appropriate to prevent excessive periods of

'

downtime.

-

4

a

J

_ , .

_..

.

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-,,

-,

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.. _

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.

.

13

M2.3 Inocerable Pressurizer Proportional Heater Trains

a.

Inspection Scope (71707)

The inspector reviewed the circumstances that caused both trains of the Pressunzer

i

Proportional Heaters to be inoperable on October 18,1996. The inspection also

'

included a review of operator actions during the event.

b.

Observations and Findinas

On October 18,1996, with the reactor at 90% power, operators observed a

decreasing trend in reactor coolant system (RCS) pressure. A drop from 2230 to

2220 psig was noted. At that time, pressurizer proportional heaters E-2P-A and

E-2P-B were ir service as well as back-up heaters E-2-A and E-2-F.

There are 120 pressurizer heaters divided into eight groups with a total heating

,

capacity of 1500 Kw. Six of the eight groups (E-2A, E-2B, E-2C, E-2D, E-2E, and

E-2F) are backup heaters with fixed outputs, while the other two groups E-2P-A and

E-2P-B are proportional heaters whose output varies with the demand of the pressure

control program. Technical Specification Section 3.3.C.1 requires that at least one

j

bank of proportional heaters be operable during normal system operation whenever

i

the reactor coolant system Tavg is greater than 500 F.

During the event, once operators noted the pressure drop, they placed standby

backup heater, E-2-C, in service and the pressurizer pressure recovered to 2228 psig

i

and later to 2232 psig.

Electrical maintenance personnel conducted an investigation of the heaters electrical

j

circuit and discovered that there were blown fuses in phase A and phase C of

proportional heater E-2P-A. Operators then declared E-2P-A inoperable. Further

investigation by the electrical maintenance personnel revealed that both fuses on the

A phase of proportional heater E-2P-B were also blown. Operators then declared

E-2P-B inoperable. With both banks of proportional heaters inoperable, the Technical

'

Specification (TS) Limiting Condition for Operation (LCO) of section 3.3.C.1 was not

met and operators entered TS 3.0.A at 12:16 pm. Specification number 2 of

TS 3.0.A required a reactor shutdown within one hour. Meanwhile, maintenance

personnel were able to replace the blown fuse in E-2P-B. After a successful testing

of the heater, E-2P-B was declared operable and a reactor shutdown was averted.

Later, E-2P-A was also repaired, tested and declared operable.

Testing of the proportional heaters is normally accomplished per Attachment E,

Testing Pressurizer Proportional Heaters, of Procedure 1-1, Plant Heatup. The test

verifies power output from each bank of proportional heaters to ensure their

functionality. Electrical Maintenance also conducts a semi-annual Preventive

Maintenance (PM) for these heaters and had just completed one within the past

month and a half. The PM, E-11-SA-G(H), Check Heater Amperage, checks the

output of the proportional heaters under full load condition to verify functionality.

. . ~. .

.

. __

-.

- -

. - . . _ -

-

=

. . _ ._

.-

. . - -

-

1.

I

14

i

in addition, under the Supplemental Engineering Reliability Program, plant Engineering

Department personnel conduct a quarterly infrared surveillance of the heater circuitry.

j

Those tests failed to detect the fuse problem.

l

j

The inspector reviewed all applicable documents and noted that the licensee had

observed instances of blown fuses in the past in the proportional heater circuits.

However, no significant actions had been taken to thoroughly understand and correct

the problem. The licensee indicated that reviews were ongoing to develop a design

enhancement for the system. Meanwhile, the licensee stated that the electrical PM

'

would now be conducted at monthly intervals, and that the PED infrared checks

would be done on a weekly basis.

.

l

However, based on problems experienced in the past, the frequency of monitoring

j

the heaters was increased in response to this problem. In addition, extensive actions

to identify the root cause of the failed fuses to eliminate the problem was not

pursued.

c.

Conclusion

.

!

l

Operators responded well to the failure of both banks of proportional heaters.

Electrical maintenance personnel demonstrated excellent support to operations by

,'

timely identifying and resolving the problem, thereby preventing an unnecessary

transient to the plant. The licensee recognized a weakness in the process and what

appeared to be appropriate actions were planned and are being taken to control the

j

problem (UOR No. 96-99).

'

M2.4 Control Element Assembly Problems Durina Reactor Startuo

i

a.

Inspection Scope (62707)

The inspector observed plant restart activities in the control room and reviewed the

i

licensee's efforts to correct the problems encountered while withdrawing certain

control element assemblies (CEA) during reactor startup.

t

4

b.

Observations and Findings

I

During the reactor startup on October 10 - 11,1996, the inspector noticed some

discrepancies with the operation of the CEAs. There were instances of dropped

,

rods (similar to these experienced during the September 1,1996, reactor startup)

i

and instances of the reed switch CEA position indication not keeping up with the

pulses for CEA movement.

3

i

The significant problems were as follows:

I

I

Partial drop of Rod 4 (in Group 4)

Partial drop of Rod 4 (in Group 4)

3

Rod 2 reed switch indication not keeping up with the pulses

j

Partial drop of Rod 3 (in Group 4)

i

.-

.

.

..

-_=

--

. - - _ - . - .

.

. . .

.- .

_

-

-

l

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.

15

Following each rod drop, the affected group (Group 4) was re-inserted and realigned

in accordance with Abnormal Operating Procedure (AOP) 2-21, Misaligned (Dropped)

CEA, however, the situation presented a challenge'to the operators during their

attempt to restart the reactor. The restart was aborted and all previously withdrawn

regulating rods (groups 1,2, and 3) were re-inserted, so that extensive trouble

shooting efforts could be conducted to resolve the problems.

The inspector observed some of the troubleshooting activities and noted that they

were conducted properly. At the end of these activities, a modification to procedure

l

3-6.2.1.19, Rod Drop Time and Functional Check, was required so that it could be

-

used for exercising the group 4 rods, one at a time for functional testing purposes.

The modification was made under the temporary procedure change process

(TPC # 96-331). The inspector reviewed the approved copy of the change and

verified that it had been properly processed, and did not create a situation for

operating the reactor outside previously analyzed conditions. Operators had

l

conservatively assigned the previously calculated and approved boron concentration

l

required to maintain the reactor subcritical with group 4 withdrawn as that to be

!

maintained with a single CEA withdrawn. The inspector reviewed the calculations

l

with the shift technical advisor and verified that they were accurate and in

accordance with the numbers and charts provided in the Technical Data Book (TDB).

i

The repair activities included the following:

!

for Rod #2, replacement of the timer module, the upper gripper power switch

and the pull down power switch.

for Rod #3, replacement of the upper gripper power switch and the pull down

coil.

for Rod #4, replacement of the timer module, the upper gripper power switch,

and the pull down switch.

The CEAs functioned well following the repair and the reactor was successfully

restarted.

c.

Conclusions

Operators performed wellin response to the CEA problems. They followed the

appropriate procedures and maintained the plant in a safe condition. During and

following troubleshooting activities, plant personnel showed good safety perspective

and ensured that the plant was always maintained safely. However, the problems

with the CEAs were similar to those that occurred during reactor startup the previous

l

month and it appeared that the right solution of the previously identified problems

l

had not been made. These equipment malfunctions posed an unnecessary challenge

to the control room operators during their startup of the plant.

l

,

.

.

16

til. Enaineerina

E2

Engineering Support of Facilities and Equipment

E2.1 Desian Issues Affectina Cold Weather Operations in the Turbine Buildina (uodate

URI 50-309/96-08-02 and 50-309/96-08-04)

a.

Inspection scope (37551)

The inspectors reviewed and discussed the status of Maine Yankee's engineering

efforts to address the design issues affecting cold weather operations.

I

On October 21,1996, the inspector met with representatives frorn the corporate

engineering department (CED) to discuss the status of ongoing licensee activities to

address the concerns with the High Energy Line Break (HELB) and Flooding

Conditions in the turbine building. Specifically, the review was to understand and

assess Maine Yankee's turbine building modifications to ensure that safety related

components would continue to function during a design bases HELB or flood in the

turbine building. The review was also to determine if the modifications are suitable

j

for cold weather operation.

j

b.

Observations and Findinas

A HELB Condition could occur in the turbine building if a postulated rupture of high

energy lines in the building occurred. Originally, the limiting or worst case scenario

was postulated to be a break of a 30 inch main stream wire. However, recent

reviews have determined that a postulated slot break of the main feed wire could

present the worst case scenario because a higher energy at a lower pressure could

<

be released without the Turbine Building panels blowing open.

A flooding in the Turbine Building could occur if a postulated failure of non-safety

related water pipes in the Building occurred. Previous analyses performed to mitigate

the effect of the postulated flood had considered a guillotine break of the circulating

water system pipe as the worst case scenario. This break would result in a spill of

about 110,000 gpm. However, the flood protection saving panels installed were not

capable of mitigating a flood of 3,500 gpm resulting from a crack in the circulating

water would not provide enough static head to open the swing panels.

In NRC Inspection Reports 50-309/96-03 (Section 3.0) and 50-309/96-08

(Section E2.2 and E2.4, respectively), the issues involving flooding and HELB

concerns in the turbine building were discussed. As compensatory measures, Maine

Yankee had secured the roll-up doors open at least 5 inches to ensure flood relief

would be available. As for the HELB concern, the licensee had left some banks of

roof louvers open, removed some of the plywood covering from the walllouvers and

secured the building unit heaters which were to be used only as needed. These

compensatory measures were suitable for temporarily addressing the concerns and at

.

.

17

that time, they did not create any habitability or environmental concerns in the

turbine building. However, with the onset of cold weather conditions, the potential

habitability or environmental concerns needed to be resolved.

In a letter to the NRC, dated August 14,1996, Maine Yankee committed to

j

conducting an analysis of the postulated HELB condition with a " winter condition"

ventilation lineup. The analysis was to be submitted to the NRC by October 1,1996,

along with a description of any plant modifications which appear to be warranted. In

another letter to the NRC dated October 1,1996, Maine Yankee indicated the need

for an extension of the submittal date of the analysis.

j

The inspectors reviewed the licensee's close out plans which included the actions to

be taken to address each of tha issues. Engineering analysis were still ongoing to

determine the turbine building environment based on the worst case scenario for a

HELB in the turbine building. The licensee was considering a feedwater line slot

break as the worst case. A design discrepancy evaluation (DDE No. 96-63) was

generated to address this issue.

i

To address the flooding concerns, a design change was being generated to raise the

berm at the entrance to the corstrol room from 4 to 6 inches (present flooding in the

control room) and to install a gravity flap in the north wall of the turbine building

allowing water build-up to leave the turbine building. The inspectors will review the

i

details of each issue resolution when they become available. At this time, the

inspectors identified no safety concerns, with the licensee's activities and will

continuo to follow-up on these issues.

c.

Conclusions

Maine Yanken was making good efforts to address the turbine building design issues

affecting cot Neather operations. The activities of the corporate engineering and

licensing depa anent personnel appeared technically sound and detailed. There was

good management attention and involvement in addressing these issues.

IV Plant Suocort

R1

Radiological Protection and Chemistry (RP&C) Controls

R1.1 Multiple Contamination Events and Unolanned Exoosures

a.

Insoection Scope (71750)

The inspector reviewed the licensee's corrective actions concerning several

unplanned exposures during this and the previous inspection periods.

b.

Observations and Findinas

The first unplanned exposure was to an operations department radiation waste

handling operator. This event occurred during a routine filter changout of the

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Duratek System. The Duratek System is used by Maine Yankee to remove

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radioactive waste from waste liquid prior to disposal. The second unplanned

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exposure occurred to a contractor during repair activities on the auxiliary steam

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system in the primary auxiliary building. The loss of contamination control occurred

when a radioactively contaminated sheet of metal was taken outside the restricted

area, and returned to the station warehouse.

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On April 16,1996, the pre and post filters of the Duratek liquid waste ion exchange

demineralizer processing unit were changed out. A Maine Yankee operator and a

radiation protection technician were assigned to the task and no abnormal conditions

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were identified during the change out process.

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On July 11,1996, Maine Yankee radiation controls supervisory personnel reported

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that during a review of the second quarter TLD data it was identified that the station

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radiation waste handling operator had received an unplanned skin dose of 6.94 rem.

This is approximately 14% of the extremity annual limit. On July 12,1996, station

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management suspended the operator's access to the restricted area pending a dose

evaluation and investigation of the event.

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On July 16,1996, the preliminary results of the investigation were presented to

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station management and corrective actions were implemented to allow filter

changout to recommence on July 19,1996. The corrective actions included

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mandatory issuance of extremity dosimetry for filter changeouts, increased frequency

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of extremity dosimeter processing and increased personnel and area survey

requirements in potentially discrete (HOT) particle and highly contaminated areas.

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The inspector reviewed the completed corrective actions, held discussions with

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radiation protection technicians and supervisors and discussed the event with the

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radiation control manager. The corrective actions were very comprehensive in scope

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with excellent use of independent analysis to determine the source of the exposure.

The investigation was unable to pinpoint the exact cause of the exposure, but the

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most plausible scenario appeared to be that a discrete (Hot) particle was deposited

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on the workers rubber glove during the filter changeout process and caused the

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exposure. The inspector did note that one of the conclusions in the radiological

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incident report (RIR) was that "since the operator is not expected to recognize that a

failed gasket seal might change radiological conditions, he would have no reason to

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alert the technician to this fact." This statement does not appear to coincide with

stated Maine Yankee policy that operators and all radiation workers are expected to

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use a questioning attitude during all plant evolutions. With the bag filter being used

for the first time, extreme caution should have been required for the evolution. Also

with the seal being identified as being disturbed there appears to have been good

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reason to alert the technician to the potential of an abnormal condition in the filter

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housing. This problem was discussed at a management meeting in Region I on

August 9,1996 as documented in NRC Inspection Report No. 50-309/96-09.

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in another situation on October 1,1996, a Maine Yankee maintenance mechanic

brushed against a lead shield blanket and received an unplanned exposure of

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approximately 88 mrem from a discrete (hot) particle which lodged on the seat of

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the worker's coveralls. The particle swipe sample was analyzed by the radiation

protection department and determined to primarily consist of magnesium 54 and

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cobalt 60.

And yet in another situation on August 22,1996, Maine Yankee Radiation Protection

personnel reported to the NRC that several steel plates had been identified with

internal contamination. Upon further review the licensee determined that the

contamination was in fact external and milling activities would remove the

contamination. The steel plate was determined to have been used during the last

refueling outage, on the reactor coolant system decontamination project and

subsequently returned to the warehouse. On October 7,1996, Maine Yankee

notified the NRC to provide an update to the previous notification.

Maine Yankee is currently reviewing the station program that surveys material prior

to release from the station restricted area. The inspector will review the licensee's

finding in a subsequent inspection.

c.

Conclusions

The inspector determined that Maine Yankee radiation protection department properly

responded to the unplanned exposure events. The analysis of the data was thorough

with proper review by independent expertise. Most of the conclusions were

appropriate. However, the inspector noted that one of the conclusions for the

Duratek event did not properly take into consideration the experience and training of

the two individuals that were involved in the event. The inspector determined that

while no personnel exposures exceeded regulatory limits, the contamination exposure

controls were weak.

R1.2 Imolementation of Radioactive Liauid and Gaseous Effluent Control Proaram

a.

Inspection Scope (84750)

Inspection of this area consisted of:

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physical walkdown of facilities and equipment, including air cleaning

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systems;

review of selected licensee's procedures, and

review of selected radioactive liquid and gaseous discharge permits with

respect to TS ODCM requirements.

b.

Observations and Findinas

The inspector toured all effluent RMSs, several process RMSs, and the air cleaning

systems and noted the below findings:

All RMSs were operable at the time of this inspection.

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The background readings of effluent RMSs were reasonably low to allow

monitoring of any unusual releases

The differential pressure between the high efficiency particulate air (HEPA)

charcoal filters were within the licensee's acceptance criteria

The effluent control procedures were detailed, easy to follow, and ODCM

requirements were incorporated into the appropriate procedures.

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The discharge permits were complete and met the TS/ODCM requirements for

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sampling and analysis at the frequencies and lower limits of detection

established in the TS.

c.

Conclusion

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The licensee established, implemented, and maintained effective radioactive liquid

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and gaseous effluent control programs.

R2

Status of RP&C Facilities and Equipment

R2.1 Effluent / Process Radiation Monitorina Systems

a.

Insoection Scoce (84750)

The inspectors reviewed the most recent calibrated results, upgrades, and

workorders for the following effluent and process RMSs to determine the

implementation of the TS/ODCM requirements and the Updated Final Safety Analysis

Review (UFSAR) commitment for:

Liquid Radwaste Effluent Monitor

Service Water Effluent Line Monitor

Condenser Air Ejector Monitor

Steam Generator Blowdown Line Monitor

Plant Vent Stack Noble Gas Monitors (Normal and High Range)

Waste Gas Holdup System Monitor

Containment Noble Gas Monitor, and

Main Steam Line Monitors

b.

Observations and Findinas

The Instrumentation and Controls (l&C) Department had the responsibility of

performing electronic and radiological calibrations for the above effluent process

RMSs. The I&C system engineer had the responsibility to maintain the operability for

the above RMSs and upgrade the system, as necessary. All calibration results

reviewed were within the licensee's acceptance criteria.

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During the review of the above RMSs radiological calibration data, the inspector

independently verified several calibrated results, including linearity tests and

conversion factors. The comparison results between the licensee's values and

inspectors' calculation values were in agreement.

The licensee issued an RMS Manual, " Radiation Monitoring System (Design Basis

Summary Document)", in June,1996. The purpose of this manual was to document

the design basis of the effluent process RMSs installed at the site. The manual also

provided a means to verify and validate system reliability and functionality. The

inspectors reviewed this manual and noted that it contained the following useful

information for users (e.g., control room operators, chemistry staff, and l&C staff):

Functional Requirements and Capabilities

External Events Considered in the Design Basis for the System

Historical Narrative of Modifications and Analysis

Synopsis of System and Component Testing

Piping System Analysis of Record

Setpoint Sumw y and

Component Summary of Design Conditions

Several important RMS upgrades were completed since the last inspection of the

REMP. These upgrades included: addition of a polished sleeve for the liquid radwaste

effluent monitor, which helped to minimize radionuclide plateout; refurbishment of

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the three air particulate detector (APD) paper drives; and replacement of several self-

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rolling ribbon cables.

The inspectors' review of open and closed licensee workorders pertaining to the

RMSs indicated no recurring equipment problems other than recurring problems with

the self-rolling ribbon cables (which, as a consequence, were recently replaced as

need).

c.

Conclusion

Notwithstanding the RMSs calibration discrepancy noted in Section R7.1 of this

report, the RMSs were well maintained. Licensee completion of several RMS

upgrades and the RMS manual demonstrated the licensee's commitment to a strong

REMP.

R2.2 Air Cleanina Systems

a.

Insoection Scope (84750)

The inspector reviewed the licensee's most recent surveillance test results to

determine the implementation nf TS requirements and UFSAR commitments, and

walked down the following ventilation systems:

Spent Fuel Pool

Containment Ventilation / Purge

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Control Room (CR) Recirculation and CR Breathing Air, and

Primary Auxiliary Building

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The inspectors reviewed the following surveillance test results:

Visual Inspection

In-Place HEPA Leak Tests

In-Place Charcoal Leak Tests

Air Capacity Tests

Pressure Drop Tests, and

Laboratory Tests for the lodine Collection Efficiencies

b.

Observations and Findinas

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All test results of the above systems were within the licensee's acceptance criteria

with the exception of: 1) the laboratory test methodology for the iodine collection

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efficiency for the containment ventilation / purge system, and 2) tLe failure of the in-

place leak test for control room recirculation system.

Section 4.11.D,2.b of the TS, Containment Ventilation / Purge System, requires that

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radioactive elemental iodins be used as a challenging agent to determine the iodine

collection efficiency. In past surveillance, methyl iodide was used as the challenging

agent, rather than elemental radioiodine. All other air-cleaning systems require the

methyl iodide be used as a challenging agent to determine the iodine collection

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efficiencies. Although the use of elemental radiciodine as a challenging agent to

determine iodine collection efficiency is available (described is ASTM D-3803-79),

use of methyl iodide is the general industrial practice because it is readily available,

cost effective, and is considered to provide better test results. Isee Regulatory

Guides 1.58 (for post accident condition) and 1.140 (for normal ventilation), and

ANSI N510-1975 for details.] During an October 22,1996, telephone call with

licensee representatives, the inspectors were informed that the licensee had decided

to change TX 4.11.d,2.b to denote methyl iodide as the challenge agent. The

licensee committed that this TS change would be approved by PORC and Nuclear

Safety Audit and Review Committee (NSARC) and submitted to the NRC by

March 1997. The inspectors assessed the safety significance of this matter to be

negligible because elementaliodine can be used as a charcoal challenge agent. 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 Poliev.

The in-place halogenated hydrocarbon leak test for the control room recirculation

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system failed on October 31,1995. After replacing the charcoal filter gaskets ad

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repairing the filter frarne, the licensee retested on December 5,1995, and the result

was within the TS acceptance criteria.

c.

Conclusion

The licensee adequately implemented TS requirements except for one minor violation.

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R2.3 Ventilation Systems Air Balance (IFl 50-309/96-12-03)

a.

Inspection Scope (84750)

Section 9.13 of the final safety analysis report (FSAR) describes the general

ventilation systems and each building ventilation system throughout the plant. Air

balance for the Primary Auxiliary Building, Fuel Building, Containment Spray Pump

Area, and Service Building were reviewed.

b.

Observation and Findinas

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The inspectors discussed with the heating, ventilation, and air conditioning (HVAC)

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engineer, the maintenance and testing of the ventilation systems committed in

Chapter 9.13 of the FSAR. Section 9.13.2.5 of the FSAR, Turbine Building, details

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turbine building ventilation system supply and exhaust air flow rates.

Section 9.13.2.7 of the FSAR, Office Building, details office building ventilation

system supply and exhaust air flow rates. No surveillance requirements are detailed

in Sections 9.13.2.5 or 9.13.2.7 of the UFSAR. The inspectors noted that there

were no routine surveillance tests documented for the Turbine and the Office

Buildings. Such tests are not specifically required by the technical specifications.

The independent safety assessment team identified a number of examples of weak

and inadequate testing indicating a broad and programmatic concern and this area is

pending further review (See also, NRC Inspection Report No. 50-309/96-09,

Section E2.1).

The inspector noted that the Service Building was divided into two areas, a

contaminated area and a clean area. The licensee maintained a positive pressure for

the clean area and a negative pressure for the contaminated area. Section 9.13.2.4

of the FSAR, Service Building, denotes that HV-8 removed 6,200 cfm from the

Service Building contaminated area in order to maintain a negative pressure. The

inspectors determined that this fan should have been denoted as FN-8 (an exhaust

fan) in the UFSAR. The fan HV-8 (5,000 cfm) was an air supply fan. The licensee

acknowledged the error and stated that the UFSAR would be revised accordingly.

c.

Conclusion

Overall, maintenance and surveillance of the ventilation systems was adequate. The

licensee had not documented the plant air balance surveillance results to demonstrate

consistency with Chapter 9.13 of the FSAR assumptions. Section 9.13.2.4 of the

FSAR had not been updated. These items will be reviewed during a subsequent

inspection (IFl 50-309/96-12-03).

R3

RP&C Procedures and Documentation

R3.1 Off Site Dose Calculation Manual Review

a.

Inspection Scooe (84750)

The inspection reviewed the ODCM implemented at the Maine Yankee site, including:

1) dose factors,2) setpoint calculation methodology, and 3) bioaccumulation factors

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for aquatic sample media. The inspectors also reviewed the 1995 Annual

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Radioactive Effluent Report.

b.

Observations and Findinas

The ODCM provided descriptions of the sampling and analysis programs, which are

established for quantifying radioactive liquid and gaseous effluent concentrations,

and for calculating projected doses to the public. All necessary parameters, such as

effluent radiation monitor setpoint calculation methodologies, site-specific dilution

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factors, and dose factors, were listed in the ODCM. The licensee adopted other

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necessary parameters from Regulatory Guide 1.109.

The inspectors reviewed the 1995 annual radioactive effluent release report. This

report provided data indicating total released radioactivity for liquid and gaseous

effluent. This annual report also summarized the assessment of the projected

maximum individual and population doses resulting from routine radioactive airborne

and liquid effluent. Projected doses to the public were well below the TS limits. The

inspectors determined that there were no anomalous measurements, omissions or

adverse trends in the report.

c.

Conclusion

Based on the above review, the inspectors determined that the licensee's ODCM

contained sufficient specification, information, and instruction to implement and

maintain the radioactive liquid and gaseous effluent control programs. The

inspectors also determined that the content of the Annual Report was very good and

met the TS reporting requirements.

R6

RP&C Organization and Administration

R6.1 Radioactive Liauid and Gaseous Effluent Proaram Review

a.

Insoection Scope (84750)

The inspectors reviewed changes to the organization and administration of the

radioactive liquid and gaseous effluent control programs and the REMP.

b.

Observations and Findinas

The inspectors determined that there had been no changes since the last inspection

conducted from May 1-5,1995. The Chemistry staff had primary responsibility for

conducting the radioactive liquid and gaseous effluent control programs. The

departments of Operations, System Engineering, Radwaste Operations and l&C

support the radiological effluent control programs relative to air cleaning systems,

radioactive liquid discharge, and radiation monitoring system calibrations.

The licensee reorganized the REMP in September 1996. The REMP moved from the

Radiation Protection Section (Dosimetry) to the cognizance of the Environment Health

and Safety and Emergency Preparedness (EMS & EP) section. The inspectors

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interviewed the EMS & EP section head regarding the implementation of the REMP.

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The inspectors assessed that this change was an improvement in oversight because

many technical aspects of EP and the REMP are similar in nature. The former Maine

Yankee chemistry section head, who had very good knowledge in the areas of

effluent controls and the REMP, was assigned to oversee the environmental specialist

who was assigned the REMP. An individual from another nuclear power plant was

recently assigned as the acting chemistry section head for one year.

c.

Conclusion

Oversight of the REMP was improved. No degradation was noted as a result of the

reorganization.

R7

Quality Assurance in RP&C Activities

R7.1 Quality Assurance Audit Report Reviga

a.

Inspection Scoce (84750)

The inspection consisted of a review of Quality Assurance (QA) Audit Reports

required by the TS and a review of corrective actions implemented to address audit

findings. The inspector reviewed the 1995 QA Audit Report No., MY-95-02,

" Chemistry / Radiological Effluent Technical Specifications (RETS)/Off-Site Dose

Calculation Manual (ODCM)", and the 1996 Audit Report No., MY-96-02, " Chemistry

Audit". These audits were conducted by the Quality Programs Department,

b.

Observation and Findinas

The inspectors noted that individuals with appropriate technical expertise were used

to assist the audit team leader, which was considered to be a strength.

The 1996 audit findings, focused mainly on administrative aspects of the Chemistry

program. No " technical" issues of regulatory significance were identified.

The 1995 audit identified that the primary vent gas monitor RM-3902Y had not been

calibrated at the frequency required by TS 4.1. No calibration records were found for

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the time period 9/92 to 3/95. TS required this instrument to be calibrated once every

18 months. The tracking system for surveillance had been established to key on

individual procedures rather than on an individual RMS. RMS calibration procedures

were established so that a single procedure would provide calibration guidance for

several RMSs of the same manufacturer and model. l&C supervisory review also

failed to initiate a timely calibration of RM-3902Y. Licensee corrective actions

included (1) calibrating RM-3902y and (2) improving the administrative controls for

tracking of RMS calibration / surveillance due dates by tracking by RMS channel

numbers rather than by RMS calibration procedure numbers. The inspectors

considered the corrective actions to be appropriate. The inspectors' assessment of

the safety significance of the failure to calibrate RM-3902Y at the frequency required

by TS 4.1 was minimal due to the fact that (1) daily grab samples were taken (due to

a pre-existing agreement with the State of Maine) and (2) the March,1995,

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calibration of RM-3902Y demonstrated no significant change in the calibration factor.

This licensee-identified and corrected violation is being treated as a Non-Cited

Violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev.

The inspectors noted that the failure to calibrate RM-3902Y was not formally

communicated to be NRC by either submittal of an LER or annotating the Annual

Radioactive Effluent Release Report. The inspectors questioned licensee personnel as

to the basis of the reportability determination that no LER was required for this

matter and found the determination adequate. The inspectors also discussed with

the licensee a means of notifying the NRC of future cases of RMS inoperability such

as by writing as LER or by noting such cases in the Annual Radioactive Effluent

Release Report. Licensee representatives noted to the inspectors that a

February 1996 revision to section 2.3.3 of the ODCM dictated that if an RMS could

not be returned to an operable status within 30 days, an explanation of the delay

was to be provided in the next Annual Radioactive Effluent Release Report. The

inspectors had no further questions regarding how the licensee would notify the NRC

of future cases of an inoperable RMS.

c.

Conclusion

Based on the above reviews, the inspectors determined that the licensee met the QA

audit requirements. Licensee corrective actions to audit findings were considered

appropriate.

R8

Miscellaneous RP&C lssues

R8.1 Review of Updated Final Safety Analysis Reoort (UFSAR) Commitments

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A recent discovery of a licensee operating their facility in a manner contrary to the

UFSAR description highlighted the need for a special focused review that compares

plant practices, procedures and/or parameters to the UFSAR description.

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While performing the inspection discussed in this report, the inspectors reviewed the

applicable portions of the FSAR that related to the areas inspected. The following

inconsistency was noted between the wording of the FSAR and the plant practices,

procedures and/or parameters observed by the inspectors.

Section 9.13.2.4 of the FSAR noted that HV-8 was an exhaust fan and was designed

to maintain a negative pressure in portions of the Service Building. The inspectors

determined that this fan should have been denoted as FN-8 (exhaust fan) in the

FSAR. The f an HV-8 was an air supply fan. This FSAR discrepancy was considered

to be minor in nature and is discussed in more detail in Section R2.3 of this report.

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P1

Conduct of Emergency Preparedness (EP) Activities

P1.1 Emeraency Plannina Self Assessment Proaram and Corrective Action Trackina

System

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a.

Insoection Scope (82701)

The inspector reviewed the licensee's action item tracking system and the

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Emergency Planning self-assessment program to determine the effectiveness

of licensee controls.

b.

Observations and Findinas

The inspector reviewed the corrective action tracking system (CATS) for EP items.

There were 14 open items on CATS. Most of the open items pertained to exercise

and drillissues. One long range item being tracked was the possible addition of

another fiber-optic telephone line into the site from another area. This line is being

contemplated to preclude the loss of long distance capabilities and the NRC systems,

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which occurred in June of 1995 when the AT&T fiber-optics line in Freeport, ME was

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damaged.

Additionally, the inspector reviewed five self-assessments of different areas of the

emergency preparedness program. These self-assessments were performed on

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emergency preparedness equipment inventories, exercise development review,

training in the use of portal monitors, offsite scenario interface, and monthly

communication drills. The self-assessment of monthly communication drills identified

that the February drill was missed (see Section P2). The inspector noted that the

other self-assessments adequately identified additional areas where improvements

could be made in the emergency preparedness program, and that they were being

actively pursued.

The EP items were being properly tracked. The closure of EP items required the

approval of the Environmental Health and Safety / Emergency Preparedness Section

Head.

c.

Conclusions

Both the CATS and the self assessment program appeared to be an effective licensee

control.

P2

Status of EP Facilities, Equipment, and Resources

P2.1 Emeraency Plannina Eauipment inventories and Surveillance

a.

Insoection Scone (82701)

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The inspector reviewed facility equipment inventories and surveillance conducted

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from January 1996, through September 1996, for completeness and accuracy. The

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inspector also conducted an audit of emergency equipment in the Control Room,

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Operations Support Center (OSC), Technical Support Center (TSC), Emergency

Operations Facility (EOF) to determine if any changes had been made and whether

the facilities were operationally ready.

b.

Observations and Findinas

The inspector noted that the inventories were thorough and that items found

depleted were immediately replaced and equipment found inoperable was replaced

with back up equipment, and it was noted on the inventory the equipment was out

for repair. All dosimetry had been exchanged in the time frame required by the

emergency plan. All survey instruments were calibrated on the same date instead of

on various dates throughout the six month calibration period as had been the

previous practice.

While reviewing the communication drill surveillance, the inspector noted that the

February 1996 notification drill was not conducted. This missed notification drill had

been identified by the licensee during a July self-assessment. The licensee found

that the request to conduct the drill was apparently lost and the follow up to ensure

its conduct had not occurred. The NRC Resident inspectors had been informed of

the missed notification drill on July 3,1996. The inspector determined that all other

surveillance were completed in accordance with the emergency plan with the

exception of the TSC, OSC and EOF ventilation discussed in Section P8.2.

During an inspection tour of the control room, TSC, OSC, and EOF, the inspector

conducted a selected inventory of the equipment lockers. The equipment was found

to be operationally ready and within calibration, as required by the emergency plan.

The inspector also noted that the licensee's potassium iodide supply had an

expiration date of 2000. The inspector tested the emergency notification system

telephone line from the EOF to the NRC Operations Center and found it in working

order.

c.

Conclusions

Facility inventories were complete, radiation survey instrumentation was within the

calibration requirements and the emergency response facilities were found to be in a

state of operational readiness.

P3

EP Procedures and Documentation

P3.1 Review of Emeraency Response Plan Chanaes (Closed. URI 50-309/96-007-01)

a.

Inspection Scope (82701)

The inspector reviewed recent emergency response plan changes to assess the

impact on the effectiveness of the EP program. The inspectors also assessed the

process that the licensee uses to review emergency plan changes.

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b.

Observations and Findinas

The inspector reviewed emergency plan change 96-01 and 96-02 and the process

used to evaluate the changes to determine if they met the requirements of

10 CFR 50.54(q) and did not decrease the effectiveness of the emergency plan.

Change 96-01 revised Table 6-1 giving the functional responsibility for rescue

operations and first aid to the Security Supervisor instead of the Plant Shift

Supervisor (PSS). Change 96-02 added the 60 minute goal of staffing the

emergency response facilities to Section 6.1 of the emergency plan. This change

addressed the NRC discrepancy that was identified during the licensee's graded

exercise June 1996. URI 50-309/96-007-01 is closed.

c.

Conclusions

Based on the licensee's determination that the changes did not decrease the overall

effectiveness of the emergency plan and after limited review of the changes by the

inspector, in accordance with 10 CFR 50.54(q), no NRC approval is required to

implement the changes. However, implementation of these changes may be subject

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to further inspection at a later time to confirm that the changes have not decreased

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the overall effectiveness of your emergency plan.

P5

Staff Training and Qualification in EP

PS.1

-meraency Plannina Trainina Proaram Evaluation

a.

inspection Scone (82701)

The inspector reviewed EP training records, training procedures, lesson plans,

emergency plan, and EPIPs associated with on-shift dose assessment to evaluate the

licensee's EP training program and obtain information for NRC's Temporary

Instruction 2515/134 " LICENSEE ON-SHIFT DOSE ASSESSMENT CAPABILITIES."

b.

Observations and Findinas

The inspector reviewed thirty percent of the emergency response organization

training records. Additionally, the inspector reviewed the dose assessment training

for operations personnel, emergency directors, and radiation protection. The

inspector found that all training was current.

The inspector was given a demonstration of the offsite dose projection system

(ODPS)in the control room by the PSS and the shift technical advisor (STA). The

demonstration included obtaining the meteorological and radiological monitoring data,

entering it, in the correct format, into the ODPS system, obtaining the results, and

printing it out. In addition, emergency response organization (ERO) management, and

the radiation protection staff are trained in the use of ODPS, the Meteorological Post

Accident Computer Model (METPAC) computer dose projection system, and the

backup nomogram.

The ERO is staffed three deep in all major positions.

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c.

Conclusion

The training program was being effectively implemented and the ERO is adequately

staffed. The inspector concluded that the licensee maintained on-shift dose

assessment and adequate back-up capabilities to ensure that on-shift dose

assessments could be performed.

P6

EP Organization and Administration

P6.1 Emeraency Plannina Staffino and Manaaement Chanaes

a.

Insoection Scope (82701)

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The inspector reviewed the licensee's EP group staffing and its management to

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determine what changes have occurred since the last program inspection

(July 1995), whether changes if any had an adverse effect on the EP program.

b.

Observations and Findinas

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Responsibility for EP program is assigned to the Manager, Licensing and

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Engineering Support. Reporting to this Manager is the Environmental Health

and Safety / Emergency Planning Section Head. There is one onsite and one

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offsite Emergency Planning Coordinator who report to the Section Head.

Additional resources from the Yankee Atomic Electric Company are available

to assist in the administration and implementation of the program.

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The inspector interviewed the President of Maine Yankee Atomic Company,

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Vice President Operations, the Plant Manager, Environmental Health and

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Safety / Emergency Planning Section Head, Senior Emergency Preparedness

Coordinator, Principal Emergency Preparedness Coordinator, and Emergency

Preparedness Specialist, Yankee Atomic Services Division. All personnel interviewed

were members of the ERO and were very familiar with and supportive of the

emergency preparedness program, its issues and requirements.

c.

Conclusions

There were no changes in the emergency planning staff since the last inspection.

The staff is knowledgeable and appears adequate to administer and implement the

program properly.

P7

Quality Assurance (QA)in EP Activities

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P7.1 Review of Annual Emeraency Plannina Proaram Audit Reoorts

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a.

Inspection Scope (82701J

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The inspector reviewed the QA audit reports of the EP program, conducted in 1995

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and 1996, to determine compliance of NRC requirements and licensee commitments.

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b.

Observations and Findinas

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The reports reviewed were MY-95-14 and MY-96-14, as well as the corresponding

audit plans and checklists. Tnere were no adverse findings in either report. Audit

j

report 95-14 had four observations, three which indicated areas for improvement in

the facility and equipment inventories and one on training' course descriptions and

lesson plan content. The 96-14 audit report had three observations which consisted

of emergency plan clarifications, minor facility maintenance items and administrative

inconsistencies. The inspector also noted that there were no recurring items in either

of the audit reports.

i

The inspector interviewed the QA Supervisor. The QA Supervisor stated that he

uses the " Risk-Based Quality Verification Process," a portion of Procedure 21-205,

Revision 6, Attachment C, to assess appropriate areas of the program. By applying

the associated factors designated in the procedure, he establishes which areas of the

program may need more resources applied during the audits. Additionally, the

inspector noted, and the QA supervisor indicated, that emergency preparedness

expertise from organizations outside of Maine Yankee assisted with both of the

audits.

The audit reports were distributed to the appropriate levels of management. The

results of all audits on emergency plan offsite group interface were made available to

]

the outside agencies. Timely and appropriate corrective actions were taken on the

j

observations.

c.

Conclusion

The licensee conducted audits that were thorough and met the requirements of

10 CFR 50.54(t).

j

P8

Miscellaneous EP issues

P8.1 Updated Final Safety Analvsis Reoort (UFSAR) Inconsistencies

A recent discovery of a licensee operating its facility in a manner contrary to the

UFSAR description highlighted the need for a special focused review that compares

plant practices, procedures, and/or parameters to the UFSAR description.

Section 12.5 of the UFSAR refers to the emergency plan. Since the UFSAR does not

specifically include emergency plan requirements, the inspector specifically addressed

Section 6.3.3 Offsite Radiation Levels Assessment and Section 6.6.4 Medical

Treatment in the emergency plan. The inspector also reviewed on shift dose

assessment capabilities and training as discussed in Section P5.

The inspector visited the Midcoast Hospital in Bath, ME, which is the primary facility

for care of contaminated injured patients frorn Maine Yankee. The inspector toured

the facihties used for handling of contaminated injured personnel, interviewed two

emergency room doctors and the head nurse regarding their radiation training and

reviewed a video tape of the 1994 medical drill held at the facility. All had received

.

.

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32

1

training from the licensee and one doctor had received training at the Radiation

Emergency Assistance Center / Training Center (REACTS), Oak Ridge, Tennessee.

The telephone number for REACTS assistance was listed in the procedure for the

emergency room.

No inconsistencies with the emergency plan requirements were noted.

P8.2 Emeraency Resoonse Facilities Ventilation System (URI 50-309/96-12-04)

During the review of the emergency facility surveillance and information provided by

{

the Resident inspectors, the inspector found that the 3rd quarter emergency

ventilation system surveillance, performed in accordance with Plant Engineering

Department (PED) Procedure 3.17.5.1, Attachment 8, "Special Operating

Instruction," identified that damper MVD#2, which is normally open, failed to close.

The damper is solenoid-operated to open and spring return to close. When this by-

pass damper fails to close, unfiltered air is admitted into the TSC, OSC and EOF.

The damper is located in Room 116 of the Maine Yankee staff building. The

ventilation system failure had been documented on September 26,1996, in Unusual

Occurrence Report Number 90-090.

While touring the TSC, OSC, and EOF, the inspector observed maintenance in

progress to repair the ventilation system. The inspector observed that the solenoid

i

operator for MVD#2 had been replaced with a new, larger model, and that the

!

maintenance technician was installing a new position indicator and repairing the

wiring for the position indicator light in Room 118.

!

The inspector reviewed the plant alteration specifications, and EPIP 2-50-8,

ATTACHMENT B, " EMERGENCY VENTILATION SYSTEM," and noted that the Plant

Alteration Specification, Section 3.e, stated, in the fifth paragraph, that the air is

brought in through an air tight duct and that MVD#2 is normally closed, but opens

during an emergency condition to allow air through the filtration system. In actuality,

MVD#2 is normally open and is closed under emergency conditions to prevent

unfiltered air from entering the TSC, OSC and EOF. Because of the inconsistencies

between the plant alteration specifications and the EPIP 2-50-8, Attachment B,

the inspector asked the Onsite Emergency Preparedness Coordinator to pursue the

following questions:

1.

Had the plant alteration specifications, drawings, and EPIP 2-50-8,

Attachment B, been compared for accuracy and correctness of damper

labeling and description?

2.

Had the damper actuator component substitution been evaluated to see if it

complies with the original design specifications?

3.

Since the actuators for both MVD#1 and MVD#2 were being replaced and

were notably larger than the original design and were inside the duct, would

there be an effect on air flow characteristics and capacity?

4.

Were the rubberized gaskets and transition boots between the duct wcrk and

fans were covered under a preventive maintenance program?

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5.

Because of the discrepancies noted in question 1, are the position indication

lights for the actuators in Room 118 correct?

Section 7.1.2 of the emergency plan states: "The first floor ventilation system for the

)

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EOF and TSC is equipped with HEPA and charcoal filtration which is placed in

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operation when the TSC is activated. The system is designed to provide

1

approximately 1000 cfm of filtered air to all ERF's collectively, and to also provide

2000 cfm of recirculated / filtered air to the TSC."

!

During the inspection, the Onsite EPC transmitted a memorandum to PED requesting

!

responses to the inspectors questions. This is considered to be an Unresolved item

j

j

(URI 50-309/96-012-04).

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S1

Conduct of Security and Safeguards Activities

S1.1 Control of Safeauards Information (VIO 50-309/96-12-05)

a.

Insoection Scope (71750)

The inspector reviewed the licensee's response and corrective actions upon

identification of a failure to properly control Safeguards information.

This review was initiated due to several past problems with the control of Safeguards

information. Security event reports were issued in May 1996, and June 1993

documenting failure to properly control Safeguards information. in addition, eight

loggable events have occurred between 1993 and 1996.

b.

Observations and Findinas

On October 17,1996, Maine Yankee security personnel informed the PSS that a

security administrative assistant called from home to report that safeguards

information had been left unattended on a desk in an unsecured office inside the

protected area. Although the office is located inside the protected area, it is not

locked and is accessible to persons other than security officers.

Security personnel and the PSS immediately went to the area and all safeguards

material was recovered at 5:33 pm. The licensee determined that the administrative

assistant had left the site at 3:07 pm which left the safeguards information in a

,

compromised condition for approximately two hours. The event was reported to the

NRC within one hour as required by 10 CFR 73.71. After review by Maine Yankee

j

security personnel it was determined that no station personnel had been in the

security office area during the two hour period, due to the close proximity of the

security officer day room and locker room.

The Maine Yankee NRC Approved Security Plan and 10 CFR 73.21.d (2) requires in

]

i

part, that while unattended, Safeguards Information shall be stored in a locked

security storage container. Failure to properly control Safeguards information on

October 17,1996, is a violation of NRC approved Security Plan and 10 CFR 73.21

(VIO 50-309/96-012-05).

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c.

Conclusion

a

The inspector identified a violation for failure to properly control Safeguards

i

information. The inspector also determined that Maine Yankee properly responded to

j

the event as required by the station security plan and NRC requirements. At the

completion of the inspection the licensee had not identified the root cause of this

,

J

violation. This violation is notable in that there have been several instances of failure

i

to properly control safeguards information in the last three years.

S7

Quality Assurance in Security and Safeguards Activities

.

2

S7.1 Review of Annual Security Procram Audit

a.

Insoection Scope (71750)

The inspector reviewed the results of the annual security program as required by

j

10 CFR 73.55.

I

b.

Observations and Findinas

i

During the inspection period, the inspector reviewed the 1996 Annual Security

Program audit that was conducted on September 9-13, and 17-18,1996, by

technical specialists from nuclear power plants in Region I that were SALP

,

,

category 1. The audit scope was well defined and the findings were well presented.

]

No major discrepancies were identified but several observations were identified

during the audit. The inspector reviewed the auditors observations and the licensee's

a

corrective actions to resolve the identified concerns. These were found to be

appropriate. These observations included a finding of the use of ro::ks to fill holes

{

under the protected area barrier, an incomplete search of a package by a security

officer, a lack of complete documentation of the Vehicle Barrier System and need to

enhance several security procedures. The audit team determined that the control of

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Safeguards Information (SI) was adequate.

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c.

Conclusion

The inspector determined that Meine Yankee, with assistance from off site security

.

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specialists, properly performed the required annual audit of the security program.

The audit was comprehensive with the proper involvement of the station quality

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,

!

programs department. The inspectors held discussions with the security operations

1

supervisor and determined that the corrective actions were appropriate to resolve the

j

observations. However, the observation concerning the control of Si material

appears to require more extensive corrective actions to resolve the problem of

continued failures in this area.

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35

V. Manaaement Meetinas

X1

Exit Meeting Summary

X1.1 Routine Resident inspection Exit Meetina

The inspectors presented the inspection results to members of the licensee on

October 31,1996. The licensee acknowledged the findings presented.

X1.2 Radioloaical Control Inspection Exit Meetina

The Radiological Control inspection results were presented to members of the

Licensee on September 27,1996. The licensee acknowledged the findings

presented.

X 1.3 Emeraency Preparedness insoection Exit Meetina

The Emergency Preparedness inspection results were presented to members of the

licensee on October 11,1996. The licensee acknowledged the findings presented.

X3

Manaaement Meetina Summarv

X3.1 Indeoendent Safety Assessment Team Exit Meetina

On October 10,1996, the Independent Safety Assessment Team conducted a final

exit meeting to discuss the major findings and conclusions of the team. The meeting

was held at the Wiscasset Middle School, Wiscasset, Maine, and was in two parts.

Part 1 was between the NRC and Maine Yankee and was open for public observation

only. The second part was primarily to respond to questions from the public on the

ISA process and findings.

Attached are the " handouts" provided by NRC staff and Maine Yankee. The NRC

transcripts of both parts of the meeting will be issued under separate

correspondence.

,

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PARTIAL LIST OF PERSONS CONTACTED

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Licensee

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R. Blackmore, Plant Manger

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J. Connell, Technical Support Department Manager

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S. Evans, Environmental Health and Safety / Emergency Preparedness, Section Head

C. Giggey, Plant Engineer

J. Grant, Plant Support

J. Hebert, Manager, Licensing and Engineering Support

G. Leitch, Vice President Operations

J. Mathieson, Onsite Emergency Preparedness Coordinator

T. Marstaller, Plant Engineer

J. McArdle, Senior Emergency Planner, Yankee Nuclear Service Division

J.' McCann, Licensing Section Head-

P. Metivier, Security Manager

J. Niles, Assistant Operations Manager

G. Pillsbury, Emergency Preparedness Training

P. Radsky, Chemistry Section Head

C. Shaw, Plant Manager (left post)

F. Smith, Chemistry Section Head

S. Smith, Operations Manager

E. Soule, Plant Engineering Manager

J. Temple, Senior Emergency Preparedness Coordinator

W. Tracy, Acting Supervisor - QA

M. Veilleau, Maintenance Manager

J. Weast, Licensing Engineer

D. Whittier, Vice President Licensing and Engineering

NRC

L. Eckert, Radiation Specialist

J. Jang, Sr. Radiation Specialist

J. Lusher, Emergency Preparedness Specialist

W, Olsen, Resident inspector

J. Yerokun, Senior Resident inspector

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INSPECTION PROCEDURES USED

IP 37551:

Onsite Engineering

IP 40500:

Effecuveness of Licensee Controls in Identifying, Resolving, and Preventing

Pr-)blems

IP 62707:

Msintenance Observation

IP 71707:

Plant Operations

IP 71750:

Plant Support Activities

IP 73753:

Inservice inspection

IP 82701:

Operational Status of the Emergency Preparedness Program

IP 83729:

Occupational Exposure During Extended Outages

{

IP 83750:

Occupatior a! Exposure

lP 84750:

Radioactive Waste Trestment, Effluent and Environmental Monitoring

IP 92700:

Onsite Followup of Written Reports of Nonroutine Events at Power Reactor

Facilities

,

IP 92901:

Operations Followup

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IP 92902:

Followup - Engineering

IP 92903:

Followup - Maintenance

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ITEMS OPENED, CLOSED, AND DISCUSSED

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Items Opened:

50-309/96-12-01

VIO

Failure to install Primary Vent Stack Sampling Filters as required

during conduct of maintenance in accordance with TS 5.8.a.3.

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(Section 02.1)

f

50-309/96-12-02

URl

Unplanned Reactor Power increase to 2457 Mwt (Plant Limited

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to 2440) During RCS Delithiation on September 23,1996.

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(Section 04.1)

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50-309/96-12-03

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Documentation of air balance surveillance testing results to

demonstrate consistency with Chapter 9.13 of the FSAR.

Update of section 9.13.2.4 of the FSAR (Section R2.3)

50-309/96-12-04

URI

Questions regarding Emergency Response Facility Ventilation

.

System maintenance. Maintenance activities might have

negatively affected the ERF ventilation system (Section P8.2)

50-309/96-12-05

VIO

Failure to Properly Control Safeguards information. (Section

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S1.1)

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Items Closed:

50-309/96-007-01 URI

Emergency plant changes 96-01 and 96-02 reviewed and found

i

to be in compliance with 10 CFR 50.54(q) and did not decrease

the effectiveness of the emergency plan. (Section P3.1)

Items Discussed:

50-309/96-08-02

URI

High Energy Line Break in the Turbine Building (Section E2.1)

50-309/96-08-04

URI

Turbine Building Flood Protection (Section E2.1)

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LIST OF ACRONYMS USED

AEOD

Office for Analysis and Evaluation of Operational Data

ADP

Air Particulate Detector

ARP

Alarm Response Procedure

BAST

Boric Acid Storage Tank

CATS

Corrective Action Tracking System

CEA

Control Element Assembly

CED

Corporate Engineering Department

CFR

Code of Federal Regulations

CR

Control Room

DBS

Design Basis Screen

ENS

Emergency Notification System

EP

Emergency Preparedness

ERB

Event Review Board

ERDS

Emergency Response Data System

,

ERO

Emergency Response Organization

ESF

Engineered Safety Feature

ESS

Emergency Support System

FES

Front End Support

gpm

Gallons Per Minute

GPO

Government Printing Office

HELB

High Energy Line Break

HEPA

High Efficiency Particulate Air

IFl

Inspection Follow-Up item

IFS

Inspection Follow-Up System

IMC

Inspection Manual Chapter

INCA

Incore Analysis

I&C

Instrument and Control

IPAP

Integrated Performance Assessment Process

ISI

in-Service Inspection

LCO

Limiting Condition of Operation

LER

Licensee Event Report

MD

Management Directive

METPAC

Meteorological Post Accident Computer Model

MWt

Megawatts Thermal

NCV

Non-Cited Violation

NMSS

Office of Nuclear Material Safety and Safeguards

NOV

Notice of Violation

NRC

Nuclear Regulatory Commission

NRR

Office of Nuclear Reactor Regulation

ODCM

Off-Site Dose Calculation Manual

OE

Office of Enforcement

01

Office of Investigations

OSS

Operational Support System

PDT

Primary Drain Tank

PIPB

inspection Program Branch

PORC

Plant Operations Review Committee

PPR

Plant Performance Review

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PSS

Plant Shift Supervisor

PVS

Primary Vent Stack

QA

Quality Assurance

RA

Regional Administrator

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REMP

Radiological Environmental Monitoring Program

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RHR

Residual Heat Removal

RlR

Radiological incident Report

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RMS

Radiation Monitoring System

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RP

Radiation Protection

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RP&C

Radiological Protection and Chemistry

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RPS

Reactor Protection System

SALP

Systematic Assessment of Licensee Performance

SPDS

Safety Parameters Display System

SOS

Shift Operating Supervisor

STA

Shift Technical Advisor

TE

Technical Evaluation

Tl

Temporary Instruction

TS

Technical Specification

UFSAR

Updated Final Safety Analysis Report

UOR

Unusual Occurrence Report

URI

Unresolved item

VCT

Volume Control Tank

WO

WorkOrder

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MAINE YANKEE HANDOUT

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INFORMATION

INDEPENDENT SAFETY

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ASSESSMENT TEAM

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OPEN PUBLIC MEETING

OCTOBER 10,1996

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MAINE YANKEE COMMFNFS ON ISA REPORT

(Talking Points Version]

]

Report reflects excellent effort by the NRC and the State

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Report supports Maine Yankee's position that plant is

operating safely

[ Safe operation for over 24 years: report conprms that when issues are identifed which are safety

signtpcant, necessary actions are taken]

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ISA Report is balanced in content and reasonable in its

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subjectivity

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Maine Yankee does not have significant disagreement with

the technical facts in the report

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MAINE YANIME COMMENTS ON ISA REPORT

Many ISA issues were identified by MY prior to the

inspection

(Culture Assessment, Learning Process, Maintenance improvement Program (including a Procedure

'

Adherence Initiative). Engineering Quality improvement Program, Safety Analyses improvement Plan,

Industrial Safety improvement Initiatives and Supervisory improvement Plan)

New issues raised by the ISA are receiving prompt

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attention

[NPSH, surveillance testing, timeliness ofcorrective actions, design / licensing basis issues]

NRC root.cause statements are reasonable when viewed in

.s

context

[Willdiscuss in more detaillater]

Overall, being subjected to the ISA was a significant

learning experience

.

[ Regulatory thresholds, industryprocaces, opportunityfor re-calibrating interncipolicies andpractices]

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Commend all Maine Yankee personnel for their dedicated

and competent response to all challenges presented by the

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ISA

[I wanted to publicly thank themfor their dedication. Personnel at MYperform theirjobs proudly, they

endure long hours ofnever-endingpublic and regulatory scrutinv, and demonstrate that they truly care

about this plant and the safety ofthe public and that they respect to thefullest, the regulatory envannment

in which we work]

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ISA ROOT CAUSE # 1

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" Economic pressures to contain cost"

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Report correctly notes that notwithstanding these

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pressures, " management has effectively operated the

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plant within the budget constraints"

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" Management has effectively prioritized available

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resources, but financial pressures have caused the

'

postponement of some needed program improvements

!

and actions"

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[ Safety is always paramount duringplant operation and during budget decisions]

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[Whether we like talking about it or not, Maine Yankee has to priariti:e its activities while

ensuring that saf.ety issues are addressed as necessary]

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[ Postponed activities were not viewed at the time to have safety sigmficance; we are

recon.ridering the methods that we use to make these types ofdeci.sions]

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ISA ROOT CAUSE # 2

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" Poor problem identification as a result of complacency and a

lack of a questioning attitude"

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Maine Yankee agrees with statement in the context of

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the full report

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Agrees with NRC statement that involved areas were

-

perceived by . management to

be of low safety

significance

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[I would like to emphasi:e that we do not believe that this comment applies to all

employees nor does it apply to the company as a whole - our workforce was not

<

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complacent regardingissuesthatpersonnelbelievedweresafetysignificant-thoseissues

were promptly addressed to the best oftheir ability]

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[We, the managers at Maine Yankee will absorb the blamefor those circumstances where

l

we did not exhibit an adequate questioning arntude. It is ourjob to ensure thatpotentially

j

, signtjicant and credible "what ifs"are addressedin every circumstance. In that regaral

the ISA helped us recalibrate our way ofapproaching some issues]

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MATNE YANKEE RESPONSE METHODOLOGY

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Response to the NRC ISA Report within 60 days

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Commitment to Excellence Action Plan provides an

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immediate response to ISA findings and other Maine

Tankee issues

i

Provides

comprehensive

plan

for

achieving

and

'

'>

' maintaining excellent Maine Yankee performance

Resources

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Organization

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Board of Directors' Oversight

.

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Programs

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People

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Maine Yankee's focus is clear, and its commitment to

success is unwavering

[We are committed to providing adequate resources and running this plant safety or not at all-

-

when you cut through all ofthe issuesfacing us, it is that simple!]

S

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NRC STAFF HANDOUT

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INFORMATION

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INDEPENDENT SAFETY

,

ASSESSMENT TEAM

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6

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OPEN PUBLIC MEETING

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OCTOBER 10,1996

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INDEPENDENT SAFETY ASSESSMENT

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UNITED STATES

T-

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NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-4001

.

g 4g#

October 7, 1996

.....

CHAIRMAN

Mr. Charles D. Frizzle, President

Maine Yankee Atomic Power Company

329 Bath Road

Brunswick, Maine 04011

Dear Mr. Frizzle:

I am forwarding the report on the Maine Yankee Atomic Power Station by the

Nuclear Regulatory Commission's Independent Safety Assessment (ISA) team.

The

purpose of the ISA was to determine whether Maine Yankee was in conformity

with its design and licensing bases; to assess operational safety performance;

and to evaluate Maine Yankee's self-assessment, corrective actions, and plans

for improvement.

Overall performance at Maine Yankee was considered adequate for operation.

However, a number of significant weaknesses and deficiencies were identified

that will result in violations.

These weaknesses and deficiencies appear to

be related to two root causes:

economic pressures to contain costs and poor

problem identification as a result of complacency and a lack of a questioning

attitude.

The ISA review was conducted in response to findings made by the NRC's Office

of the Inspector General (0IG) in a report dated May 8,1996.

It included an

assessment of the analytic code support provided for Maine Yankee by the

Yankee Atomic Electric Company. The OIG report found, among other things,

that Maine Yankee had experienced problems with the RELAP/5YA computer code,

used for analyzing how the emergency core cooling system would function during

a small break loss-of-coolant accident (LOCA), and in response, had modified

that code.

OIG also found that these problems with the computer code had not

been reported to the NRC, as required, and that because of these problems,

Maine Yankee's use of the code was not in accordance with NRC requirements.

NRC reviews did not uncover these deficiencies.

The team was large and multi-discipliced in order to provide a thorough, in-

depth review.

Its 25 members, led by an NRC manager, included three

representatives of the State of Maine.

To ensure an independent perspective,

the NRC members were selected from NRC offices other than the Office of

Nuclear Reactor Regulation (NRR) and the NRC's Region 1.

Only persons with r)o

significant prior responsibility for regulating Maine Yankee were chosen.

The

team's management reported to me.

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The ISA team was on site at Maine Yankee between July 15 and 26, 1996, and

)

again between August 12 and 23, 1996.

During these time periods, team members

also conducted assessments at Maine Yankee's corporate headquarters in

Brunswick, Maine, and at the Yankee Atomic Electric Company offices in Bolton,

Massachusetts.

.

The ISA team reviewed the use of selected analytic codes for performing non-

,

LOCA safety analyses, as well as the capability of the safety-related support

systems to perform in accordance with the assumptions made in those analyses.

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The review determined that the conditions of approval in NRC Safety Evaluation

Reports have been met although weaknesses in documentation and validation of

plant specific code applications are vulnerabilities which warrant your

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attention.

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The team determined that cycle-specific core performance analyses were

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excellent. However, weaknesses were fcund in more complicated, less

frequently performed system safety analyses.

These weaknesses did not cause

the results to exceed Maine Yankee's design and licensing bases.

However, the

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team questioned the capability of the containment spray system and the

component cooling water systems to meet the design basis assumptions for a

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LOCA initiated from greater than 2440 MWt.

These issues, along with the

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RELAP/5YA deficiencies, will be reviewed by NRC's Office of Nuclear Reactor

Regulation.

The team identified significant deficiencies in the areas of maintenance and

engineering, as well as weaknesses in the overall approach to testing and the

corrective action program.

Specifically, the lack of routine testing of

certain safety systems resulted in the existence of a significant deficiency

of which Maine Yankee was unaware.

In addition, the ISA noted certain design

errors.

Either Maine Yankee was unaware of these errors, or it was aware of

them and had failed to take action to address them.

,

I should add that Maine Yankee deserves credit for having formed a counterpart

team of highly qualified personnel to interface with the ISA team during its

review. The existence of this team was both helpful to the ISA team's

activities and valuable as a means of ensuring that Maine Yankee learned as

much as possible from this effort.

In addition, it meant that as problem

areas were identified, Maine Yankee was in a position to devote resources

promptly to necessary corrective actions.

We have scheduled a meeting for October 10, 1996, during which we will discuss

the assessment and respond to questions you may have.

I request that

following this meeting, you determine the actions needed to ensure the long-

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term resolution of the deficiencies noted.

I also request that by

December 10, 1996, you provide to the Commission your plans for addressing the

root causes of the deficiencies identified by the ISA.

The NRC's Region I and

its Office of Nuclear Reactor Regulation will be responsible for followup of

the issuns identified in this assessment, in terms of overseeing corrective

actions and taking any enforcement action deemed appropriate.

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In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure

will be placed in the NRC Public Document Room.

Should you have any questions

concerning this assessment, I would be pleased to discuss them with you.

Sincerely,

/DL

g~a ykw

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Shirley Ann Jackson

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Enclosure:

Independent Safety Assessment Report

for Maine Yankee Atomic Power Company

cc:

See page 4

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cc w/ enclosure:

Mr. Charles B. Brinkman

Mr. Christopher R. Shaw

Manager - Washington Nuclear

Plant Manager

Operations

Maine Yankee Atomic Power Station

ABB Combustion Engineering

P.O. Box 408

)

12300 Twinbrook Parkway, Suite 330

Wiscasset, ME 04578

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Rockville, MD 20852

Mr. G. D. Whittier, Vice President

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Thomas G. Dignan, Jr., Esquire

Licensing and Engineering

Ropes & Gray.

Maine Yankee Atomic Power Company

One International Place

329 Bath Road

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Boston, MA 02110-2624

Brunswick, ME 04011

Mr. Uldis Vanags

.

Mr. Patrick J. Dostie

State Nuclear Safety Advisor

State of Maine Nuclear Safety

State Planning Office

Inspector

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State House Station #38

Maine Yankee Atomic Power Station

Augusta, ME 04333

P.O. Box 408

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Wiscasset, ME 04578

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Mr. P. L. Anderson, Project Manager

Yankee Atomic Electric Company

Mr. Graham M. Leitch

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580 Main Street

Vice President, Operations

Bolton, MA 01740-1398

Maine Yankee Atomic Power Station

P.O. Box 408

Regional Administrator, Region I

Wiscasset, ME 04578

U.S. Nuclear Regulatory Commission

475 Allendale Road

Mary Ann Lynch, Esquire

King of Prussia, PA 19406

Maine Yankee Atomic Power Company

329 Bath Road

First Selectman of Wiscasset

Brunswick, ME 04578

Municipal Building

U.S. Route 1

Mr. Jonathan M. Block

Wiscasset, ME 04578

Attorney at Law

P.O. Box 566

Mr. J. T. Yerokun

Putney, VT 05346-0566

Senior Resident Inspector

Maine Yankee Atomic Power Station

Mr. James R. Hebert, Manager

U.S. Nuclear Regulatory Commission

Nuclear Engineering and Licensing

P.D. Box E

Maine Yankee Atomic Power Company

Wiscasset, ME 04578

329 Bath Road

Brunswick, ME 04578

Friends of the Coast

P.O. Box 98

Edgecomb, ME 04556

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EXECUTIVE SUMMARY

Background

In December 1995, the Union of Concerned Scientists forwarded anonymous

allegations to the State of Maine, and the State submitted the allegations to

the NRC. The allegations were that Yankee Atomic Electric Company knowingly

'1

performed inadequate analyses to support an increase in the rated thermal

power at which Maine Yankee Atomic Power Station (MYAPS) may operate.

After

performing a technical review, the NRC Office of Nuclear Reactor Regulation

(NRR) issued a confirmatory order on January 3, 1996, limiting power operation

at the plant to the original licensed power level of 2440 MWt.

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The NRC Office of the Inspector General (0lG) completed an inquiry into this

allegation on May 8,1996.

0IG established that MYAPS had experienced

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problems with, and made modifications' to, the RELAP/5YA computer code which

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was used in the emergency core cooling analysis fcr a small-break loss-of-

,

coolant accident.

OIG also reported weaknesses in the NRC review and followup

activities which contributed to NRC failure to detect these deficiencies.

In

response to these findings, as well as to respond to concerns by the Governor

of Maine about the safety and the effectiveness of regulatory oversight of

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MYAPS, the NRC Chairman initiated an independent safety assessment of MYAPS.

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This assessment was to be performed by a team comprised of staff who were

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independent of any recent or significant regulatoiy oversight responsibility

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for MYAPS. Additionally, the assessment was to be coordinated with the State

of Maine to facilitate participation by State representatives consistent with

the Commission's policy on cooperation with States at commercial nuclear power

plants (57 FR 6462, February 25, 1992).

{

Licensing and Design-Basis

.

Maine Yankee was in general conformance with its licensing-basis although

significant items of non-conformance were identified.

The licensing-basis was

understood by the licensee but lacked specificity, contained inconsistencies,

and had not been well maintained.

The use of analytic codes for safety analyses was very good. Cycle specific

core performance analyses were excellent.

More complicated, less frequently

performed safety analyses contained weaknesses, but the analyses were found to

be acceptable based on compensating margin.

Conditions of use specified in

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the safety evaluation reports were found to be satisfied, but not documented.

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The quality and availability of design-basis information was good overall.

Despite uncorrected and previously undiscovered design problems, the design-

basis and compensatory measures adequately supported plant operation at a

power level of 2440 MWt.

However, the team could not conclude, and the

licensee did not demonstrate, that at a power of 2700 MWt the design-basis

assured adequate NPSH for the containment spray pumps and the heat removal

,

capability of the component cooling water system in the event of a loss-of-

coolant accident.

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Operations

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Performance in the area of operations was very good, with strengths noted in

the areas of operator performance during routine and transient operating

conditions; shift turnovers and pre-evolution briefs; use of risk information

to assure safe operations; and the involvement of management in day-to-day

e

operations.

Weaknesses were noted in the area of "workarounds" and

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compensatory measures which unnecessarily burdened the operators or

complicated their response to transient conditions. Additionally, log keeping

practices and post-trip reviews lacked rigor.

Maintenance and Testing

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Performance in the area of maintenance was good overall however, testing was

weak.

The results of the review of equipment reliability for the auxiliary

feedwater, emergency feedwater, high pressure safety injection, and emergency

diesel generator systems showed mixed equipment performance.

Strengths were

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noted in the areas of knowledge and use of risk methodologies for planning,

prioritizing, and scheduling work; the control and limited use of temporary

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sealants; and a motivated and dedicated work force. Although material

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condition was considered good overall, a number of significant material

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condition deficiencies were noted as was a decline in material condition

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following the 1995 steam generator tubing out age.

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Inadequacies in the scope of testing programs were identified, as were

weaknesses in the rigor with which testing was performed and in the evaluation

of testing results to demonstrate functionality of safety equipment. A lack

of a questioning attitude and stressed resources resulted in the use of poor

surveillance procedures and ineffective evaluation of surveillance test data.

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Engineering

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The quality of engineering work was mixed but considered good overall.

Strengths were noted in the capability and experience of the engineering

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staff, day-to-day engineering support of maintenance and operations, in the

quality of most calculations, and in the routine use and application of

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analytic codes.

However, engineering was stressed by a shortage of resources,

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and there was a tendency to accept existing conditions.

Specific weaknesses

]

were noted with inconsistent identification and resolution of problems,

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inadequate testing, and work on some calculations and analytic codes.

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Self Assessment and Corrective Actions

Weaknesses were identified in the areas of problem identification and

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resolution.

While licensee self-assessments were generally good, they

.

occasionally failed to identify weaknesses or incorrectly characterized the

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significance of the findings. Additionally, some corrective actions were not

timely and others were ineffective, leading to repetitive problems.

Licensee

planning was generally effective, although some weaknesses were found in the

overall implementation of improvement plans.

Some economic pressures resulted

in limitations on resources, which impaired the licensee's ability to complete

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improvement projects that affected plant safety.

Equipment problems were not

resolved and improvement programs were not effectively implemented because

the licensee perceived them to be of low safety significance.

Root Causes and Overall Conclusions

While overall performance at Maine Yankee was adequate for operation, a number

of deficiencies were identified by the team in each of the areas assessed.

These deficiencies, which included weak identification and resolution of

problems; weak scope, rigor, and evaluation of testing; and declining material

condition stemmed from two closely related root causes.

These root causes

were (1) economic pressure to be a low-cost energy producer has limited

available resources to address corrective actions and some plant improvement

upgrades and (2) there is a lack of a questioning culture which has resulted

in the failure to identify or promptly correct significant problems in areas

perceived by management to be of low safety significance.

The economic pressures discussed in Section 4.3 resulted in limitations on

resources and interfered with the licensee's ability to complete projects and

other efforts that would improve plant safety and testing activities.

Examples include the failure to adequately test safety related components

(Section 3.2.4); the long-standing deficient design conditions, such as the

undersized atmospheric steam dump valve (Sections 3.1.3.1 and 3.3.1) and

environmental qualification issues (Section 2.3.9); and the lack of effective

improvement programs, such as the design basis reconstitution program

(Sections 3.3.3 and 4.3.3).

These and other examples discussed in the report

illustrate the licensee's willingness to accept existing conditions, many of

which became operator workarounds (Section 3.1.1.1).

Examples of issues which illustrate complacency and the failure to identify or

promptly correct significant problems, include previously undiscovered

deficient conditions of the service water and auxiliary feedwater water

systems (Section 3.2.2); inadequacies in ventilation systems (Section 2.3.7);

post-trip reviews which lacked rigor and completeness (Section 3.1.2.7);

emergency operating procedures that may not adequately address an inadequate

core cooling event and a steam generator tube rupture under certain conditions

(Section 3.1.3.1); lack of a questioning attitude during test performance and

evaluation that was not conducive to discovering equipment problems, but

rather to accepting equipment performance (Sections 2.2.1, 3.2.2, 3.2.4); and

licensee self-assessments that occasionally failed to identify weaknesses, or

incorrectly characterized the significance of findings (Section 4.1).

In

addition, some corrective actions were not timely and others were ineffective,

leading to repetitive problems (Section 4.2).

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UNITED STATES NUCLEAR REGULTORY COMMISSION

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INDEPENDENT SAFETY ASSESSMENT TEAM

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OF

MAINE YANKEE ATOMIC POWER STATION

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PUBLIC MEETING OCTOBER 10,1996

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SELECTION OF MAINE YANKEE

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ALLEGATIONS REGARDING RELAP/5YA

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OFFICE OF INSPECTOR GENERAL INQUIRY

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STATE OF MAINE CONCERNS

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INDEPENDENT SAFETY ASSESSMENT

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LARGE EXPERIENCED TEAM

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INDEPENDENT OF NRR AND REGION I

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PARTICIPATION BY STATE OF MAINE

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MAINE YANKEE INDEPENDENT SAFETY ASSESSMENT

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MISSION

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Provide an independent assessment of the conformance of Maine Yankee

Atomic Power Station to its design and licensing bases including

appropriate reviews at the site and corporate office.

2.

Provide an independent assessment of operational safety performance

providing risk perspectives, where appropriate.

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3.

Evaluate the effectiveness of licensee self-assessment, corrective actions,

and improvement plans.

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Determine the root cause(s) of safety significant findings and draw

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conclusions on overall performance.

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Maine Yankee

Independent Safety Assessment Team

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Edward L. Jordan

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Team Manager

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Ellis W. Merschoff

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Team Leader

Region 11

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Uldis Vanags

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Ola B. West

Patrick Dostie

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David Decrow

State Representatives

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Kriss M. Kennedy

Ronald Lloyd

Thomas O. Martin

Alan L Madison

Jack E. Rosenthal

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OPS / Training

Maint./ Testing

Eng. Design / Tech.

Management & Organization

Analytic Code Support

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AEOD

AEOD

Region IV

AEOD

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Larry Bell

Russell Bywater

John Boardman

Harold Christensen

G. Norman Lauben

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Region ill

AEOD

Region il

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John Kauffman

Peter Prescott

George Hausman

Brian Haagensen

Leonard Ward

AEOD

AEOD

Region lli

Contractor

Contractor

Robert Christie

Cyril Crane

Contractor

Contractor

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George Cha

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Contractor

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Michael Shlyamberg

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STATE PARTICIPATION

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TECHNICAL TEAM - DAY-TO-DAY PARTICIPATION IN EACH OF

THE FIVE FUNCTIONAL AREAS BEING ASSESSED.

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ULDIS VANAGS

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PATRICK DOSTIE

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DAVID DECROW

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PROCESS TEAM - OBSERVE THE PROCESS AT KEY

MILESTONES TO ASSURE ISAT IS FAIR, BALANCED, AND

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OBJECTIVE.

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PETER WILEY

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CITIZEN'S GROUP - PERIODIC BRIEFINGS FOR GOVERNOR AND

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CITIZEN'S GROUP TO KEEP THEM INFORMED OF PROGRESS.

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DR. DON ZILLMAN

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MR. ROGER HEWSON.

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MS. ELIZABETH ARMSTRONG

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DR. EDWARD LAVERTY

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MR. THOMAS BROUSSARD

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LICENSEE SUPPORT ORGANIZATION

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EXTENSIVE RESPONSE LIBRARY

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EFFECTIVE LINK TO LINE ORGANIZATION

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THOROUGH EXTENT OF CONDITION REVIEWS

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SAFETY ASSESSMENT SCHEDULE

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JUNE 17-JULY 12

TEAM PREPARATION

JULY 15

PUBLIC ENTRANCE MEETING

JULY 15-26

FIRST ONSITE EVALUATION PERIOD

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AUGUST 12-23

SECOND ONSITE EVALUATION PERIOD

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OCTOBER 8

ISSUE REPORT

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OCTOBER 10

PUBLIC EXIT MEETING

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SAFETY ASSESSMENT ACTIVITIES

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WALKDOWN SYSTEMS

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EXTENDED CONTROL ROOM OBSERVATIONS

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VERTICAL SLICE REVIEW

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PROGRAM / PROCESS / PROCEDURE REVIEW

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ANALYTIC CODE REVIEW

HORIZONTAL REVIEW TO SER

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INTERVIEWS

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REGULATIONS - MEASURE CONFORMANCE

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ASSESSMENT - MEASURE MARGIN OF SAFETY

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ACCEPTABLE

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NRC ASSESSMENT STANDARDS

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GOOD

ACCEPTABLE

Safety

Properly Focused

Normally Well Focused Acceptable

Performance

Programs

Effective Control

Some Deficiencies Exist Instances of

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Insufficient Control

Self Assessment

Effective

Some Issues Not

May not Occur

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Identified

Until Problem is

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Apparent

Corrective Actions

Comprehensive

Some Not Complete

Not Thorough

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Root Cause

Recurring Problems

Normally Thorough

Do Not Probe

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Eliminated

Deeply

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SAFETY ASSESSMENT RESULTS

OVERALL PERFORMANCE ADEQUATE FOR OPERATION

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DESIGN AND LICENSING BASIS - GENERALLY IN

CONFORMANCE

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OPERATIONS - VERY GOOD

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MAINTENANCE - GOOD

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TESTING - ACCEPTABLE

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ENGINEERING - GOOD

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ACCEPTABLE

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LICENSING AND DESIGN BASIS

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LICENSING BASIS

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• LACKS SPECIFICITY

> CONTAINS INCONSISTENCIES

> NOT WELL MAINTAINED

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> CYCLE SPECIFIC - EXCELLENT

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DESIGN BASIS

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> GOOD QUALITY

• GOOD AVAILABILITY

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> ADEQUATELY SUPPORTS 2440 MWt

> OPERATION AT 2700 MWt NOT DEMONSTRATED

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LICENSING / DESIGN BASIS

OPERABILITY ISSUES RAISED

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COMPONENT COOLING PIPING INSIDE CONTAINMENT

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REACTOR WATER STORAGE TANK LEVEL TRANSMITTERS

EQUIPMENT QUALIFICATION FOR SUBMERGENCE

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VENTILATION

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LOGIC CIRCUITRY

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CHECK VALVE TESTING

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SAFETY ASSESSMENT

OPERATIONS

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OVERALL PERFORMANCE VERY GOOD

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STRENGTHS

p

!

> OPERATOR PERFORMANCE

l

• USE OF RISK INFORMATION

l

• MANAGEMENT INVOLVEMENT

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> SHIFT TURNOVERS

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> PRE EVOLUTION BRIEFS

i

WEAKNESSES

[

i

> WORKAROUNDS AND COMPENSATORY MEASURES

• POST TRIP REVIEWS

i

> LOG KEEPING

j

l

!

i

15

I

'

_ ._. _ __. _ _ _ .

_ .._ _ _ _ _ _ _ _. _ _ . _ _ _

-

..

SAFETY ASSESSMENT

MAINTENANCE

OVERALL PERFORMANCE GOOD

.

i

e

STRENGTHS

i

> KNOWLEDGE /USE OF RISK

j

> MOTIVATED / DEDICATED WORK FORCE

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> CONTROL OF TEMPORARY REPAIRS

i

• QUALITY OF MAINTENANCE

{

!

WEAKNESSES

!

• DECLINING MATERIAL CONDITION

• INCONSISTENT EQUIPMENT RELIABILITY

l

,

\\

1

16

. ..... _ - -

- - - .- . . - . - -. - .

- .

.

.

.;

i

SAFETY ASSESSMENT

TESTING

OVERALL PERFORMANCE ACCEPTABLE

~

-

.

e

TESTING WEAKNESSES

INADEQUATE SCOPE

(

>

WEAK RIGOR

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'

>

MTAK EVALUATIONS

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i

!

!

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TESTING STRENGTHS

!

t

>

STEAM GENERATOR

>

INSERVICE TESTING

CONTAINMENT LEAK RATE TESTING

I

,

I

!

i

17

!

_

-

- -

-

- -

- - - - - -

-

-

- - - -

- -

-

- -

- - - - .

- - .

- - .

- . . - _ _ _ _

- .

- - . - .-. - -

_. -

"

SAFETY ASSESSMENT

ENGINEERING

!

OVERALL PERFORMANCE WAS MIXED

e

QUALITY OF ENGINEERING WORK WAS GOOD.

• QUALIFIED CAPABLE STAFF

• GOOD ELECTRICAL DESIGN WORK

• SUPPORT TO OPERATIONS AND MAINTENANCE

> SUPPORT PROVIDED BY YANKEE ATOMIC

!

e

LIMITED OWNERSHIP

,

t

t

> EQUIPMENT QUALIFICATION

> FIRE PROTECTION

• TESTING

i

e

INUONSISTENT IDENTIFICATION AND RESOLUTION OF

PROBLEMS

[

i

'

> VENTILATION

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-

> ATMOSPHERIC DUMP VALVE

!

> AUXILIARY FEED PUMP

I

t

.

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18

l

-

- - .

--

- ------

-

. - - - -

-

- -

- -

- - - - - - - - . . . - . _ - - _ _ - - - - - _ . - - - -

...!

-

EFFECTIVENESS OF SELF ASSESSMENT, CORRECTIVE ACTIONS,

AND IMPROVEMENT PLANS

,

OVERALL EFFECTIVENESS WAS ACCEPTABLE

I

.

e

SELF ASSESSMENT

> INTERNAL / EXTERNAL EFFECTIVENESS MIXED

• OVERSIGHT COMMITTEES EFFECTIVE

l

• FRAGMENTED PROBLEM IDENTIFICATION PROCESS

!

CORRECTIVE ACTION PROGRAM

> FRAGMENTED PROCESS

!

> TRENDING AND TIMELINESS WEAK

[

> OCCASIONALLY INEFFECTIVE

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IMPROVEMENT PLANS

I

'

• MANY INDIVIDUAL PLANS

l

• NO INTEGRATED PLAN

!

> RESULTS MIXED

[

!

19

- - -

.

- - - - - -

-

- - - - -

-

-

.

. . .

. _ _. __

_

_

_

..

_

. _ _ _ _ . _ _

. .

. _ _ _ _ _ . _ .

.

.1

.

!

ROOT CAUSES OF SIGNIFICANT FINDINGS

j

'

.

e

ROOT CAUSE 1

Economic pressure to be a low-cost energy producer has limited

'

available resources to address corrective actions and some plant

improvement upgrades. Management has effectively prioritized

available resources, but financial pressures have caused the

postponement of some needed programs and actions.

i

l

i

!

>

,

!

20

'

L

I

i

- -

-

-

-

- - -

_

_

.

_

..

.--

-

-

-

.

.

ROOT CAUSES OF SIGNIFICANT FINDINGS

.

e

ROOT CAUSE 2

1

There is a lack of a questioning culture which has resulted in the

failure to identify or promptly correct significant problems in areas

perceived by management to be of low safety significance.

Management appears complacent with the current level of safety

performance and there does not appear to be a clear incentive for

improvement.

.

21

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