IR 05000029/1990004

Insp Rept 50-029/90-04 on 900213-0402.Two Noncited Violations Noted.Major Areas Inspected:Operational Safety, Security,Plant Operations,Maint & Surveillance,Engineering Support & Radiological Controls

ML20042G080
Person / Time
Site:	Yankee Rowe
Issue date:	04/30/1990
From:	Pasciak W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20042G078	List: IR 05000029/1990004
References
50-029-90-04, 50-29-90-4, NUDOCS 9005110075
Download: ML20042G080 (22)

Text

.

..;

...

'

t

+

.

,

!

,

U.S. NUCLEAR REGULATORY COMMISSION

[

REGION I

~

Report No:

50-29/90-04

!

Docket No:

50-29 Licensee No:

DPR-3 Licensee:

Yankee Atomic Electric Company

'

580 Main Street Bolton, Massachusetts 01740-1398 Facility Name: Yankee Nuclear Power Station

. Inspection at: Rowe, Massachusetts Dates:

February 13 - April 2,1990 Inspectors:

T. Koshy, Senior Resident Inspector

'

ey, ident Inspector Approved By-A e h 30- EO

..

, PalTiak~, Acting Chie N Date Reactor Projects Section No.'3A Inspection Summary: Inspection en February 13 - April 2,1990 Report No.- 50-29/90-04 Areas Inspected:.-Routine inspection on' daytime and backshifts by two resident. inspectors of: operational safety; security; plant operations; maintenance and surveillance; engineering support; radiological controls; actions on previous inspection findings; licensee event reports; licensee

-

response to NRC initiatives; and,' periodic reports.-

Results:

- 1.

Gener41 Conclusions'on Adequacy. Strength or Weakness in Licensee Programs The licensee decision to replace the safety injection tank demonstrated a-sound commitment to plant. safety.

Electrical maintenance efforts were noteworthy lin supporting plant operations through EDG breaker maintenance.

I 900511o075 900430

{DR ADOCK 05000029 PDC

-

h

,

e

.

<

y.n c.

t

.

'

_st

-:*1 i

,

Y [ kt

,!

.I

'

"

<

,..

.

.

'

!

l-i 2. -

..

.

i

<

.

Violations-i

!

Two non-cited violations were identified during this inspection.

'

period.

j

':

,

Improper contamination control (Section 6.2).

'i

---

-- ' An incomplete background check (Section"10.2)

.

.

'

'3.

Unresolved Items

_;

-l One unresolved item was identified during this inspection pc.aiod:

j

The omission of safety-related systems in response to NRC Li c

--

e Bulletin 79-14.

i

!

'

!

.

ll

,- ;

!

"

,

r

'i i

!

-

!

_

.

i

.' f

,

'

-

i

.[

'

w

~

.

s ' >

f

).-

'

i

.

L i

k t

!

'T1

4

- ----_-- -_-__. _ - _ _ _ - _

,

c..

,1 x

..

.

+

TABLE OF CONTENTS f

PAGE 1.

Executive Summary...................

2.

Persons Contacted....................

3.

Summary of Facility Activities............

,

4. -

Operational Safety Verification (IP 71707).......

4.1 Plant Operations Review

............

4.2 Safety System Review...............

4.3 Review of Temporary Changes,. Switching and.

Tagging......................

4.4

_ Storage of Transient Equipment in Safety Related

'

Areas (RI-TI 87-03)_...............

r 5.

Plant Operations'(IP 71707, 93702)...........

5.1 Plant Load Reduction for Condenser Tube Testing.

5.2 Missing spent Fuel Pit Cooling Pipe Support..

5.3 Loss of Power to 480V Emergency Bus No. 1.... 22

.

5.4 Safety Injection Tank Replacement.......

.

6.

Radiological Controls (IP 71707)...........

6.1 Radiologica1' Incident During Safety Injection Pump Surveillance Test.............._.14 6.2 Radwaste Container Surveys...........

7.

Maintenance / Surveillance (IP 61726, 62703)......

'7.1 Emergency Diesel Generator Breaker Maintenance.

8.

Emergency Preparedness Drill (71707).........

9.-

Engineering and Technical Support (71707, 62703)

..

9.1 -

Defects.in Steam Generator No. 4 Girth Weld..

9.2 Cracks in Fuel Transfer Chute Concrete....... 19 10. Security (IP 71707)....-

.............

.

10.1 Observations of Physical Security...

.

...

10.2-Unauthorized Access to the Protected Area

...

_

10.3 Fitness-For-Duty (FFD) Training

...... -..

4

- 11

' Licensee Event Reporting (LER) (IP 90712).......

11.1 LER 90-$01....................

. 12.

Review of Periodic Reports (IP 90713)

........

13. Management Meetings (IP 30703)......

.....

• The NRC Inspection Manual-inspection procedure (IP) or temporary i

Instruction (TI) used as inspection guidance is listed for each

!

applicable report section.

l i

i l

i l

l

'

.

-,

.. -

,,

4.

..

,

'#

i Executive Summary Plant Operations (Module 71707)

Yankee Nuclear Power Station (YNPS, Yankee or the plant) maintained continuous power operation throughout the inspection period.

The licensee reduced power to about fifty percent on March 16-18, 1990, to perform i

condenser leak testing and tube plugging.. The plant operated at full

.,

- rated power for the remainder of the inspection period.

Safety Assessment / Quality Verification (Module 71707)

l On March 9, 1990, the utility board of directors authorized replacement of l

the safety in,Jection tank which was identified as having a significant weld i

defect in January 1990.

Replacement is scheduled for the June 1990 refueling outage.

'

Radiological Controls (Module 71707)

An NRC identified non-cited violation was documented involving an I

auxiliary operator who failed to adhere to the radiation work permit.

'

The licensee was prompt and effective in resolving the issue within the

inspection period, i

Maintenance and Surveillance (Modules 61726,62703)

Surveillance activities performed by instrumentation and controls (I&C)

,

personnel were effective, Electrical maintenance effectively supported plant

~

operationt in performing EDG breaker maintenance.

Some weakness was noted in the licensee lack of spare parts for those breakers, j

Emergency Preparedness (Module 71707)

i The operations staff' properly evaluated emergency action levels (EALs)

during the March 8, 1990, loss of the No. 1 Emergency Bus.

The licensee conducted an EP drill on February 27, 1990 to improve facility communications and to familiarize personnel with EP procedure changes.

Security (Module 71707)

"

The licensee identified a violation involving an individual, with an t

incomplete background investigation, who was granted unescorted access to

'

- the protected area. The NRC determined that the response to this incident was (

prompt and thorough.

'

t i.

I C

L

!

'.

-

c+

ns

.

,

':es

-<

xy

_

H, ? ',. t

'$

'M(h i!

.

I t.

- :.

i-r c..

,

Engineering and Technical Support (Modules 71707,62703)

<

In' response to NRC questions it was noted that the licensee has identified lower girth weld defects in the No. 4 steam generator. The licensee has not evaluated the other steam generators or the upper girth weld in No. 4

.

steam generator.- The licensee plans to perform further evaluations during the June 1990 refueling outage.

,

The inspector. identified 'an unresolved item regarding the omission of

!

-safety-related systems evaluated in' response to NRC Bulletin 79-14.

<

i;

. Identified deficiencies and corrective actions warrant management

"[

k+ ~

attention,

-

'

y l'

.

!

i

.!

s

-i

.

!

!

'

l-f k

i

!

J k

.

,

p

!

[

,

.-

i h

f

.

,

x

,

~.,

-

_

-.

-

^

'

.

!

.. _-

,

E'

e i.

e

.-

.

.

I

!

t L

.

'

f DETAILS

=2.

Persons Contacted E

Yankee Nuclear Power Station t'

,

$

T. Henderson, Plant Superintendent.

k R. Mellor, Technical Director

.;

Yankee Atomic Electric Company (YAEC)

N. St. Laurent, Manager of Operations i

The _ inspector also interviewed other licensee employees during the

..

inspection, including members of the operations, radiation protection,

.

chemistry, instrument and control, maintenance, reactor engineering, l

"

security, training, technical services and general office staffs.

l n

3.

Summary of Facility and NRC Activities i

Yankee Nuclear Power Station (YNPS, Yankee or the plant) maintained

continuous power operation throughout the inspection: period.

The

,

licensee reduced power to about fifty percent on March 16-18, 1990, to-perform condenser leak testing and tube plugging. The plant operated at full rated power for the remainder of the inspection period.

Throughout the majority of this period one resident inspector was i

assigned. A second inspector, Mr.-Thomas Koshy, assumed the duties of Senior Resident Inspector _ effective April 8,1990.

l Effective March 18, 1990, Mr. P. K. Eapen completed his assignment as l

.

the cognizant NRC Region I Division of Reactor Projects (DRP) Section Chief for YNPS. Mr. John F. Rogge, Regional Coordinator in NRR and

.

'

L former Senior Resident Inspector in NRC Region II was selected as

'

.;

cognizant DRP Section Chief for YNPS.

In the interim prior to Mr.

Rogge's arrivali Mr. Walter J. Pasciak, Facilities Radiation

-!

Protection Section Chief in the Division of Radiation Safety and Safeguards, assumed the Section Chief responsibilities-for YNPS.

l On March 20, 1990, Mr. Patrick M. Sears, Licensing Project Manager i

in NRR responsible for YNPS visited the site and supported the NRC resident inspector with backshift operational safety verification inspection.

,

f

i

!

,

- -. _ _ _.

. _ _ _.

,,

t

.

e

.

.,

i

-

.

'

4.

Operational Safety Verification 4,1 Plant Operations Review

The inspector observed plant operations during regular and backshift tours of.the following areas:

i Control Room Safe Shutdown System Du11 ding Primary Auxiliary Building Fence Line (Protected Area)

>

Diesel Generator Rooms Intake Structure

.

Vital Switchgear Room-Turbine Building

~

Cable Tray House Spent Fuel Pit (SFP) Building

,

Safety Injection Building

The following items were checked during daily routine facility

.

l tours: shift staffing, access control, adherence to procedures i

and limiting conditions of operation (LCOs), instrumentation, recorder traces, protectivo systems, control room annunciators, area radiation and process monitors, emergency power source

operability, operability of the Safety Parameter Display System

(SPDS), control room logs, shift supervisor logs, and operating orders.

On a weekly basis, selected Engineered Safety Feature l

(ESF) trains were verified to be operable.

The condition of

plant equipment, radiological controls, security and safety were assessed. On a biweekly frequency, the inspector reviewed safety-related tagouts, chemistry sample results, shift turnovers, portions of the containment isolation valve lineup

-

and the posting of notices to workers.

Plant housekeeping and fire protection were also evaluated.

Inspections of the control room were performed on weekends and i

backshifts as follows: February 19, 20,.25, 26, 28 and March-8,

,

9, 10, 20, 24, and 26. Holiday inspection was performed from

'

10:30 a.m. to 1:30 p.m. on February 19. Operators and shift supervisors' were alert, attentive and responded appropriately to annunciators and plant conditions.

Cognizant shift personnel were knowledgeable of plant conditions and ongoing maintenance and surveillance activities.

Shift turnovers were conducted professionally with strong personnel access control practices.

Shift documentation adequately characterized operating history and the observed off-normal o

conditions, such as equipment problems, were resolved in a timely manner.

One noted exception to the normally high quality of control room decorum was observed on March 7, 1990, when there was approximately twenty people in the control room.

This was due, in part, to the number of training department personnel who i

~

.

.

.

'4

.

.

!

., -

.

!

i i

)

were establishing on-shift hours for senior reactor operator

'

qualifications.

This occurred at a time when several instrumentation and control (I&C) personnel were performing

,

corrective maintenance on process and area radiation monitoring systems. When identified by the NRC, the shift supervisor

promptly directed unessential personnel to leave the area.

No subsequent deficiencies were identified.

4.2 Safety System Review The emergency diesel generators, EDG fuel' oil, residual heat removal, and safety injection systems were reviewed to verify

'

proper alignment and operational status.

The review included verification that (1) accessible major flow path valves were a

correctly positioned, (ii) power supplies were energized, (iii)

l lubrication and component cooling was proper, and (iv) components

[t.

were operable based on a visual inspection of equipment for

,

leakage and general conditions.

System walkdowns to assess the material condition of the ECCS, HPSI, LPSI and the low pressure.

-

safety injection accumulator were performed.

Selected accessible valves were verified to be in the correct position and locked i

when required by plant procedures.

l No unacceptable conditions were identified

-

4.3 Review of Temporary Changes Switching and Tagging

'

Temporary change requests (TCRs), which were approved in support of implementing lifted leads and jumper requests and

-

mechanical bypasses, were reviewed to verify that: controls established by AP-0018, " Temporary Change Control," were met;

'

no conflicts with the Technical Specifications were created; the requests were properly approved prior to installation; and a safety evaluation in accordance with 10 CFR 50.59 was prepared if required.

Implementation of the requests was reviewed on a

,

sampling basis.

[

The switching and tagging log was reviewed and tagging activities were inspected to verify plant equipment was

>

controlled in accordance with the. requirements of AP'0017,

'

" Switching and Tagging of Plant Equipment."

Licensee administrative control of off-normal system configurations by the use of TCR and switching and tagging i

procedures as reviewed above, was in compliance with procedural instructions and was consistent with plant safety.

No unacceptable conditions were identified.

,

!l;

.

..

_ ;

-8 s.:

-

p.

..

I

.

8

.

('

L 4.4 Storage of Transient Equipment in Safety Related Areas (RI TI

'

87-03)

On March 5, 1987, NRC Region I issued Tempore.ry Instruction No.

87-03, " Storage of Transient Equipment in.c fety Related Areas" l

a to review licensee equipment storage programs and practices which

!

have the potential to adversely impact suety-related equipment.

'

As defined in the instruction, transient equipment includes:

,

dollies, block and tackle, filled gas bottles, heavy equipment (stationary or on rollers) such as welding machines or tool F

cabinets, scaffolding (stationary or on rollers), and temporary i

office spaces along with housed furniture, cabinets, etc.

,

Specific inspection guidance included reviewing the licensee internal response to NRC Information Notice No. 80-21, " Anchorage

}

and Support of Safety-Related Electrical Equipment," review for i

adequacy of established administrative controls and procedures, and verify implementation of programmatic controls.

The inspector reviewed licensee procedures AP-0041, Rev. 10,

.

" Plant Inspection Schedule" and Ap-0216, Rev. 5, " Plant Housekeeping Program." These procedures provide adequate i

guidance to control transient equipment in safety-related areas.

'

Minor deficiencies identified during plant tours were promptly

.

corrected by.the licensee.

No safety significant deficiencies

'

were identified.

.

Inspector review of quality assurance surveillances and plant audits indicated that the licensee was implementing the established procedures. Deficiencies noted during these surveillances were adequately documented for corrective action.

,

Yankee Rowe is a Systematic Evaluation Program (SEP) facility.

>

The safe shutdown system (SSS) provides seismic safe shutdown capability at YNPS.

No deficiencies were identified in areas impacting SSS structures, systems or components.

5.

Plant Operations

~5.1 Plant Load Reduction for Condenser Tube Testing

,

L The licensee reduced power to approximately fifty percent on

L March 16 - 18, 1990, to perform helium leak testing and tube plugging.

One tube in the east waterbox and five tubes in the l

west waterbox were plugged.

The licensee completed the conder.ser i

work and started increasing plant load on February 17.

Full l

rated power was achieved at.3:50 a.m. on February 18.

'

.

w

-.

.

,

.-

,

I

10 No unacceptable conditions were identified during this evaluation.

5.2 Missing Spent Fuel Pit Cooling pipe Support During a routine plant tour on March 5, 1990, the inspector noted

'

a-pipe support discrepancy on the four inch piping line between.

,

the shutdown cooling heat exchanger and the spent fuel pit (SFP)

cooler.

Specifically, hanger CRF-H-10 on pipe line CFR-152A-2 was missing the U-bolt fastener and associated hardware such that

no support was being provided by the hanger.- This piping system provides cooling to the SFP cooler located in the primary r

auxiliary building fan room. The piping system and hangers are a Safety Class 3 (SC-3) system.

The licensee evaluation determined that the field installation was inconsistent with equipment specified by plant drawing Nos.

9699-MP-AH-10 and 9699-FP-54A.

Hanger CRF-H-10 is required for

this piping system. Missing support hardware included the U-bolt t

and hexagonal nut fasteners.

'

The licensee performed engineering evaluation / calculation YRC-795 which concluded that the pipe was not overstressed and was at 61 percent of the stress allowed by ANSI B 31.1, 1977. Adjacent hangers CRF-H-9 and CRF-H-13 were determined to independently provide adequate support for the pipe in the absence of CRF-H-10. A visual examination determined that no evidence of piping deformation existed.

The cause of the missing support fasteners is unknown.

The licensee stated that the condition may have existed since initial plant construction.

The licensee initiated nonconformance report-(NCR) No.90-009.

The plant operations review committee (PORC) reviewed and

approved the NCR on March 13, 1990.

Included in the NCR was i

acceptance of the engineering evaluation, root cause analysis,

corrective actions and actions to prevent recurrence.

brrective

,

actions entailed issuing maintenance request (MR) No.90-517 to i

install the pipe support.

The licensee examined the adjacent

!

pipe supports to ensure the pipe was not in an overstressed

, _

condition, PORC also discussed previously identified pipe support l

anomalies and concluded that the other incidents were unrelated.

The licensee determined that no 10 CFR 21 reporting was required, The licensee was prompt in addressing this issue. The

,

maintenance support department (MSD) and the Yankee Nuclear

!

l:

Services Division (YNSD) provided a technically sound evaluation

'{

of the as-found conditions and corrective measures resolution.

i Implementation of MR 90-517 was adequate in establishing proper pipe support.

The licensee was unable to install the U-bolt l

,

-

,.

l

.

.

Il i

+

!

hanger as specified in plant drawing 9699-MP-154-AH-10 due to interference.

The licensee installed an equivalent hanger on the threaded rod and angle iron design per plant drawing

>

9699-MP-154-AH.

No unacceptable conditions were identified

.in the design or field installation.

[

The inspector reviewed some previously identified hanger

-

discrepancies. The licensee issued the following NCRs in response to licensee identified pipe support anomalies:

NCR 89-016, " Incorrect Support Type," addresses the August

--

'

18, 1989, identified pipe support discrepancy where MR

89-1972 was issued for a missing component on a Unistrut

,

support for the six inch shield tank cavity (STC) fill and drain line PRT-302A-1.

Licensee review determined the

!

installation did not conform to piping diagram 9699-FP-46B and hanger drawings that specified variable spring type

,

hangers.

Actual supports instal. led are Unistrut dead weight

>

supports.

NCR 89-017, "SI System Pipe Supports Do Not Meet Design

--

Drawing Configuration," addresses the October 20, 1989,

identified discrepancy where personnel performing Quality assurance (QA). surveillance on the safety injection (SI)

,

system noted that seven out of ten pipe supports examined

'

did not meet the design drawing configuration.

Specific

discrepancies included: weld related deficiencies, member orientation and added material.

NCR 90-003, " Support CRCH-5-2 Does Not' Match Drawing

--

MP-144-CH-10,"' addresses the January 22, 1990, identified discrepancy where the charging system pipe support stiffener

~

plate to base plate weld lacked the required fillet weld thickness.

Both NCR Nos.89-016 and 89-017 questioned the adequacy of the response and documentation derived from the vendor performed walkdown in response to NRC Inspection and Enforcement Bulletin (IEB) No. 79-14, " Seismic Analysis for Safety-Related Piping

Systems." YNSD evaluation YRP 170/90, dated February 7, 1990,

,

examined.the vendor suppl.ied analysis, plant design changes,

,

drawing configurations and installed conditions for the identified nonconformances.

The licensee evaluation concluded that there is reasonable assurance that piping and drawings prepared by the vendor for their work scope for IEB 79-14 provide accurate representation of actual field conditions.

YNSD similarly concluded that there was reasonable assurance that piping and pipe support ~ drawings for the SI system inside the

,

,

.- ',;

c

-.

.

-

.

vapor container (VC) provide an accurate representation of actual field conditions.

.

Although YNSD determined that adequate support existed for the nonconformances identified by NCR Nos.89-016 and 89-017, YNSD recommended field verification of SI and cavity fill. system

piping outside the VC, including the NCR identified supports, to

"

verify supports and to ensure accurate representations of field conditions. The safety related piping outside of the VC were

excluded from the 1979 inspection scope even though IEB 79-14

addressed all the safety related systems. The NRC inspector was concerned about the above omission of systems outside VC.

The-licensee is reviewing the historic data to support the adequacy of the response to the bulletin.

This item is unresolved (50-29/90-04-01).

The NRC subsequently identified a missing pipe support CRF-H-10 on a fourth plant system which may have existed since original plant construction.

Additionally, the NRC identified a missing hexagonal fastener on the adjacent hanger CRF-H-13 following licensee evaluation and corrective measures for NCR 90-009 YNSD recommended that a service request be written to perform these actions..The licensee stated that service request No. 90-27 was initiated with implementation scheduled for late 1990.

YNSD evaluation YRP No. 170/90 documented plant drawings which were not updated to reflect as-built conditions and design changes, a vendor pipe analysis supporting equipment different than what was installed, and contradictory plant drawing data. However, the licensee f.echnical evaluation determined that the piping and support evaluations remain valid.

This area requires additional management attention.

5.3 Loss of Power to 480V Emergency Bus No. 1 At 10:58 a.m. on March 8, 1990, during surveillance OP-4204, " Test of Operation of the Safety Injection Pumps and Determination of ECCS Subsystem Leakage," power was lost to the No. 1 Emergency Bus when bus tie breaker BT-1B unexpectedly tripped following manual trip of the No. I high pressure safety injection (HPSI)

pump at the conclusion of testing. When BT-1B opened, the feeder breaker BT-1A, which connects the offsite power through Bus 6-3, opened and the No I emergency diesel generator (EDG)

automatically started.

However, the generator output breaker EG-1 failed to close resulting in deenergization of the Emergency Bus (EBus).

The operating shift responded to the alarm

!

indications caused by the deenergized bus, placed the EDG control switch in the "0FF" position, reset and closed the BT-1A breaker, and manually closed the BT-1B breaker to restore power to the EBus through the normal power supply.

The licensee declared

.

~

i.

.

.

.

.- ;

.

.

.

!,

!

L the No. 1 EDG inoperable and entered the TS 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limiting condition for operation (LCO) action statement. At 1:39 p.m., a four. hour report was made to the NRC Operations Center via the emergency notification system (ENS) per 10 CFR 50.72 (6)(2)(11)

as determined by event reportability evaluation report (ERER) No.

>

90-09.

Emergency action. levels (EALs) were evaluated and the licensee determined that the incident did not constitute

[

declaration of an emergency classification.

'

The momentary loss of power to the EBus resulted in the following equipment losses: the No. I emergency motor control center j'

(EMCC-1), the No. I safety injection train, primary control rod position indication, control rod drive mechanism, and'the inplant Gaitronics communications system.

Power was restored to the Emergency Bus at 10:59 a.m.

Licensee testing of tie breaker BT-1B was inconclusive in determining why the breaker tripped open.

Repetitive tests were unsuccessful in recreating the trip condition.

No deficiencies in operator actions were identified.

The root cause could not be determined.

To test the output breaker EG-1, Emergency Bus 1 needed to be deenergized.

Emergency motor control center No.1 (EMCC-1) is normally aligned to EBus 1.

To troubleshoot EG-1, the licensee proposed aligning EMCC-1 to the No. 2 Emergency Bus. During the time needed to transfer the EMCC-1 loads from EBus I to EBus 2, the equipment lost during the above described incident would'

again be unavailable. YNSD performed a technical and safety evaluation YRP 264/20 for this proposed action. The operations technic 61 staff drafted procedure OP-2000.248, Original Revision,

"No. 1 Emergency Power Train Functional Test," to perform the evolution.

The plant operations review committee (PORC) approved-the YNSD evaluation and the procedure on March 9, 1990.

At 3:14 p.m. the same day, the licensee switched EMCC-1 to EBus 2.

The transition time was approximately two minutes. The licensee considered this a momentary transition of plant loads which did not necessitate entry into TS 3.0.3 for loss of primary rod position indication as required by TS 3.1.3.2.

Maintenance performance in support of this evolution was a

noteworthy. A more detailed description of maintenance activity is described in Section 7.1 of this report.

Following maintenance and testing, the licensee restored EMCC-1 to the normal power' supply EBus 1 at 4:17 a.m. on February 10.

'

This transition was performed in approximately one minute. The

.

l'

.

.

.

.

.

,

.

.

.

licensee completed OP-2000.248 and exited the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> TS action statement at 6:45 a.m. the same morning.

Inspector review noted good licensee performance.

Control room personnel restored the momentary loss of electrical loads in a

,imely and calm manner. The licensee demonstrated a strong safety perspective in planning and implementing OP-2000.248.

The procedure provided the operation with the proper guidance for the evaluation including entry into TS 3.0.3 if problems were encountered in switching EMCC-1 to EBus 2.

Inspector review noted licensee planning and actions were consistent with agency guidance on TS 3.0.3.

No unacceptable conditions were identified in event classification and reportability.

5.4 Safety Injection Tank Replacement On March 9, 1990, the utility board of directors approved replacement of the safety injection tank (SIT) which was identified as having a significant weld defect in January 1990.

Replacement is scheduled for the June 1990 refueling outage.

Details of the tank defect and licensee actions are documented in NRC Inspection Report Nos. 50-29/90-01 and 50-29/90-03.

The decision to replace the SIT demonstrates a sound commitment to plant safety.

6.

Radiological Controls Radiological contrels were reviewed on a routine basis relative to industry radiological standards, administrative and radiological control procedures, and regulatory requirements.

Selected work evolutions were observed to determine the adequacy of program implementation commensurate with the radiological hazards and importance to safety.

Independent surveys were performed by the inspector to verify the adequacy of radiological controls and instructions to workers.

Radiation protection (RP) program implementation was generally good.

Radiation work permits (RWPs), radiological surveys and air sampling adequately characterized the radiological hazards. Work controls implemented by the RP staff was effective in providing a high level of radiological safety.

No deficiencies were identified in radiological posting and labeling.

6.1 Radiological Incident During Safety Injection Pump Surveillance Test One exception to the consistent radiological work controls routinely observed by the inspector occurred on March 15, 199. - _

_

__

__

-

+

.

j

...

.

..

,

'

i During surveillance testing of the No. 2 high pressure safety

,

injection (HPSI). pump, the inspector observed an auxiliary

operator (AO) repairing a leaking contamination control device

'

(drip bag) inside a posted contaminated area without the protective clothing (cloth gloves and rubber gloves) required by

>

radiation work permit No. 90-0205." Routine Shift Duties-All

Shif ts, Operations and Radiation Protection," for contaminated

,

l areas greater than 1000 dpm. The A0 acknowledged the inspector

'

concern, stopped handling the contamination control device, a

informed the RP staff, repaired the drip bag and cleaned the leakage using the required protective clothing gloves. The RP staff directed the A0 to frisk his hands to determine if they had

'

become contaminated. The results of this survey indicated that the A0 had fortuitously not been contaminated.

The-post surveillance radiological survey indicated that the area

.

smearable radioactivity was less than 1000 dpm/100 cm-sq.

The

liquid ectivity in the SI system was 3.02 E-4 microcuries/ml.

i The inspector noted several other weaknesses associated with this incident.

The subject A0 demonstrated poor radiological safety judgement in the presence of newly hired individual performing A0

duties.

The leaking flange had an equipment maintenance request i

.

(MR No. 89-29) which was outstanding since January 3,1989, and,

-

the contamination control device was poorly craf ted to perform

'

its intended function.

l

.

The licensee promptly issued a radiclogical occurrence report which. adequately characterized the sequence of events, corrective

,

.

actions and actions to prevent recurrence.

The operations shift

'

supervisor (SS) counseled the A0s in the importance of following i

good radiological safety practices-and the consequences of not adhering to the RWP.

The RP staff replaced the control device with an improved design pending completion of the MR.

Corrective actions were effective in addressing personnel performance and radiological safety aspects of this incident.

The licensee completed corrective actions within the scope of this inspection.

Inspector review determined it was an isolated incident and not repetitive as set forth in 10 CFR 2, Appendix C, Section V. A. for a non-cited violation of the RWP procedure

[

(AP-0806) and Technical Specification 6.11 (50-29/90-04-02).

'

6.2 Radwaste Container Surveys

!'

The licensee demonstrated good initiative in the area of radiological surveys for radwaste containers.

The licensee

!

i developed a device which enables personnel to measure exposure

-

'

r'ates at precise distances from the container surfaces.

The

!

l devise also allows the individuals to survey the containers from i

l

C l

I

U

,-

.

.

'

.

e

,

.

16

greater distances thereby maintaining occupational exposure as

'

low as reasonably achievable. Although simple in design, the device is an effective ALARA tool which increases survey precision.

i 7.

Maintenance / Surveillance

The inspector observed and reviewed maintenance and surveillance problem investigation activities to verify compliance with

,

.

l-regulations, administrative, maintenance and surveillance procedures,

,

codes and standards, proper QA/QC involvement, safety tag use.

!

c equipment alignment, jumper use, personnel qualification, radiological

'

controls for worker protection, fire protection, retest and restoration of equipment, deficiency review, resolution and reportability per Technical Specifications.

,

Routine surveillance activities conducted by instrumentation and controls (I&C) personnel were effective in ensuring equipment operability in support of plant operations and other plant programs.

IhC personnel routinely demonstrated a proper questioning. attitude in evaluating equipment performance.

Individuals secured proper authorization to perform surveillance tests and maintained continuous

communication with-the control room for critical tests to reactor protection systems (RPS) and engineering safety feature (ESF) systems.

s 7.1 Emergency Diesel Generator (EDG)

_

During a plant event on March 8, 1990, the No. 1 EDG bus tie P eaker BT-1B tripped open unexpectedly during surveillance testing of the No. I high pressure safety injection (HPSI) pump.

Also, the EDG output breaker EG-1 failed to automatically close

as designed.

Both of these breakers are General Electric type

.

AK-2A-50S units.

Plant operations details of this event are described in Section 5.3 of this report.

The licensee issued maintenance request (MR) No.90-499 to perform troubleshooting maintenance on the breakers.

Repetitive-inspection and testing of BT-1B was inconclusive in determining why the breaker tripped open.

All licensee attempts to recreate the trip condition were unsuccessful.

The licensee concluded that BT-1B was functioning properly.

The licensee experienced difficulty in troubleshooting EG-1.

Electrical maintenance efforts proceeded into the early morning of March 10 when personnel determined the failure was due to i

compression of the moveable main contact crossbar buffer * assembly washers. These washers were determined to have caused the main contact crossbar to become out of adjustment resulting in the breaker tripping free upon receipt of the closing signal.

The i

L i

LE

.

r

.

S

,

{

..

.

,

o i

-

.,

licensee stated that this conditicn caused a binding on the breaker trip shaft thereby preventing the trip shaft from

,

resetting.

The crossbar buffer assembly was adjusted and the breaker operated satisfactorily. The licensee-secured the services of a vendor representative who assisted in the troubleshooting and problem resolution. The licensee plans to

,

revise procedure OP-4506, " Inspection of ECCS Circuit Breaker,"

,

to improve the inspection process to include measuring the crossbar buffer assembly position.

The licensee issued MR No.90-516 to inspect the other generator

,

output breakers EG-2 and EG-3.

The licensee determined that similar conditions did not exist on either breaker.

'

However, a damaged cutout switch was identified on EG-2.

The licensee declared the No. 2 EDG inoperable at 5:35 a.m. on March

'

13 pending replacement of the switch.

The licensee entered the appropriate 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.

No replacement switch was

-

available in the plant storeroom.

The licensee obtained a

commercial grade replacement switch from the vendor.

Onsite

testing was performed pending receipt of the vendor qualification test documentation.

The cutout switch was installed, the breaker was' tested, and the EDG train was returned to service at 7:15 p.m.

~;

on March 15.

t Inspector review noted no deficiencies in the licensee technical evaluations and maintenance activities.

The ability of the electrical maintenance staff to perform onsite maintenance of emergency breakers is a licensee strength.

Limited reliance on contractor / vendor support is noteworthy.

The electrical maintenance supervisor demonstrated a high level of involvement and dedication throughout these activities.

No deficiencies were

'

identified in the qualification of commercial grade parts. Quality assurance personnel were noted to provide independent assessment of selected critical phases.

t One area which warrants further licensee ' evaluation is the

availability of replacement parts for these breakers.

The licensee does not have replacement breakers to support operations

,

and maintenance.

Therefore, the onsite supply of replacement parts is critical.

For the repairs of EG-2, the licensee

-

expended approximately 62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br /> of a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> TS LCO action statement which would have required a plant shutdown.

The lack of onsite repl.acement parts provides an unnecessary challenge to

.;

plant personnel and constitutes an operational limitation.

• Otherwise, licensee efforts in this area are exemplary, l'

W

_ _ _ _ _ _

,

.

,

o:

.

.

.

." o

,

'

i 8.

Emergency Preparedness (EP)

[

On February 27, 1990, the inspector observed portions of an emergency preparedness drill conducted to test Emergency Response Facility (ERF)

communications. The drill objectives were to improve communications during an emergency, to familiarize personnel with EP procedure

,

changes and revised data sheets, and to complete post-accident sampling requirements. The licensee conducted a predrill briefing to review major procedure changes, drill objectives and initial

'

conditions. This drill did not involve. control room participation.

The licensee conducted' the drill in an orderly, professional manner, i

Technical support center (TSC) and operations support center (OSC)

status boards were updated consistent with the progress of the scenario.

"

9.

Engineering and Technical Support 9.1 Defects in Steam Generator No. 4 Girth Weld In response to steam generator girth weld cracks identified at another NRC Region I licensed facility, the inspector questioned

the licensee about steam generator girth weld evaluations

performed at Yankee.

The maintenance support department (MSD)

+

provided the inspector with a timely and detailed explanation.

The licensee has identified defects on the.No. 4 steam generator.

'

lower girth weld. The inspector reviewed the following licensee technical evaluations and correspondence associated with this issue:

MSG Memorandum 385/85, dated November 20,~1985, Indication

--

. Evaluation, Yankee Plant Steam Generator #4; MSG Memorandum 111/87, dated May 29, 1987, Indication

--

Evaluation Follow-up: Steam Generator #4, Weld SG-4-4;

+

PED Memorandum 314/88, dated July 29, 1988, Service Request

--

88-149, Steam Generator Transition Weld UT; MSD Memorandum 94/88, dated August 30, 1988, Inspection of I

'

--

Steam Generator Transition Weld SG-4-4;

,

PED Memorandum 372/88, dated September 20, 1988, Steam

--

'

-Generator Transition Region Weld; PED Memorandum 30/89, dated January 27, 1989, Steam

--

Generator Weld Ultrasonic Evaluation; and YRP Memorandum 1202/89, dated August 28, 1989, Steam

--

Generator Welds.

.

'

<

O

g

'

,

.

<

s i

'

.

..

.;

'

e..

,

i E

L-

,

The latest licensee evaluation YRP 1202/89 concluded that all

,

'

metallurgical and geometric conditions were present at fabrication, no apparent changes have taken place since

monitoring began, the identified defects are not. service related, and monitoring needs to continue in the future.

Inspector review noted some weakness in the licensee evaluation, f

Only steam generator No. 4 has been examined for girth weld

-

defects. Also, only the lower girth transition weld in the No. 4 steam generator has been evaluated. Most of the defects

!

identified at the other Region I licensed facility were in the upper girth weld. Additionally, the above listed memoranda

describe technical difficulties in performing certain examination

,

techniques in the Yankee steam generators.

g i

The inspector discussed these concerns with the licensee who acknowledged the need for additional examination.

The licensee

,i stated that additional evaluations would be performed during the

'

June 1990 refueling outage.

!

-

9.2 Cracks in Fuel Transfer Chute Concrete During a routine plant.walkdown on February 8, 1990, the inspector noted some cracking in the fue' transfer chute concrete.

The inspector questioned the licensee regarding this i

observation.

The licensee stated that the cracking was previously identified and evaluated in October 1989 by the YNSD structural staff in support of the plant life extension (PLEX)

project.

- l The licensee structural evaluation determined that the defects F

were surface cracks. The white surface deposits were attributed to calcium leaking from the concrete.

The observed rust was determined to have come from external metal components and not from exposed metal rebar. The licensee plans to repair the surface cracks following the June 1990 refueling outage.

The licensee plans to perform additional assessment and maintenance as dispositioned in the PLEX project.

Inspector review noted the licensee technical evaluation was sound. The concrete serves as radiological shielding for the transfer of spent fuel assemblies.

Inside the concrete shielding

'

is the steel transfer integrity chute, No unacceptable conditions were identified in the licensee assessment or corrective action plans.

..

L

f.w

.

.

F.'

.

,

~

,.

?

.. s L

10.'

Security 10.1 Observations of Physical Security Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accordance with the security plan and approved

_

procedures. This review included the following security

'

measures: guard staffing, vital and protected area barrier integrity, maintenance of isolation zones, and implementation of access controls including authorization, badging, escorting, and searches.

No inadequacies were identified.

The licensee was effective in analyzing and trending security events.

The f.ecurity. staff compiled event data for 1988 and 1989. Equipment failures and personnel errors were compared with industry averages to identify areas where the security program lagged behind the industry or otherwise needed improvement.

Similarly, areas of noteworthy performance were highlighted.

10.2 Unauthorized Access to the Protected Area At 4:45 p.m. on February 15, 1990, the licensee reported to the NRC: Operations Center via the emergency notification system (ENS)

that a new company employee with an incomplete background investigation had plant protected and vital area access for approximately twerity-six minutes on February 13.

The individual did not access any vital areas. At 3:45 p.m. on February 15, the licensee discovered the error and immediately terminated the authorization.

The lic;nsee issued licensee event report (LER) No. 90-S01 which detailed the sequence of events, corrective actions and actions to prevent recurrence.

The licensee determined the root cause was personnel error by the manager who reviewed the records for access control authorization.

The error was identified during a i

routine review of badgin0 records.

Corrective actions and

,

actions to prevent recurrence involved revising the administrative control forms to provide guidance designed to

'

improve personnel attention to detail through required checklist-l

-

item completion.

l

'

The licensee was effective in self-identifying the incident, in evaluating programmatic guidance and in evaluating personnel performance.

Licensee corrective actions and actions to prevent recurrence were adequate.

Inspector review of this incident, the associated LER (Section 11.2), and NRC enforcement guidance contained in 10 CFR 2, Appendix' C, V.G.1 determined that this

.

-

{.-

,

,.

..

-.

s

,

!

incident met the criteria for a non-cited violation of Sections

1.3.1 and 1.6 of the Security Plan (50-29/90-04-03).

'

10.3 Fitness-For-Duty Training On February 20, 1990, the licensee conducted FFD drug

..

paraphernalia identification training for the plant staff. The

,

licensee secured the services of a contractor who presented the training as an open dialogue drug awareness seminar. The v

following issues were presented and discussed:

Common drugs of abuse and how to identify them;

'

--

Identification of drug packaging and paraphernalia;

--

Recognition of drug concealment techniques; and

.'

--

Recognition of symptoms of drug use, possession and sale.

--

.

In addition to seminar discussion, examples of drugs and drug paraphernalia were displayed.

Plant security maintained line-of-sight cognizance over all controlled substances which

,

were encapsulated in heat sealed containers.

The licensee demonstrated good initiative in continuing training efforts in the newly implemented FFD program.

Security program y

and regulatory requirements were adequately addressed in

controlling training materials.

No unacceptable conditions were

identified.

11.

Licensee Event Reporting (LER)

The inspector reviewed the below listed licensee event reports (LER) to determine'that with respect to the general aspects of the event:

(1) the report was. submitted in a timely manner; (2)

description of the events was accurate; (3) root cause analysis was

.I performed;-(4). safety implict.tions were considered; and (5) corrective'

-

actions implemented or planned were sufficient to preclude recurrence of a similar event.

~

11.1 LER 90-S01, " Unauthorized Access to Projected Area," addresses the February 15, 1990, incident where a new Yankee employee had

-

unescorted access to plant protected and vital' areas without a complete background investigation for approximately twenty-six minutes.

The licensee determined the root cause was personnel error by the manager who reviewed the re::ords for access

'

authorization. The licensee revised administration control forms to provide guidance designed to improve personnel attention to detail through required checklist item completion. No deficiencies were identified in the licensee response to this incident.

,

L

.

"

>

-,

-_

-

. - _ _.

-

. _

_ _ _ _

=>

.

i a

L o

,

,.

>

-

12.

R_eview of Periodic Reports Upon receipt, the inspector reviewed periodic reports submitted.

,

pursuant to Technical Specifications. This review verified, as applicable: (1) that the_ reported information was valid and included the NRC required data; (2) that test results and supporting information were consistent with design predictions and performance specification; and (3) that planned corrective actions were acequate

-

for resolution of the problem.

The inspector also ascertained whether

,

I any. reported information should be classified as an abnormal

[.

occurrence. The following reports were reviewed:

.

Monthly Statistical Report for plant operations for the month of

--

February 1990; YAEC Response to Generic' Letter 89-10 " Safety Related Motor

--

Operator Valve Testing and Surveillance, letter 89-185, December 27, 1989; YAEC Response to Station Blackout Rule 10 CFR 50.63 - Supplement

--

I; Proposed Change No. 231 to TS to Allow Utilization and Increased

--

Testing at Power of a New Neutron Flux Instrumentation System, letter BYR 90-004, January 5, 1990; Proposed Change No. 232 to TS Governing the Emergency Core Cooling

--

System, letter BYR 90-006, January 18, 1990; 1989 Annual Report, letter BYR 90-022, February 28, 1990;

--

Semiannual Effluent Release Report, letter BfR 90-021, February

--

28.-1990; Semiannual Effluent and Waste Disposal Report Including-Annual

--

Radiological Impact on Man, letter BYR 90-036, March 30, 1990; Personnel Exposure Report by Duty Function, letter BYR 90-018,

--

February 23, 1990; and YAEC 1989 Certified Financial Statements.

--

No deficiencies or safety concerns were identified.

13. Management Meetings At periodic intervals during this ' inspection, meetings were held l

with senior plant management to discuss the flndings.

A. summary

.of findings for the report period was also discussed at the conclusion

' !

-

of the inspection and prior to report issuance. No proprietary information was identified as being included in the report.

l l

l

_. -