Text

Insp Rept 50-267/88-03 on 880201-29.Violations Noted.Major Areas Inspected:Licensee Action on Previously Identified Findings,Operational Safety Verification,Followup of Unusual Events,Esf Walkdown & Physical Security Observation
	ML20150F008
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/25/1988
From:	Farrell R, Michaud P, Westerman T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20150F000	List: ML20150E999; ML20150F005; ML20150F008;
References
	50-267-88-03, 50-267-88-3, NUDOCS 8804040236
	Download: ML20150F008 (12)
	v • d • e

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

'

. . . .

, , . .-

t .

>

% >

'

APPENDIX B

4

, U. S. NUCLEAR RECULATORY COMMISSION

'

>,

REGION IV ,

1

'

NRC Inspection Report: 50-267/88-03 License: DPR-34 -

.

Docket: 50-267.

Licensee: Public Service Company of Colorado (PSC)

Facility Name: Fort St. Vrain Nuclear Generating Station

Fort St. Vrain (FSV) Nuclear Generating Station, Platteville.

~

Inspection At: .

Colorado I

t

Inspection Conducted: February 1-29, 1988 ;

/f )~8O8

i

Inspectors: [

R. E. Farrell,"Senior Resident Inspector (SRI) Date

'

Y(A/r

F. W. Michaud, Rbsident Inspector (RI)

3*/79

Date

,

!

Approved: } tt) e d 3/2b

D6te'

T.'F. Westerman, Chief

Reactor Projects Section B

'

!

,

W

,

t '

i

8804040236 890331

PDR ADOCK 05000267

0 DCD

_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

. . .

. .

2

Inspection Summary

Inspection Conducted February 1-29,1988 (Report 50-267/88-03)

Areas Inspected: Routine, unannounced inspection of folicwup of licensee

action on previously identified findi;.gs, operational safety verification,

followup of unusual event, engineered safety features walkdown, monthly

surveillance observation, monthly maintenance observation, radiological

protection, and physical security observation.

Results: Within the eight areas inspected, one violation was identified (the

failure to implement and follow procedures for maintenance and operations

activities, paragraph 4).

l

!

l

t

!

I

!

I

l

._--____ _ ___ _ _ _ _ _ _ - _ _ _ _ _ _ - - - _ _ _ _ _ _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _

'. .

'

.

. .

3

DETAILS

1. Persons Contacted

FSV

L. Brey, Manager, Nuclear Licensing and Fuels

"M. Ferris, Manager, Quality Assurance (QA) Operations

C. Fuller, Manager, Nuclear Production

M. Holmes, Manager, Nuclear Licensing

F. Novachek, Manager. Technical / Administrative Services

P. Tomlinson, Manager, QA ,

D. Warembourg, Manager, Nuclear Engineering

R. Williams Jr. , Vice President, Nuclear Operations

J. Reesy, Staff Assistant, Nuclear Engineering

, *F. Borst, Nuclear Training Manager

M. Deniston, Shift Supervisor

S. Hofsetter, Nuclear Licensing

M. Block, Superirtendent, Nuclear Betterment l

L. Scott, Manager, QA Service

R. Sargent, Assistant to Vice President, Nuclear Operai1ons

R. Webb, Maintenance Supervisor

,

The NRC inspectors also contacted other licensee and contractor personnel

during the inspection.

Denotes those attending the exit interview conducted March 8, 1988.

!

2. Followup of Licensee Action on Previously Identified Findings

(Closed) Open Item 267/8507-06: Shorten Time Between Change Notice (CN)

Issue And Notation On Drawing - In some cases, a caution that changes had

. been made under a CN was not reflected on the affected drawings for 30

days or more af ter a CN was issued. This presented a concern that a

modified system or component could be in service for that amount of time

without adequate drawings. By utilizing a computerized document update

information system, the licensee has shortened the time involved to mark

all affected drawings to approximately 1 week, with the drawings in the

2 control room, shif t supervisor's of fice, and records center updated the

same day a CN issue notification is received. The NRC inspector verified

'

these activities are taking place by direct observations and a review of

documentation. This item is closed.

(Closed) Open Item 267/8507-07: Devcon Epoxy Only Qualified to 200 F.

Epoxy used to attach thermocouples to control rod drive assemblies was

qualified to enly 200*F, while actual operating temperatures can exceed

200 F. Two tests were performed by the licensee to establish this

adhesive's acceptability. One test performed under Fuel Handling

Procedure 100-31 involved a visual examination and measurement of force

,

,e, , , e +,,_mmr.n.,rs. -r, ,-g ~ a,---., -----.,--w.m..---,,, e,-e s---,ww,w<- - - , -

__

.. .

. .

4

required to remove the epoxy from a CRD element, which had been subjected

l to varying power operating conditions in the reactor core between 1979 and

l 1984. The second test, T-288, involved subjecting epoxy to greater than

1 300 F temperature and then performing a pull test to verify that

thermocouples remained sufficiantly attached. Based on these tests, the

,

licensee concluded the Devcon ep;xy was acceptable for use in applications

'

up to 300 F. The NRC resident inspectors reviewed the licensee's tests

and evaluation and found them acceptable. This item is closed.

3. Operational Safety Verification

The NRC resident inspectors reviewed licensee activities to ascertain that

the facility is being operated safely and in conformance with regulatory

requirements and that the licensee's management control system is

ef fectively discharging its responsibilities for continued safe operation.

The NRC resident inspectors toured the control room on a daily basis

during normal working hours and at least weekly during backshif t hours.

The reactor operator and shif t supervisor logs and Technical Specification

compliance logs were reviewed daily. The NRC resident inspectors observed

proper control room staffing at all times and verified operators were

attentive and adhered to approved procedures. Control room

instrumentation was observed by the NRC inspectors and the operability of

the plant protective system and nuclear instrumentation system were

verified by the NRC resident inspectors on each control room tour.

Operator awareness and understanding of abnormal or alarm conditions were

also verified. The NRC resident inspectors revieved the operations order

book, operations deviution report (0DR) log, clearance log, and temporary

configuration report (TCR) log to note any out-of-service safety-related

systems and to verify compliance with Technical Specification

requirements.

The licensee's station manager and superintendent of operations were

observed in the control room on a daily basis, with the superintendent of

operations frequently in the control room during the day and during

special tests or evolutions.

The NRC resident inspectors verified the operability of a safety-related

system on a weekly basis. The PCRV overpressure protection system,

120 VAC vital power distribution system, reactor plant cooling water

system, and firewater system were verified operable by the NRC resident

inspectors during this report period. During plant tours, particular

attention was paid to components of these systems to verify valve

positions, power supplies, and inst umentation were correct for current

plant conditions. General plant condition and housekeeping were

acceptable.

Shift turnovers were observed at least weekly by the NRC resident

inspectors. The information flow appeared to be good, with the shift

l

- . - . .- .

'

.. ..

. .

5

supervisors routinely soliciting comments' or concerns from reactor

operators, equipment operators, and auxiliary tenders.

No violations or deviations were identified in the review of this program

area.

4. Followup of Unusual Event

On February 10, 1988, at 3:47 p.m. (MST), "A" helium circulator tripped

due to a low speed signal with the reactor at 75 percent power. The

circulator trip resulted in a reactor runback to between 50 percent to

60 percent reactor power and then reactor power was further reduced by the

plant operators to 25 percent power. While attempting to balance

feedwater between Loop 1 and Loop 2, an upset in the helium circulator

auxiliaries supplied by feedwater resulted in the tripping of "B" and "D"

helium circuiators at 4:07 p.m. (MST).

The tripping of two circulators (A & B) in one loop r.;ulted in a loop

shutdown (ESF actuation). The reactor operators manually scrammed the

reactor from 25 percent power with only one helium circulator running.

At 6:40 p.m. (MST), the licensee identified that an unplanned release was

occurring and an unusual event was declared. An operator had been

dispatched to vent the surge tank associated with the liner cooling water

system. The licensed operator dispatched to perform this function

inadvertently opened the wrong valve venting the tank to the plant stack

rather than to the gaseous radwaste system. The total release over

approximately 200 minutes was small. (4.26 X 105 microcuries of noble gas

activity)

The plant maintained forced circulation cooling at all timas. The SRI

responded to the event and was onsite all night. The Colorado Department

of Health was in contact with the site and was briefed by the licensee as

well as the SRI.

The licensee has subsequently determined that the "A" helium circulator

trip occurred due to an apparent interchange of speed indication signal

cables during a recent equipment calibration. The trip occurred when the

"B" helium circulator was placed in manual control for calibration.

a. Background

The unusual event of February 10, 1988, and associated unplanned

release started with the trip of helium circulator "A".

Helium circulator speed cable daily calibration was in process when

circulator "A" tripped.

When a circulator's speed cables are calibrated, the circuiator is

taken from auto to manual control to minimize the chances of a trip.

.

_ _ _ _ _ _ .

.

. .

6

Helium circulator "A" speed cables had been successfully calibrated

and circulator "A" returned to auto control. Helium circulator "B"

was placed in manual control and calibration of the "B" circulator

speed cables was in process when circulator "A" tripped.

The licensee determined that on February 2, 1988, while calibrating

the speed modules (SM) on circulator "A", SM 2109 could not be

balanced while getting its signal from cable 18194. The technician

decided to check if the problem was in SM-2109 or in the cable 18194.

The licensee suspected the speed problems were in the cables. Seven

spare speed cables are available from each circulator's SM. The

technician unplugged cable 18194 from SM 2109 and plugged in

cable 18133. With cable 18133 installed, SM-2109 balanced and was

left in this configuration by the technician. Cable 18133 does not

sense circulator "A" speed but is a spare speed cable from the "B"

circulator,

b. Design Information

There are two speed indications from each circulator: a steam

turbine speed indication and a water turbine speed indication. The

water turbine speed indicator is much easier to read than the steam

indicator and generally the one the operators use. Since both drives

are on a common shaft, the speed should be the same regardless of

which turbine is driving the circulator.

There are 12 speed cables coming from the speed modules of each

helium circulator. Four of these cables are utilized for speed

control. One cable for steam turbine speed, one cable for water

turbine speed, and two spares.

Eight cables from each circulator are dedicated to the plant

protection system (PPS). Three of these cables are used at one time

(one for each logic channel). Five cables are dedicated spares.

c. Speed Control

The speed control circuitry looks at the water turbine indicated

speed and the steam turbine indicated speed and controls from the

higher of the two indicated speeds (no difference if everything

working correctly).

As long as the "B" circulator speed was less than or equal to the "A"

,

circulator speed, the control system saw no problem and chose the "A"

!

circulator steam turbine speed to control circulator "A" With

l cable 18133 (a "B" circulator speed cable) controlling SM-2109 (the

"A" circulator water turbine SM) the problem arose during calibration

of "B" circulator speed when the "A" circulator was in auto control

and the "B" circulator speed exceeded the "A" circulator speed,

l

1

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

.

. .

_- _ _ _ _ __

.

. '.

. .

7

.When this happened, the control circuit for "A" circulator, selecting

the higher speed indication, selected the "A" circulator water

turbine speed. -This was actually the "B" circulator speed, since a

"B" cable was feeding this speed module. This falsely told the

control circuit that the "A" circulator was running faster than the

control circuit required, so the control circuit began closing the

"A" circulator steam speed valve.

Since the control circuit was actually reading "B" circulator speed

it saw no change in the "A" circulator speed indication and continued

to close down the "A" circulator speed valve. When the "A"

circulator reached the low setpoint of the circulator

speed-to-feedwater flow program, which forces a limit on primary to

secondary flow ratio, the the PPS which was correctly reading

circulator "A" speed tripped the circulator.

d. Findings

The technician calibrating the SM was utilizing licensee

Procedure SR-RE-17-W, Issue 10, "Circulator Speed Modifier Weekly

Check."

The procedure did not address cable termination.

When the technician removed the installed cable (18133) he was no

longer performing surveillance activities, but was performing

maintenance activities. Maintenance activities are governed by the

licensee's Administrative Prosedure P-7, Issue 12, "Station Service

Request Processing." Procedure P-7 as modified by Procedure

Deviation Request 88-0006, dated January 13, 1988, specifically

states, in Section 2.0, that the procedure applies to corrective and

preventative maintenance and not to calibration activities.

Procedure P-7 is the licensee's procedure for controlling maintenance

activities. Procedure P-7 requires initiation of a Station Service

Request to authorize, document, and control maintenance activities.

Failure to follow Procedure P-7 is an apparent violation of NRC

regulations (267/8803-01).

The operator venting reactor plant cooling water system surge tanks

was guided by System Operating procedure (SOP) 46, Issue 39, "Reactor

Plant Cooling Water System." SOP-46 in Step 3.7, "Venting the Vapor

Space in T-4601 or (T-4602)," details the steps for venting the

reactor plant cooling water surge tank vapor space to the gas waste

system. The steps call for first opening V-4653 for Surge

Tank T-4601 (V-4654 for Surge Tank T-4602). Then the operator is to

open V-461691 for Surge Tank T-4601 (V-461692 for Surge Tank T-4602).

Opening these two valves for each surge tank vents the vapor space of

each tank to a common line leading to the gas waste system. When

these steps are completed, the operator opens Valve V-46193, which

. . . . . .

-____ _ _ _ _ _ _ _ .

. '.

. .

8

opens the common line from the two surge tanks to the gas waste

system relieving the pressure in the , urge tanks.

All of the valves mentioned in the preceding paragraph are manual

valves. Adjacent to the valves, V-461691 on Tank T-4601 and V-461692

on Tank T-4602, are hand operated valves, V-461634P and k-461635P,

respectively. Opening Valve V-461634P after opening Valve V-4653

vents Surge Tank 1-4601 to the plant exhaust stack. Opening Valve

V-461635P af ter opening Valve V-4654 vents Surge Tank T-4602 to the

plant exhaust tank.

The valves are now clearly marked as to function. At the time of the

incident, the valves were marked with small stamped metal tags

identifying the valves by number.

Procedure 50P-46 in Step 3.7 clearly listed the valves to be opened.

The valves were identified in the procedure by valve number

corresponding to the valve numbers attached to the valves. The

operator opened either or both Valves V-461634P and V-461635P, rather

than V-461691 and V-461592. This vented the gaseous content of

Tanks T-4601 and/o" T-4602 to the plant stack resulting in an

unplanned radioactive release. The failure to follow

Procedure 50P-46 is second example of Violation (267/8803-01).

5. Engineered Safety Features (ESF) Walkdown

The NRC resident inspectors performed a walkdown of all accessible

portions of the prestressed concrete reactor vessel (PCRV) overpressure

protection system to verify its operability. Sections 4.3.6 and 6.8 of

the FSAR and Technical Specifications 3.2, 3.3, 4.2.7, and 5.2.1 were

reviewed by the NRC resident inspectors to ensure familiarity with the

system and requirements. The as-found system configuration was compared

with drawing PI-11-5 to check their agreement. Valve positions and

labeling were verified to be correct by the NRC resident inspectors,

including the installation of lotking devices on valves where required.

All cortions of the system were physically inspected, w th the exception

of the internals of the PCRV safety valve tank T-1101 which contains the

relief valves and rupture discs. These components will be inspected

during the next outage when T-1101 is opened. During this inspection,

attention was paid to equipment conditions, housekeeping, and any items

which could degrade performance. The overall condition of this system was

considered good.

No violations or deviations were identified in the review of this program

area.

6. Monthly Surveillance Observation

The NRC resident inspectors observed the licensee's performance of

selected surveillance activities as listed below. The surveillance

procedures were reviewed for conformance with Technical Spe;ification

. .

o .

9

requirements and to ensure they had been properly reviewed and approved

prior to commencing any tests. The NRC resident inspectors witnessed

portions of the preparations, conduct, and v/ stem restoration for each of

these surveillance tests. Test results were independently reviewed by the

NRC resident inspectors to ensure they met applicable Technical

Specification requirements. Surveillance activities observed during this

reporting period included:

SR 5.4.1.1.8.b-M, "Reheat Steam Temperature Scram Test," performed on

February 1,1988. This surveillance tests each hot reheat steam

temperature scram channel to verify alarms, actuations, and

indications. The as-found values were measured and recorded,

acceptance values calculated and independently verified, and

calibration of the bystable amplifiers and thermocouple amplifiers

was checked at 600 F, 900 F, and 1200 F utilizing test signals.

These amplifiers were adjusted as required in accordance with this

procedure and the as-left values were recorded. No discrepancies

were noted.

SR 5.10.8-M, "Monthly Check of Fire Hose Stations," performed on

February 2, 1988. This surveillance verified the condition of each

fire hose station in the reactor and turbine buildings, and was

independently versfied by the NRC resident inspectors. Each

station's hose valve was verified shut and not leaking, hoses and

nozzles properly connected, and general equipment conditions

observed. No discrepancies were noted.

E3R 8.1.lbc-M, "Radioactive Gaseous Effluent Systein Test," performed

on February 25, 1988. This surveillance test verifies the operation

of the gaseous waste release system automatic functions. Instruments

which provide inputs to cause automatic isolation and ventilation

system realignments were tripped using a test signal, then each

associated damper or valve which was repositioned by the automatic

signal was verified to be in its proper position. The instruments

and equipment were then restored to their normal lineup. No

discrepancies were noted.

No violations or deviations were identified in the review of this program

area.

7. Monthly Maintenance Observation

On February 4,1988, the licensee noticed the pressure in the emergency

feedwater supply to the Loop 1 helium circulator Pelton wheel drives was

equal to the feedwater header pressure (approximately 3000 psia). This

condition indicated a problem with Pressure Control Valve PV-21243, which

'

should reduce the pressure to approximately 1700 psi. The licensee took

the emergency feedwater header out of service at 5:57 a.m. (MST), on

February 5, 1988, to perform repairs on PV-21243 and entered Technical

Specification Limiting Condition for Operation (LCO) 4.0.3, since the

conditions of LCO 4.3.4, "Emergency Condensate and Emergency Feedwater

i

- - _ _ _ _ _

'

. .

. ..

10 o

,

Headers LCO," were no longer satistted- LCO 4.0.3 requires the reactor to

be ;hutdown in an orderly manner within a 24-hr Jr period. Also c' :icaale

and providing a 24-hour grace period was LCO 4.2.2.a, "Operable C. culator -

LCO." Repairs were made to valve PV-21243, which included replacement of

the valve trim. The associated pressure controller, PIC-21243, was

calibrated in accordance with Procedure RP-EQ-16, Issue 2, dated

October 15, 1986. The NRC resident inspectors observed the repairs and

calibration, which were completed satisfactorily. No aiscrepancies were

noted. The emergency feedwater header was retu ned to service at

1 a.m. (MST), on February 6, 1988, and LCO 4.0.3 and 4.2.2.a were formally

exited at 5:15 a.m. (MST), after allowing the system to run following its

return to service.

The NRC resident inspectors also followed the licensee's actions to

correct the problems in the helium circu'ator speed cables. The

circulator speed signals to both the indicators and the plant protective

system had been exhibiting erratic behavior at the elevated temperatures

associated with operation at higher power levels. Troubleshooting

following the February 10, 1988, event, described in paragraph 4 of this

report, indicated a problem with the twinax cable "Cannon" connectors at

the helium circulators. These special connectors have the male end

attached to the circulator housing and the female end attached to the

cables. These female pin connectors have a spring-like device which in

some cases had relaxed, allowing a slight gap in the pin connection at the

elevated temperatures. The connectors on each of the four helium

circulators were disassembled and both the nale ar.d fema'e pins were

checked with a micrometer to ensure their size was within a tolerance of

0.060 inch to 0.064 inch. A number of female pins were replaced, and the

connectors reassembled. Since returning to power on February 12, 1988,

the licensee has experienced no significant problems wish the helium

circulator speed cables or the associated indications and protective

circuitry.

4

At 10:40 p.m. (MST), on February 25, 1988, the license 4 experienced a

turbine trip from approximately 50 percent power due to a <=lse low main

steam pressure signal. On investigation, the licensee discovered the root

valve to Main Steam Pressure Transmitter PT-5220 was nearly shut. This

valve had been repacked on February 11, 1988, and was left in a nearly

shut position following this work. The valve was open enough to allow the

main steam pressure to equalize across it before the turbine was placed in

service. The valve's new paa,ing shifted, evidenced by the fact that the

valve developed a packing leak about the time of the turbine trip, which

allowed the pressure downstream of the ,alve to be relieved. This re6fced

pressure was sensed by PT-5220, which then caused a turb'ne trip.

The NRC resident inspectors found no instructions in Maintenance

Procedure MP-2115 to return a valve to its as-found pcsttion following

maintenance. Although this is not safety-related equipment, the lack of a

step to return the equipment to service following maintenance is of some

concern. The licensee considers the potential probiers associated with

this significant and will revise all m.iintenance precedures fo valves to

j

_ .. _ _

. .

. . . . .

11

record the as-found position before commencing maintenance and to return

the valve to that position or leave it in another position with the shift

supervisor's knowledge and consent following completion of the maintenance

activity. The NRC resident inspectors will monitor the licensee's

implementation of these measures.

No violations or deviations were identified in the review of this program

area.

8. Radiological protection

The NRC resident inspectors observed the licensee's activities in this

area to verify their conformance with policies, procedures, and regulatory

requirements.

Health physics professionals were observed on all shifts, performing plant

tours, area surveys for radiation levels and radit. .ive contamination,

and checking the operability of area radiation man, toes and continuous air

samplers. The NRC resident inspectors verified tha the results of area

surveys were posted at entrances to radiation areas and in other

appropriate locations. Health physics supervisors and personnel were

aware of the plant status and activities which involved potential

radiological concerns.

The NRC resident inspectors observed that health physics personnel were

present and available to provide astistance whenever workers are required

to enter a radiologically controlled area.

No violations or deviations were identified in the review of this program

area.

9. Physical Security Observation

The NRC resident inspectors vcrified that there was a lead security

officer (LS0) on duty authorized by the facility security plan to direct

security activities onsite for eac's shif t. The LSO did not have duties

that vould interfere with the direction of security activi+1es.

The NRC resident inspectors verified, randomly and on the backshift, that

the minimum number of armed guarcs required by the facility's security

plan were present. Search equipment, including the X+ ray machine, metal

detector, and explosive detector, were operational or a 100 percuat

hands-on search was being utilized.

The protected area barrier was surveyed by the NRC resident inspectors.

The barrier was properly maintained and was not compromised by erosion,

openings in tl.. fence fabric or walls, or proximity of vehicles, crates or

other objects that could be used to scale the barrier. The NRC resident

inspectors observed tne vital area barriers were well maintained and not

ccmpromised by obvious breaches or weaknesses. Th NRC resident

~

- . .

...o

12

inspectors observed that persons granted access to the site were badged

indicating whether they had unescorted or escorted access authorization.

No violations or deviations were identified in the review of this program

area.

10. Exit Meeting

An exit meeting was conducted on March 8, 1988, attended by those

identified in paragraph 1. At this time, the NRC resident inspectors

reviewed the scope and findings of the inspection.

. -

. -. ..

ML20150F008

Text

Navigation menu

Search