ML20245C834

Insp Rept 50-267/89-03 on 890201-0318.Violations Noted. Major Areas Inspected:Lers,Reserve Shutdown Matl Removal, Operational Safety Verification,Radiological Controls, Monthly Surveillance Observation & Maint Observation

ML20245C834
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/13/1989
From:	Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20245C810	List: ML20245C808 ML20245C831 ML20245C834
References
50-267-89-03, 50-267-89-3, NUDOCS 8904270410
Download: ML20245C834 (17)

See also: IR 05000267/1989003

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' APPENDIX B

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LU.S. NUCLEhR REGULATORY COMMISSION

H REGION IV?

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NRC Inspection Report: 50-267/89-03- Operating License: DPR-34

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7 Docket: 50-267

'Licenseei Public Service Company of Colorado (PSC)

P.O. Box 840 v,.. ,

Denver, Colorado . 80210-0840:

.. Facility Name: Fort St. Vrain Nuclear Generating Station (FSV)

Inspection At: FSV, Platteville, Colorado

,. .. Inspection Conducted: LFebruary .1 through March 18, 1989

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, Inspectors: lR. E. Farrell, Senior Resident' Inspector (SRI)

P. W. Michaud,' Resident Inspector (RI)

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LApproved: *

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T. F. Westerman,. Chief, Project Section 8 Datfe

Division of Reactor Projects

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Inspection Summary

Inspection Conducted February 19through March 18, 1989 (Report 50-267/89-03)

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Areas Inspected: Routine,: unannounced inspection of onsite. followup of

licensee event reports-(LERs),_ reserve' shutdown material removal, operational-

safety verification, radiological controls,. monthly surveillance observation,

-monthly maintenance observation, and a coastdown and defueling' meeting.

Results: The licensee completed.the reserve shutdown material removal during

this inspection period. Several equipment malfunctions resulted in various

delays _during this effort, but support by all licensee organizations

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. contributed to the' completion of_this effort in as timely a manner as possible.

Numerous abnormal or first-time operations and configurations existed during

'this. inspection period. An appropriate-level of training, pre-job briefing,-

and monitoring contributed to-the. licensee's success in recovering from the

inadvertent actuation of the reserve' shutdown system and the high moisture

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levels in the primary coolant.

One' violation was identified by the NRC resident inspectors (paragraph 6),

which involved incorrectly posted radiological boundaries. The licensee's

health' physics organization performed in an effective manner in support of

-numerous evolutions which included radiological. concerns, with the exception of

this: violation.

8904270410 89J414

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PDR ADOCK 05000267

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. .s The licensee's preventive maintenance program.was reviewed and found to contain

some improvements, but is still not existing at a level which is desirable.

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DETAILS

1. Persons Contacted

D.' Alps, Supervisor, Security

• M. Block, System Engineering Manager

• L. Brey,' Manager, Nuclear Licensing and Resources

• M. Coppello, Central Planning and Scheduling Manager

R. Craun, Nuclear Site Engineering Manager

, *D. Evans, Operations Manager

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• M. Ferris, QA Operations Manager

• C. Fuller, Manager, Nuclear Production

• B. Gares, Executive Secretary to Vice President, Nuclear Operations

• J. Gramling, Supervisor, Nuclear Licensing Operations

M. Holmes, Nuclear Licensing Manager

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• R. Hooper,' Nuclear Technical Training Supervisor i

• J. Johns, Supervisor, Nuclear Licensing Engineering 'j

M. Niehoff, Nuclear Design Manager '

F. Novachek, Nuclear Support Manager

• H. O'Hagan,' Outage Manager q

• J. Reesy, Nuclear Support Engineering Manager i

• D. Rodgers, Nuclear Computer Services Manager

• R. Sargent, Assistant to Vice President, Nuclear Operations

• L.. Scott, QA Services Manager

• V. Snyder, Maintenance Dept. Manager

• P. Tomlinson, Manager, Quality Assurance

• D. Warembourg, Manager, Nuclear Engineering

R. Williams, Jr. , Senior Vice President, Nuclear Operations l

The N9C inspectors also contacted other licensee and contractor personnel

during the inspection.

• Denotes those attending the exit interview conducted March 21, 1989.

2. Plant Status

The reactor remained shut down throughout this inspection period. The

licensee's efforts were directed towards recovery from the unplanned  ;

actuation of the reserve shutdown material and the removal of moisture

from the reactor coolant system.

The reserve shutdown material removal effort was hampered with equipment

malfunctions, but was completed satisfactorily on March 2, 1989.

Moisture removal from the reactor coolant system was being accomplished by

an. evacuation of the prestressed concrete reactor vessel (PCRV) at the end

of the report period.

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3. Onsite Followup of Licen*ee Event Reports (LERs) (92700)

The NRC inspectors reviewed selected LERs to determine whether corrective

actions as stated in the LERs are appropriate to correct the cause of the

event and to verify these corrective actions have been implemented.

LER 87-15 reported a Loop 1 shutdown (ESF actuation) when the interlock

- sequence switch (ISS)lwas moved from " Low Power" to " Power." The loop

shutdown was caused by an improper wiring configuration which provided a

trip signal to both "A"' and "B" Helium Circulator's logic circuitry when

the ISS was placed in the " Power" position. The improper wiring was a

result of incorrect electrical drawings which showed certain cables to be

determinate and some used as spares that actually were not. These

incorrect drawings were used to ground spare ccaductors to reduce

electrical noise during corrective actions in response to LER 86-28-1.

The licensee's investigation of the discrepancy between electrical

drawings and the as-found configuration concluded the cables in question

should have been, but apparently were not, determinate during original

plant construction. Corrective actions included determinating the two

conductors, which corrected the discrepancy with the electrical drawings

and also removed the false trip signal from the'"A" and "B" Helium

Circulator's logic circuitry. A special test (T-360) was also performed

prior to startup to ensure the plant protective system (PPS) had no

abnormal trip signals. These actions are acceptable to close this LER.

LER 87-22 reported a. loop shutdown (ESF actuation) with the reactor shut

down. This occurred during a surveillance test due to the failure of an

electronic logic chip. The failed chip placed one logic channel in a trip

l condition. When a technician tripped another channel as part of the

surveillance test, the 2-of-3 logic was satisfied and caused a loop

shutdown. The failed logic chip was replaced and the surveillance test

was subsequently completed satisfactorily. The component failure resulted

in a conservative (trip) condition as designed, and thus did not present

an unanalyzed configuration. The licensee's corrective actions are

considered suffkient to close this LER.

LER 87-28 described a loss of offsite electrical power event during

performance of a postmaintenance test with the reactor shut down. The

postmaintenance test was to be performed on the reserve auxiliary

transformer (RAT) firewater deluge control relay. Part of the test

procedure included lifting leads so the RAT would not actually trip. The

procedure incorrectly epecified the leads to be lifted, and when the RAT

deluge system was actuated per the test procedure, the RAT breakers

tripped. The root cause of'this event was identified as personnel error

by the licensee, in that the test preparer specified incorrect leads to

disable the RAT trip function as intended. Additionally, administrative  !

procedures did not require an independent review of postmaintenance test '

procedures. The licensee corrected the deficient test procedure and

reperformed it satisfactorily. The engineer involved was counseled on the

accuracy required of test procedures and the need to ensure erroneous

results of this sort are precluded. The licensee's Administrative

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Procedure SMAP-23, " Post Maintenance Testing,"; Issue 6, includes  !

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Step 3.3.4, which requires an independent review for accuracy. These

actions provide a sufficient basis to close this LER.

LER 88-04 reported a manual reactor scram from 74 percent power due.to an

upset on the offsite electrical power grid. This grid disturbance caused ,

L several western power plants;ta shut down. The reactor operator manuallyJ

scrammed the reactor just "; the turbine automatically tripped due to load

swings. The plant subse b .1tly experienced an unplanned radioactive gas

release from the core support floor vent system, and a Notification of

Unusual Event was declared. The velease occurred when a safety valve in

the core support floor vent system lifted due to flow restrictions in the'

downstream' piping. The total activity released was approximately 15

percent of technical specification (TS) limits and resulted in a total

dose of 3.67 E-5 rem at 'the exclu: ion area boundary. The licensee

performed a special test (T-383), on the core support floor vent system

after cleaning strainers and replacing downstream filte,s. The results of

this test indicated the relief valve setpoint (5 psig) sas too low, and

that it should be changed to 10 psig. This setpoint '.nange was reviewed

by the NRC inspectors and verified to be acceptable, both for protecting

the-system from overpressurization and to provide a margin against

unplanned releases. -The new setpoint has been verified to be successful

in preventing unplanned releases from this path during subsequent reactor

shutdowns. Additionally, wiring deficiencies found in the turbine control

circuitry following this event have been corrected. This LER is closed.

LER 88 2 06 reported a manual scram actuation from 71 percent power when all

circulating water flow was lost. The loss of circulating water was due to.

an expansion joint failure on the "1A" Circulating Water Pump which

flooded the circulating water pump pit. The expansion joint failed due to

its age of approximately 15 years. The service life of these joints is

approximately 10 years. The root cause of this event was the lack of a

preventive maintenance program for rubber expansion joints. As a result

of 'this event, all eight expansion joints in the circulating water pump

pit were replaced. In addition, the licensee inspected and evaluated the

condition of all expansion joints in the plant. A number of expansion

joints have been replaced as a result of these inspections, including

those on tne condenser water boxes, diesel generator heat exchangers, and

condensate pumps. The licensee has developed and implemented a preventive

maintenance program for the inspection and periodic replacement of

expansion joints throughout the plant. These actions are sufficient to

close this LER.

4. Reserve Shutdown (RSD) Material Removal (6071(

The seven-region group of the RSD system was inadvertently inserted into

the reactor on January 19, 1989. The NRC inspectors closely monitored the

licensee's preparation and implementation of activities to remove the RSD

material'from the core.

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!Thehotservicefacility(HSF)wouldnormallyhavebeenutilizedtosetup

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,' the reserve shutdown vacuum tool. However, the HSF was configured for

control rod drive' refurbishment. Since it was' desirable to leave the HSF '

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.in.this configuration ,and with the fact that it would have taken 2 weeks

v .to' reconfigure the HSF, the licensee' decided to utilize the new fuel .

, loading' port (NFLP) to set up the RSD vacuum tool. A new Procedure,.

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! !MPF-1074, Issue 1, "In-Core Removal of RSD Material /NFLP," was written to

perform this evolution. . The'NRC inspectors reviewed this procedure and

found that it contained. adequate precautions and instructions.to ensure

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the, reactor was maintaird 'n refueling conditions and to specify

responsibilities of personnel involved in its use. Cetailed steps were-

1. ' provided:to' move,sset.up, and tu t equipment. The licensee provided

, in-depth training on this procedure, as well as other procedures, and the

overall-scheme of the RSD removal operations. The NRC inspectors attended

two of these training'sessiont and found them to be well organized and

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informative.

The' vacuum tool consists of a 400 Hz motor attached to a vacuum impeller,

a hopper to collect the RSD material, and a suction probe to be inserted

into the RSD channel; all of which are housed inside the auxiliary

transfer cask'-(ATC). The ATC provides shielding for personnel and a

radiological boundary to handle control rod ~ drive and orifice assemblies

and the~ reserve shutdown vacuum tool. An external motor generator

provides' the 400 Hz power to the vacuum motor, and is connected via wiring

inside the ATC. . Because of the configuration, the. Vacuum motor has

minimal heat removal characteristics, and as such is limited to a 4 minute

duty' cycle which must be followed by a minimum 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> cooldown period.

The general sequence of activities to remove the RSD material for each of

the'seven affected regions was:

a. Perform a shutdown margin verification in accordance with Procedure

SR 4.1.6.c/d-x, '" Shutdown Margin Evaluation. for In-Core Maintenance."

b.- Retract the control rods and insert a rewind tool to hold them after

power is removed.

c. Install a reacter isolation valve (RIV).

d. Remove the control rod drive and orifice assembly (CRD0A) using the

ATC.

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e. Transf er and store the CRD0A in an equipment storage well.

f. Assemble the vacuum. tool, test, and place inside the ATC.

g. Move the ATC to the RIV, open the RIV and vacuum the RSD material  ;

' from the region. This operation consists of lowering the vacuun

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probe until it contacts the_ RSD material (as indicated by a loss of

weight"from the ATC grapple), then energizing the vacuum motor and

slowly. lowering the probe into the RSD channel. When the end of the

probe-is approximately 6' inches from the bottom of the RSD channel, a

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mechanical stop prevents' further travel, and the probe is withdrawn

into the ATC.

h. . . Move the ATC to the NFLP, and extend and disconnect the probe and

vacuum tool.

i. Empty the vacuum tool hopper and weigh the RSD material to verify all

the material was removed.

j.- Install a refurbished CRDOA.into the region using the ATC.

The NRC~ inspectors attended shift briefings and observed good

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informational exchanges and overall coordination. Key personnel were

interviewed to ensure their understanding of their responsibilities, and

knowledge of procedures, administrative requirements, and actions to be

taken under unexpected or abnormal conditions. Management involvement was

observed throughout thisieffort. Overall, the RSD removal effort went

well, despite problems which occurred at various stages.

Following ex-core testing, the first attempt to vacuum RSD material on

February 5, 1989, was unsuccessful. Procedure MPF-1074 specified the

starting. current for the vacuum motor should decrease to less than 40 amps

within approximately 40 seconds. When it did not do so after 45 seconds,

the motor was deenergized. Investigation of the problem found wiring

damaged on the pigtail between the ATC grapple head and the vacuum motor.

In addition, external wiring on the ATC was.found to be damaged. This

wiring was repaired on February 9. ,As part of the troubleshooting to

determine the cause of the wiring damage, the 400 Hz motor generator was

tested separately and found to be defective. A new 400 Hz generator was

obtained, installed, and tested satisfactorily on February 13.

On February 14, the vacuuming effort was partially successful, but the l

probe stopped approximately 2 feet before it should have.' An

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investigation discovered the vacuum tool hopper / canister had a restriction

which prevented movement of the probe for the last 2 feet. This

restriction was relieved and the RSD material from Regions 5 and 22 l

was removed on February 15. Regions 3 and 34 were successfully vacuumed j

on February 17.

On February 19, while vacuuming RSD material from Region 28, the 400 Hz

-vacuum motor failed. The motor was disassembled and sent out for repair

on February 22. Alternate means of removing RSD material had been under 1

consideration from the outset. An auger was developed which was fairly

successful in shop tests, but which tended to grind the RSD balls together,

creating dust and other concerns. An external blower scheme was also

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developed which was similar to the existing vacuum tool, but utilizing an

external blower rather than a motor lowered into the reactor. This

external blower appeared to be a viable option and was reviewed by the NRC

. inspectors.

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l Controlled Work Procedure (CWP)89-050, "In-Core Removal of Reserve

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Shutdown Material with Roots Blower" was developed by the licensee. The I

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NRClinspectorsreviewedCWP89-050andtheassociated'designinformation

and safety evaluation. Concerns with this proposed method were discussed

with the licensee and resolved. However, the 400 Hz vacuum motor was

rebuilt and returned to the site before this alternate meth3d was

implemented. The existing vacuum tool was ceassembied and all.RSD

material was removed from the core on March 2.

A visual examination was made on the RSD material removed from each of the

seven regions. A sample from one high and one low boron

concentration material was sent offsite to be chemically analyzed for

leachable boron content, in accordance with TS 4.1.9 D.4. A visual

inspection of one RSD hopper was also performed.

Verification that'all RSD material had been removed from each of the seven

regions was accomplished by weighing the removed material and comparing it

to records of the weight of material which had been installed. A

tolerance of plus or minus 1 pound of material was developed and

documented in licensee Memo PPS-89-0433, dated February 3, 1989. The 1

pound tolerance was based on consideration of:

a. The accuracy and calibration of the scale used to weigh the material

b. Uncertainty over the accuracy of the original weights from records

c. Verification that sufficient material was inserted to fill the RSD

channel above the height of the active core

d. The amount of material which could be left in the core with no

significant effects on reactivity

e. Any effects on axial or radial power distribution that any material

left in the core would have

f. Assurance that a subsequent discharge of RSD material into the same

region would not overfill the RSD channel

The NRC inspectors reviewed the licensee's calculations and evaluation of

the 1 pound tolerance and found them acceptable.

The RSD material removal from six of the seven regions met this acceptance

criteria. However, the. material removed from Region 25 weighed 5 pounds

l more than that recorded as being loaded in 1985. Nonconformance Report

(NCR)89-050 was written to' address this discrepancy. The licensee's

evaluation and disposition of this NCR concluded that an error must have

been made during the loading in 1985 since there is no possible pathway

for RSD material to migrate from one region to another. The NRC I

-inspectors reviewed the licensee's actions in response to this NCR and l

found them acceptable. l

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

area.

t 5. Operational Safety Verification (71707)

The NRC. inspectors made daily tours of the control room during normal

working hours and at least once per week during backshift hours. Control

room staffing was verified to be at the proper level for the plant

conditions at all times. Control room operators were observed to be

) attentive and aware of plant status and reasons why annunciators were lit.

The.NRC inspectors observed the operators using and adhering to approved

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procedures in the performance of their duties. A sampling of these

procedures by the NRC inspectors verified current revisions and legible

copies. During control room tours, the NRC inspectors verified that the

required number of nuclear instrumentation and plant protective system

channels.were operable. The operability of emergency AC and DC electrical

power, meteorological, and fire protection systems was also verified by

the NRC inspectors. The reactor operators and shift supervisor logs were

reviewed daily, along with the TS compliance log, clearance. log,

operations deviation report (0DR) log, temporary configuration report

(TCR) log, end operations order book. Shift turnovers were observed at

least once per week by the NRC inspectors. Information flow was

consistently good, with the shift supervisors soliciting comments or

concerns from the reactor operators, equipment operators, auxiliary

tenders,.and health physics' technicians. The licensee's station manager,

operations manager, and superintendent of operations were observed to make j

routine tours of the control room.

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The NRC inspectors made tours of all accessible areas of the plant to j

assess the overall conditions and verify the adequacy of plant equipment,

radiciogical controls, and security. During these tours, particular 1

atter. tion was paid to the licensee's fire protection program, including j

fire extinguishers, fire fighting equipment, fire barriers, control of I

flammable materials, and other fire hazards. l

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A walkdown of the purge vacuum system, reactor building ventilation

system, reactor building area radiation monitoring system, control room

ventilation system, 480 VAC essential power distribution system, and

portions of the firewater system was performed by the NRC inspectors. I

These systems were selected because of their relation to work performed j

during various portions of the reserve shutdown material retrieval and '

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moisture removal efforts. Valve and breaker positions were verified,

where possible. When affected by a clearance, the valves or breakers were

, verified to be positioned in accordance with the clearance requirements.

Power supplies for components in these systems were verified, but were

also subject to clearances in some cases. During these system walkdowns,

the NRC inspectors verified the operability of standby or backup equipment

when components or portions of systems were inoperable due to clearances.

The NRC' inspectors reviewed seseral TCRs which were used to install

equipment in support of the outage recovery efforts. Proper reviews and

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approvals were verified for each TCR. Three of then TCRs were

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independently verified by the NRC inspectors:

TCR 89-01-01 installed a manometer on the refueling f;oor to read

reactor pressure under refueling conditions (subatmospheric),

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TCR 89-01-04 supplied power to the reserve shutdown vacuum motor l

generator from a spare breaker on Reactor Motor Control Center 5. l

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TCR 89-03-02 deenergized. rod position. indications so that oxidation

j of electrical connections for rods which were removed from the core

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would be minimized.

No discrepancies were noted during these walkdowns.

The licensee estimated that approximately 250 gallons of water' entered the

reactor coolant system through a leaking core support floor cooling water

tube. This tube was known to have a leak and was inadvertently left

unisolated during the reserve shutdown removal effort. The licensee

determined it would take approximately 2 months to remove this amount of

water through the purification system. In an effort to shorten the time

to remove the water, and the associated dates for criticality and

unrestricted operation, the licensee devised a setup to evacuate the PCRV.

This evacuation to as-close-to-full-vacuum-as-achievable, was calculated

to remove the water in approximately 1 to 2 weeks rather than 2 months

using the normal purification system lineup.

The system for evacuating the PCRV was designed and approved under Change

Notice (CN) 2916. The NRC inspectors reviewed this CN and the associated

safety evaluation in detail. The PCRV was designed for a vacuum pressure

of -12 psig as documented in FSAR Appendix E, Design Criteria DC-11-1.

Evacuation of the PCRV was previously accomplished ~in 1976 and 1982.

CN-2916 included temporary.and permanent modifications to evacuate the

PCRV through a refueling penetration, a cold trap to remove moisture, and

the installed purge vacuum pumps The discharge of the purge vacuum pumps

'was directed to the reactor building ventilation system which flows

through the reactor plant exhaust filters before exhausting through the

plant stack. An additional flowpath from another refueling penetration

to the reactor building ventilation system was used to initially draw a

vacuum on the PCRV, via a commercial blower, at a faster rate than possible

with the purge vacuum pumps. In addition, a separate cold trap and vacuum

pump was used to evacuate the core support floor (CSF) in order to minimize

the differential pressure between the CSF internals and the PCRV.

The NRC inspectors examined the licensee's design and analysis of the PCRV

evacuation process, paying particular attention to radiological concerns,

structural integrity concerns, instrumentation and monitoring issues, and

the potential effects of moisture saturated coolant on core components. .;

The radiological concerns are addressed in paragraph 6 of this report. 1

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The ~ structural effects of the' evacuation on the core support floor

p , concrete,. steel, vent system, and cooling water system were analyzed and

" documented. The effects on the'PCRV steel liner, primary closure, helium

qcirculator shutdown seals, PCRV rupture discs, and region isolation valve '

seals were also documented in the design analysis.

. .The design analysis for CN-2916 also considered'the effects of moisture on--

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control' rod drive-(CRD) and RSD system operability. With the PCRV at a

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, vacuum, purge flow-to the.CRDs had to be secured. The analysis considered

lthe' effects of: saturated vapor conditions on.these components for the 7

L days during which the PCRV was evacuated.

c The:11censee reviewed plant drawings and operating procedures, then walked

down systems which could have possibly been damaged.from evacuation of the,

PCRV to establish boundaries to isolate all affected instrumentation. All

piping subject to a vacuum was double isolated, and all critical

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instrumentation had process lines removed or vented in case of failure of

.the double valve isolation. The isolation' valves were positioned and

restored on an eight part clearance, which provided a means to return

essential equipment'to_ service more easily should,an emergency have

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occurred. The NRC~ inspectors reviewed l portions of these clearances and

found no' discrepancies.

The NRC inspectors also reviewed CWP 89-38, which provided instructions

for.the installation and removal of equipment to perform the PCRV

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evacuation. Temporary tie-ins for process flows, cooling water, and -

drains ^were described for the installed. purge. vacuum pumps, the core

support floor vacuum pump, and the commercial blower. The CWP was found

to contain sufficient steps,-adequate detail, and provisions for QC

' involvement, where required;

On March 12, 1989, the licensee entered a loss of forced circulation, with

1. a calculated allowable time of,193.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before forced circulation of

primary coolant.was- required. This provided a sufficient window in which

to perform the PCRV evacuation. The NRC. inspectors witnessed the initial

phase'of the evacuation, and subsequently' monitored its status. The PCRV

, evacuation proceeded smoothly, with approximately 180. gallons of water.

removed from the PCRV at the end of this inspection period. The NRC

s .< - ' inspectors will monitor the licensee's activities in completing the

evacuation'and restoring from it.

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The NRC inspectors. randomly verified that the number of armed security

officers required by the security plan were present. A lead security

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officer was on duty to direct security activities on each shift. The NRC

inspectors verifled that search equipment, including an x-ray machine,

. explosive detector, and metal detector,'was operational or a 100 percent

hands-on search was conducted.

The protected area barrier was surveyed by the NRC inspectors to ensure it

.was not compromised by erosion or other objects. The NRC inspectors

, t observed that vital area barriers were well maintained and not '

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compromised. The NRC inspectors also observed that persons granted access

to the site were badged and visitors were properly escorted.

6. Radiological Controls (71707)

The NRC resident inspectors observed health physics technicians performing

surveys and checking air samplers and area radiation monitors.

Contamination levels and exposure rates were posted at entrances to

radiologically controlled areas and in other appropriate areas and were

verified to be up-to-date by the NRC inspectors. Health physics

technicians were present to provide assistance when workers were required

to enter radiologically controlled areas. The NRC inspectors observed

workers following the instructions on radiation work permits concerning

protective clothing and dosimetry, and observed workers using proper

procedures for contamination control including proper removal of

protective clothing and whole body frisking upon exiting a radiologically

controlled area.

During this report period, the licensee was involved in a number of

evolutions with radiological concerns. This provided the NRC inspectnrs

with an opportunity to observe the licensee's health physics department

dealing with moderately high contamination and radiation levels, which is

not a usual occurrence at Fort St. Vrain.

Work performed in support of the reserve shutdown material removal and the

evacuation of the PCRV involved direct communication with reactor

internals, with associated high radiation levels and handling of

contaminated equipment. The NRC inspectors observed thorough involvement

and preparations by the health physics department in all phases of these

evolutions. All personnel' involved were observed to adhere to

requirements of radiation work permits (RWPs). ' The NRC inspectors

reviewed several RWPs to assess their adequacy in relation to the area

involved and the work to be performed. These included:

RWP 10571 - Reserve Shutdown Work in the New Fuel Loading Port

.RWP 10576 - Maintenance in the Auxiliary Transfer Cask

RWP 10577 - Control Rod Drive Work in the Hot Service Facility

RWP 10578 - Decontaminate / Modify Refueling Sleeve

The overall radiological controls in support of evolutions during this

report period were good, though one instance of noncompliance was found by

the NRC resident inspectors. During the reserve shutdown removal effort,

the refueling floor (Level 11) was made a radiologically controlled area

(RCA), in order to control access. The entry point was established on

Level 10 at the south stairway to Level 11. Because of relatively high

background radiation levels and the licensee's desire to monitor for  ;

potential contamination at the refueling floor, friskers were located at

the top of the stairs on Level 11. Personnel leaving the refueling floor

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frisked clean on Level'11, then walked down to Level 10', where they signed

out and exited the RCA. This arrangement was acceptable'but for the fact

that the NFLP access was located in the area between the access point on

Level 10 and the frisking station on Level 11. The NFLP was utilized for

emptying the reserva shutdown vacuum tool after removal of material from

the reactor. As such, the NFLP was a high contamination / airborne activity

area and was controlled under a separate RWP. Due to the high background

radiation levels at the access.to the NTLP, a'frisker could not be ;ocated

there. Thus, personnel exiting the NFLP. removed their protective clothing

at a step-off pad located at the NFLP access, then climbed the stairs to

Level 11 to use the friskers located there. The result was that

potentially contaminated individuals were traversing the same area which i

personnel who had frisked themselves clean were'using to exit the RCA.

This was brought to the licensee's attention and was corrected by locating i

the RCA access point at the north stairway between Levels 10 and 11. The

licensee was informed that the failure to' adequately establish and control

access to radiological contrcl areas and contaminated areas is an apparent

violation of NRC regulations (267/8903-01). >

7. Monthly Surveillance Observation (61726J

The NRC resident inspectors observed portions of surveillance testing in

support of the reserve shutdown material removal effort. The selected

surveillance procedures were reviewed for conformance with TS ,

requirements, in terms of LCOs and acceptability of results.  ;

Administrative approvals and clearances when required were verified by the

NRC inspectors prior to test. initiation. Test equipment was erified to

be within its calibration cycle. Testing was performed by qualified

personnel in all cases. Portions of the following surveillance procedures

were observed by the NRC inspectors:

SR-MA-11-RX, Issue 1, " Reserve Shutdown Material Sample and Hopper

Inspection"

SR-RE-48-X, Issue 6, " Refueling - CRD Penetration Leakage Test"

SR-4.1.3.C-X, Issue 2, " Control Rod Drive and Orifice Operability"

SR-4.1.6.C/D-X, Issue 2, " Shutdown Margin Evaluation for In-Core

Maintenance"

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SR-4.1.9.D.3-RX, Issue 1, " Refueling Penetrations Piping Examination"

SR-4.1.9.D.4-RX, Issue 2, " Reserve Shutdown Hopper Functional Test"

SR-5.2.15-A, Issue 16, "PCRV Penetration Interspace Pressure

Calibration"

SR-5.2.28-62-R, Issue 6, "Holddown Plate Bolting Examination"

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No violations or devii ..os-were identified in the review of this j

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8. Monthly Maintenance Observation (62703) 1

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The NRC inspectors observed portions of numerous safety related

maintenance activities during this inspection period. Most of these

activities were related to the RSD material removal effort. During

observation of maintenance activities, the NRC inspectors verified

the licensee was at all times in compliance with TS LCOs and that

redundant components were operable as required. Activities were

accomplished by qualified personnel. utilizing approved procedures.

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The activities observed and reviewert by the NRC inspectors included:

Operation of the reactor building crane in' accordance with Procedure

M0P-1001, Issue 5. <This crane was used' extensively in support of the

RSD removal effort. The procedure provides guidelines for

reoperation inspections in accordance wii.n ANSI B30.2 and NUREG 0612

as well as for operations involving heavy loads on the refueling

floor.

Operation of the auxiliary transfer cask'(ATC) in at cordance with

Procedure M0P-1006, Issue 1. The ATC provides shieldirg for

personnel and a radiological boundary to handle controi rod drive and

orifice assemblies and the reserve shutdown' vacuum tool. The

procedure provides precautions and instructions for handling of

various components, including expected weights of each.

Use and inspection of rigging' equipment in accordance with Procedure

M0P-1007, Issue 4.

Removal and replacement of equipment storage well covers in

accordance with Procedure MPF-1100, Issue 1.

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Removal and installation of control rod drive and orificing

assemblies (CRD0As) in accordance with Procedure MPF-1056, Issue 1.

This procedure provides instructions for moving a CRD0A with the ATC

between the PCRV, equipment storage wells, ar.d the hot service

facility.

Installation and renoval of shielding adapters in accordance with

Procedure MPF-1065. Issue 1. Shielding adapters provide personnel

shielding and a platform to mount the ATC on an equipment storage

well, hot service facility port, or the new fuel loading port.

Removal and insta)lation of reactor isolation valves (RIVs) in

accordance with Procedure MPF-1067, Issue 1, cnd operation of RIVs in

accordance with Procedure M0P-1009, Issue 1. RIVs provide personnel

shielding and a platform for mounting the ATC, fuel handling machine,

and primary seal cleaning equipment on the PCRV. These procedures

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provide prerequisites, precautions, and instructions for handling,

installation, removal, and operation of RIVs.

Removal and installation of PCRV top head holddown plates in

accordance with Procedure MPF-1091, Issue 2.

Removal and installation of' reactor penetration covers (secondary i

closures) in accordance with Procedure MPF-1094, Issue 1.

Removal and installation of helium purification train covers in

accordance with Procedure MPF-1095, Issue 1.

These activities were monitored by the NRC inspectors on a random basis

during this report period. Quality control and health physics involvement

in all phases of'these activities were observed by the NRC inspectors. No

discrepancies were noted during observations of the above activities.

During routine annual preventive maintenance on the "B" Diesel Generator,

the licensee discovered three cylinders on one engine (K-9206-X) with less

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than~ nominal, th'ough acceptable, characteristics.

On'Febfuary 27, 1989, the licensee began maintenance on this engine under

Station Service Request (SSR) 89500737. During disassembly, a metal

piec'e,.later determined to be a coolant flow director, was found in the

-engine coolant outlet plenum. An examination was made of the cylinder

. heads which had been removed for maintenance, and two coolant flow

directors were found to be missing. These coolant flow directors channel

flow to the center of a cylinder head to provide even heat removal

characteristics. Two nonconformance reports (NCRs) were written as a

result of these findings: NCRs 89-44 and 89-45. Because the type of

heads involved could not be distinguished from other installed heads, the

licensee decided to replace all 24 heads on all 4 diesel generator

engines. New heads were obtained for the "B" Diesel Generator's engines

and were installed and tested satisfactorily on March 13, 1989.

Replacement of the heads on tae "A" Diesel Generator's engines was in

process at the end of this report period and will be covered in a future

inspection report.

On March 15, 1989, the NRC inspector met with the licensee's maintenance

manager to review the licensee's preventive maintenant program. The

preventive maintenance program is described in Proce 9 e MAP-1, "FSV

Preventive Maintenance Program Description." The devc 3pment of specific

preventiva maintenance requirements is described in Procedure SMAP-27,

" Preventive and Corrective Maintenance Equipment Review." The NRC j

inspector reviewed the focus of these procedures with the licensee's  :

maintenance manager. The development of maintenance procedures for larger

activities and standardized station service requests with controlled work

procedures for smaller activities was discussed in some detail.

Weaknesses in the maintenance procedures were attributed in many cases to

the inconsistencies inherent in having different individuals developing

varicus procedures. The preventive maintenance program has resulted in an

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overall: increase in the quality of maintenance procedures and consequently

the work product. Improvements in the. licensee's ability to complete J

, preventive maintenance as scheduled suggest the program is having some

success in shifting maintenance activities from a corrective focus to s

preventive program.

There are still some indications, however, that-the licensee's maintenance

program is not yet where it should be. One example of this was the

failure of Pressure Control Valve PCV-4256, which supplies a backup source

of cooling water from the firewater system to the emergency diese'

generators' engine and room coolers, as well as other components. i'is

valve had been leaking for over 2 years, causing the dow.1 stream relief

valve (V-4599) to weep whenever the firewater header was pressurized. On

March 16, 1989, PCV-4256' finally failed during performance of

SR-5.2.10.A.1-M, monthly firewater pump and instrumentation functional

test. The failure of PCV-4256 caused relief valve V-4599 to cycle

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' rapidly, which caused repeated water hammer vibrations in the turbi.e

building firewater header. .These' vibrations resulted in the failure of

the turbine lube oil reservoir room firevater deluge valve (HV-4507). The

vibrations caused the disc of HV-4507 to fail, spraying firewater into the

turbine lube oil reservoir room to a depth of approximately 6 inches

before the header was isolated. The licensee subsequently repaired Deluge

Valve HV-4507 and Pressure Control Valve PCV-4256.

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

area.

9. Coastdown and Defueling Meeting (94702)

A meeting was held on March 7,1989 to discuss final coastdown operations

and defueling of Fort St. Vrain. This meeting was attended by the

licensee and their consultants, NRR, and Region IV. Coastdown issues

which were discussed included consideration of TS and FSAR limitations and

criteria, reactivity effects, and power peaking issues. The licensee

concluded operation during coastdown will be within the existing FSAR

analyses and TS limitations, and will present no unreviewed safety

questions. The licensee is scheduled to submit a coastdown safety j

analysis report to the NRC by May. 31, 1989. '

Various defueling plans and sequences were discussed, with emphasis on

reactivity control and monitoring, accident and safety analysis, and

computer modeling. The licensee has decided to defuel by regions in the l

reactor after considering the alternatives. This will include replacement  !

of removed regions of fuel with boronated " dummy blocks" to maintain ]

structural integrity of the core during defueling. The defueling plan is l

scheduled to be submitted to the NRC by May 31, 1989. This plan will

include the defueling action plan, safety analysis report, and technical

specifications.

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10. ' Exit Meeting (30703)

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An exit meeting was conducted'on March 21, 1989, attended by those

. identified in paragraph 1. At this meeting, the NRC inspectors reviewed

2the scoperand findings of the inspection.

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