ML20148T580

Insp Rept 50-267/87-32 on 871116-1204.No Violations or Deviations Noted.Major Areas Inspected:Fire Recovery Activities in Response to 871002 Fire

ML20148T580
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/25/1988
From:	Ireland R, Andrea Johnson, Murphy M, Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20148T544	List: ML20148T580
References
50-267-87-82, TAC-66365, NUDOCS 8802030340
Download: ML20148T580 (20)

See also: IR 05000267/1987032

Text

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APPENDIX A

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-267/87-32 Operating License: DPR-34

Docket: 50-267

Licensee: Public Service Company of Colorado (PSC)

P.O. Box 840

Denver, Colorado 80201-0840

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

Inspection At: Platteville, Colorado

Inspection Conducted: November 16-20 and Nov6*ber 31 through December 4, 1987

Inspectors: 7- < [ t[ _ .. s-

T. F. Westerman, Chief,' Reactor Projects

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Date

fection B

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kh. Q/! tts+ usv

R. E. Ireland, Chief, Plant Systems Section

VH/VV

Date

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'M. E. Mdrphy/Reacyor Inspector DRS Date

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A. . Jo son, Reactor Inspector, DRS Date

Accompanying

i Personnel: K. L. Heitner, Project Manager, NRR

F. B. Litton, Materials Engineering Branch, NRR

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Approved: dt 1 wMN

T. F. Westerman, thief, Reactor Project

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Date

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Section C

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8002030340 000120

PDR ADOCK 05000267

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Inspection Sumary

Inspection Conducted November 16-20 and November 30 through December 4,1987

(Report 50-267/87-32)

Areas Inspected: Nonroutine, announced inspection of fire recovery activities

in response to the fire of October 2, 1987.

Results: Within the areas inspected, no violations or deviations were

identified.

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DETAILS

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1. Persons Contacted

Licensee

+*R. Williams, Jr. , Vice President, Nuclear Operations

C. Fuller, Station Manager

+*F. Novachek, Technical / Administrative Services Manager

+*P. Tomlinson, Manager, QA

• L. Scott, QA Services Manager

+*J. Gramling, Supervisor, Nuclear Licensing - Operations

• H. Block, Supervisor, Nuclear Betterment

+*H. Dender, Jr. , Nuclear Licensing Coordinator

• N Snyder, Supervisor, FSV Maintenance

• H. Cappello, Supervisor, Planning / Scheduling / Store

• H. O'Hagan, NPD Outage Manager

+0. Warembourg, Manager, Nuclear Engineering

+G. Schmalz, Fire Protection Engineer l

+F. Borst, Nuclear Training Manager

+M. Ferris, QA Operations Manager

+R. Craun, Manager, Nuclear Site Engineering

+M. Holmes, Manager, Nuclear Licensing

+T. McIntire, Nuclear Site Engineering

+J. Eggebroten, Supervisor, Technical Support Engineering

l J. Wambach, Mechanical Engineer

B. Ring, QA Engineer (Metallurgist)

M. Seed, Civil Engineer

8. Tarrant, Mechanical Engineer

NRC

+*R. Farrell, Senior Resident In.ipector

• J. Milhoan, Director, Division of Reactor Safety

+G. Pick, Reactor Inspector

+0enotes presence at November 20, 1987, exit aeeting

• Denotes presence at December 4, 1987, exit meeting

I 2.0 Recovery from the Turbine Building Fire

Following the hydraulic oil fire on October 2-3, 1987, the licensee met

with the staff on October 30, 1987, and provided a preliminary report on

the impact of the fire. Section 10 of the report provided a list of

action items to be completed prior to rise to power. On November 23,

1987, the licensee provided by letter a response to NRC questions

regarding the fire and an update of Section 10 of the October 30, 1987,

report.

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The inspection of the action items identified in Section 10 of the

October 30, 1987, report is as follows:

2.1 Restoration and Testing of Plant Systems,

The licensee identified all tagged items in the fire zone using teams

of QC/QA and engineering personnel. A station service request (SSR)

was used to document the inspection. An operabili ;y test was then

3 performed to determine basic component functionalt.y and performance

where appropriate. Evaluations with regards to the environmental

y qualification (EQ) of class 1E equipment important to safety were

also performed to determine equipment status; i.e., replace or

use-as-is. Repairs and/or replacements were controlled by use of an

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engineering change notice (CN) and a controlled work procedure (CWP)

where a configuration change occurred. Where there was general

maintenance or replacement in kind, the work was controlled by an

SSR.

! Work completed under a CN and CWP was tested by a cold checkout

test (CCT) and a functional test (FT). Post-maintnenance

i testing (PMT) was performed for work accomplished using an SSR. In

some cases, a Technical Specification (TS) surveillance test (SR) or

a special test called a T-test was performed in lieu of an FT or PHT.

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As a final check to determine compliance with TS-required

4 surveillance requirements, i.e, to demonstrate operability, the .

licensee initiated a cross-reference matrix by component to the

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testing performed and the TS-required surveillance testing. This

! review was still in process at the end of this inspection and will be

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verified by the resident inspectors for completion prior to restart,

pee NRC Inspection Report 50-267/87-34.)

In addition, the licensee will perform an integrated loop shutdown

test (T-test) for both loops prior to restart. This also will be

3 witnessed by the resident inspectors. (See NRC Inspection

Report 50-267/87-34.)

! The NRC inspector concluded that the licensee has established an

i adequate program for restoration and testing of components included

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in the fire activities.

No significant deficiencies have been idetitified by the testing

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program, but problems such as open fuses, one case of crossed leads,

and inadequate termination of power cable have been identified.

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No violations or deviations were identified.

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2.2 Electrical Systems Restoration .

2.2.1 Summary of Fire Damage

TheOctober3,1987,firecausedconsiderabiedamagetoelectrical

components and cables in the immediate vicinity of the fire. In i

order to assess that damage and to determine what repairs or  ;

replacements would be needed, the licensee conducted a number of ,

detailed walkdowns. In most cases, the need for repair or  !

replacement was obvious. However, in order to assist in the  ;

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engineering evaluation that was necessary, particularly for l

electrical equipment (including cables) which must be EQ, the l

licensee prepared three-dimensional temperature profiles. These  ;

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profiles were based on the observed effects of the fire on various  ;

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materials for which the properties are well known. l

J In the immediate vicinity of the fire, a number of eled.rical i

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components had to be replaced outright or be repaired through

replacement of parts which were or might have been damaged by heat.

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In addition, approximately 100 runs of conduit (predominantly located i

4 above the fire) had to be replaced, along with some junction l

! boxes (JB) and associated cable terminations in existing or new JBs

at specific equipment items. Cables which were destroyed in two

I cable trays below the immediate location of the fire also needed

j replacement.

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2.2.2 Bases for Replacement and Repa h Work

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l The fire damage included cables and components which are classified l

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as nonsafety, safety-related, and EQ. Repair and replacement for [

both nonsafety and safety-related cables and components entailed f

i "like-kind" restoration and return to criginal prefire condition.  !

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Replacement and repair of EQ equipment required a comprehensive  !

evaluation of the aging effects of the fire on cables and components l

l in those temperature zones which were at 300'F or slightly higher, t

and in a 10-foot buffer u ne outside the 300*F profile, for which  !

specific temperatures could not be established based on jhysical j

evidence. The licensee's evaluation for EQ equipment is contained in

EE EQ-0065, Revision A and for EQ cables in EE-EQ-0066, Revision A.

Electrical replacement and repair work of all safety classes was

governed by 55Rs wherever modifications were not required. If i

modifications were required to complete the replacement and repair I

work, CNs and corresponding CWPs were utilized. [

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2.2.3 EQ Equipment

As stated above, EE-EQ-0065 cnntains a compre ensive evaluation of the ;

effects of the fire on EQ electrical components. This evaluation j

contains a component 4by-component assessment.af the impacts of fire

zone temperatures on the qualified life of components and the l

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effects, if any, of water used to extinguish the fire on those

components. Based on this evaluation, the licensee elected to

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replace or repair components or to evaluate the remaining qualified -

life of components which were not directly damaged by temperature ,

<. exposure but which may have suffered some aging of degradable  !

materials.

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In order to assure identification of all electrical equipment,  !

including EQ equipment in the fire zone, the licensee used the '

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existing plant components list (PCL) and the EQ master equipment

list (MEL) together with equipment location drawings. Through this

i means, a list of EQ equipment subject to various fire zone

temperatures was compiled.

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Based on review of EE-EQ-0065 and interviews of licensee personnel,

the NRC inspectors concluded that the licensee's evaluation was ,

, complete and that repair and replacement work was consistent with the

EQ prcgram requirements previously established by the licensee for

maintenance of EQ equipment. The NRC inspectors examined several -

components that had been replaced and concluded that the work had

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been done consistent with EE-EQ-0065 and the SSRs which governed the ,

work.  !

Based on interviews of licensee personnel and review of work

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documents, the NRC inspectors observed that all work was consistent

i with the DOR guidelines which apply to FSV. The only component

l upgrades to NUREG 0588 Category I requirements identified were the

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replacement of Cannon connectors by Namco connectors to a Moog

j servovalvt. and Collins transducer. l

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l 2.2.4 EQ Cables

, In regards to the replacement of EQ cables damaged by the fire, the $

licensee performed a comprehensive evaluation, similar to that done  :

for EQ components, contained in EE-EQ-0066. The licensee has taken i

the position, in accordance with their EQ program, that fire damaged  ;

1 cables can be replaced with "like-kind" DOR qualified cable in '

i agreement with the guidance contained in Regulatory Guide 1.89,

Section C.6, in particular, subparagraphs (b), "item to be replaced

is a part of an item of equipment qualified as an assembly;" and (c),

"used as a replacement on hand as a part of stock prior to

February 22, 1983."

The NRC inspectors reviewed PSC CN No. 2701 inv lying subcomponent EQ $

{ replacement cables in cable trays via JBs located in the fire zone.

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The NRC inspectors identified only a few EQ cables (Rockbestos

Firewall III, NUREG 0588, Category I) out of all the safety related

replacement cables in the cable trays. These EQ cables were l

terminated in JB 1444 (west end) and JB 1445 (east end) using Buchanan l

l Model NQ-511 and NQ-211 terminal blocks (TB), qualified to i

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NUREG 05B8, Category I. The terminal lugs (Amp Inc.) used for the TB

l terminations were not required to be EQ qualified as demonstrated by

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PSC engineering evaluation EE-EQ-54. EE-EQ-54 referenced another

utility's type test which demonstrated survival of LOCA c.).iditions

for FSV.

The NRC inspectors examined the manner in which the JB associated

with the cable trays wers mounted and determined that they were

adequately supported consistent with seismic requirements. Lug

attachments to the terminal blocks appeared consistent with

installation requirements. The NRC inspectors noted that, although

cables from the cable trays were not clamped as they entered the JB,

they aere either looped or tied-off to the cable trays to prevent

relative motion and tension on cables during maintenance work or

seismically induced cable tray motion.

Approximately 100 runs of conduit were replaced, principally at

elevations above the immediate fire location. Approximately 1/2 of

these runs contained EQ cables and the remainder contained both

safety and nonsafety-related cables. Review of selected SSRs and

CWPs, followed by observation of work in progress and/or completed,

confirmed that work had proceeded in accordance with defined

procedures.

Most of the replaced conduit utilized existing "trapeze" supports,

which were installed to the FSAR seismic criteria. New supports were

installed in connection with JB 1477 (CN-2711B) which was required to

reroute cables to hydraulic solenoid valves. Supports for this JB

are structurally adequate for the seismic loadings specified in the

FSAR. An internal PSC memorandum from H. A. Seed to P. F. Tomlinson,

dated November 25, 1987, confirmed that site engineering had

concluded that existing conduit supportt had not been damaged by the

fire and that they were acceptable for continued use.

2.2.5 Conclusions

Based on review of the licensee's technical evaluations, work

packages, and observations of work in progress and work completed,

the NRC inspectors concluded that all electrical components and

cables damaged during the October 3, 1987, fire have been

satisfactorilr replaced. Further, with respect to EQ equipment and

cables, all work needed to assure current qualified status and to

assure preservation of qualified status through preventive

maintenance and/or replacement of components has been completed.

2.3 Analyses of Structural Components from the Fire Area at

Fort St. Vrain Reactor

A tour of the fire area was conducted with licensee personnel. The

tour was followed by a review by the NRC inspector of the

itetallurgical analyses of the structural components taken from the

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fire area. The inctallurgical tests consisted of hardness

measurements, metallographic replication, NDE magnetic particle

inspection, and tensile testing.

Since the components were taken from the area in which the highest

temperatures were indicated, the analysis was intended to bracket

maximum structural damage in the fire area. The analysis did not

show any material deficiency attributable to the exposure. The

metallographic structure and hardness indicated that a polymorphic

change in material had not occurred. The components were acceptable

and properties were consistent with the originally installed

material. The magnetic particle inspection showed indications, but

these were shallow, capable of being buffed out, and expected to have

minor effect on structural integrity. Although geometric deformation

in certain I-beams occurred, which were replaced by the licensee, the

analysis showed no material deficiency attributable to rnetallurgical

structure change resulting from fire exposure.

The NRC inspector concluded that the material properties of the major

ferrous structural components were not degraded as a result of the

fire at the Fort St. Vrain reactor. However, the material properties

of nonferrous metal, such as aluminum and plastics (but not copper)

were degraded in the immediate vicinity. These materials were

replaced by the licensee to restore the original plant structural

integ ri ty.

2.4 HV-229,2 Hydraulics

The NRC inspectors verified by review of records and interview of

licensee personnel that repairs, including replacement of thermal

relief valves and orifices had been completed. This included both

electrical and mech 6nical work. The last work item was installation

of an electrical JB in accordance with CWP 292 which was verified by

the NRC inspector as complete on December 4,1987.

2.5 Hydraulic 011 Filter Canisters

The NRC inspector verified that the licensee had changed Controlled

Work Instruction SMAP-25, "System-22 Hand Valve Oil Filter

Replacement." to preclude the use of a pipe wrench on filter

housings. The procedure now requires use of a strap wrench to

preclude filter housing damage.

The NRC inspector also verified that bleed lines on pipe canisters

are not necessary as the capability exists to vent the system through

the installed 5-valve manifold. The NRC inspector also verified that

the FSV standard clearance points for system 91 has been updated to

reflect this capabiilty in issue 9 of that document.

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2.6 Cleanup of Hydraulic System

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The NRC inspector verified cleanup of the hydraulic oil. The NRC i

inspectors reviewed the chemistry test reports. The reports ,

indicated that the hydraulic fluid in the system had been restored to i

proper specifications.

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2. 7 Control Room Ventilation System

A review was made of certain aspects of the Control Room Ventilation  ;

System. These are discussed as follows:  !

The licensee is utilizing a control room panel instrument which '

had been calibrated to indicate tnat the control room is at a

positive pressure relative to the turbine building. However,  ;

, there is no formal scheoule for calibration of this instrument. '

The licensee plans to institute such a program, but the program i

is not in place now. This is an open item. (267/8732-01) t

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! The NRC inspector verified that the licensee had moved the [

ventilation pressure sensor from the auxiliary electric room to >

the control room in accordance with CN 2713. The NRC inspector [

j also verified completion of testing in accordance with FT 2713. L

The testing satisfactorily demonstrated that a positive pressure [

could be maintained in all modes of ventilation operation.

', These tests also included opening and closing of control room

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doors, j

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j 2.8 Cy t,rol Room and Reactor Building Filter Testing l

l The NRC inspector reviewed the following completed surveillance tests i

associated with the control room and reactor building

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SR 5.5.3a-5A Reactor Building Exhaust Filters and Charcoal j

Adsorber Samples -

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l SR 5.5.3be-A Reactor Building Exhaust Filters Charcoal Adsorber -

j Halogenated Hydrocarbon Removal ar.d HEPA Leak Test

SR-HE-7-A Control Room Makeup Filter-Charcoal Adsorber  !

l' Halogenated Hydrocarbon Removal and HEPA Filter leak }

Test  ;

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SR-HE-6-A Control Room Makeup Filter-Charcoal Adsorber Samples

The testing was perforned by a contractor (NCS Corporation). All

l' tests were found to be satisfactory. The tests included visual

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inspection, airflow distribution, charcoal adsorber halogenated i

j hydrocarbon test, and HEPA filter leak test. Test results indicated  !

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d greater than 99 percent efficiency for all tests.  !

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2.9 Fire Detection / Protection [

The licensee committed to the following actions prior to startup in [

the areas of fire detection and protection:

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Initiate management directive on fire alarms

Issue new procedure on fire protectiore operability [.

Determine compensatory actions associated with the fire [

detection systems j

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To satisfy these commitments the licensee developed FPOR-12 "Fire

Protection Operability Requirements - Fire Detectors" and promulgated  ;

it with the issue of Operations Order 87-14. FPOR-12 'pecifically '

addresses the compensatory actions to be taken with the loss of fire

area or zone detection and/or loss of the control room annunciator.

The NRC inspector reviewed Operation Order 87-14 and FPOR-12 and

attended a training session for licensed operators on the new fire

protection operability requirements.

2.10 Hydraulic 011 Storage Lockers

The NRC inspector observed that both hydraulic oil storage lockers

had been installed and that all oil drums had either been properly

stored or removed from the turbine building.

2.11 Review of SSRs for Missing Handwheels

The NRC inspector reviewed a computer run markup of SSRs that had

been performed by the licensee to determine the status of missing

System 91 handwheels. The NRC inspector also verified by walkdown of

System 91 that missing handwheels had been replaced.

2.12 Evaluation / Replacement of Plastic Valve Handles

The NRC inspector verified replacement of the plastic handwheels

associated with HV-2205 and HV-2206, which were damaged during the

fire, by review of RA-2012 and by inspect. ion of the valves to verify

replacement with metal handwheels.

2.13 Control Room Breathing Air Masks

The NRC inspector verified by inspection that five air masks are now

installed in the control room.

2.14 Reporting

The licensee submitted preliminary LER 87-23 on November 2,1987, to

describe the hydraulic oil fire.

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2.15 Hydraulic System Functional Test

In response to questions by the NRC inspector, the licensee performed

the quarterly surveillance test of the hydraulic system (SR 5.3.5-Q)

to demonstrate operability of the system. The NRC inspector reviewed

the test results. All acceptance criteria were satisfied.

No violations were identified.

3.0 Hydraulic System

3.1 System Operation

System operation utilizes both the normal and emergency headers in

each loop. One pump pressurizes the normal header and the other the

emergency header. Each header has a 6 gpm flow limiting valve going

to any one of the group headers. Thus, limiting flow to a group

header is 12 gpm. The operation is different from the system as

described in the FSAR.

The FSAR Section 9.11, Revision 5, states that under normal

conditions only one pump is running and that the second pump is in

standby. The licensee, however, operates the system with two pumps

continuously running. The licensee operation of the system is not

inconsistent with the Technical Specification LCO 4.3.7 which

requires only that two pumps be operable per loop. The redundance

provided by the second pump makes no difference whether the pump is

running or is in standby. The licensee has, however, agreed as a

result of the October 30, 1987, meeting with the staff in HQ, to

submit a change to the FSAR to eliminate any conflict. This is

considered an open item. (267/8732-02)

3.2 Indications and Alarmj

The system has a number of control room indications and alarms. Main

header and group header pressures can be monitored. Differential

pressure alarms can sense significant failures. The key to

operability is good surveillance practices. The licensee's quarterly

surveillance (functional tests) on November 24, 1987, indicate the

system is ready to operate again. Complete system recalibration is

scheduled for the Spring.

3.3 Hydraulic Leakage

The licensee is continuing to carefully monitor the hydraulic system

for excessive and uncontrolled leakage. Currently, three catch

basins in the turbine building and the power system sump in the

reactor building serve as collection points for leakage or flow from

thermal relief valves. The presence of these systems minimizes the

potential leakage and its associated fire hazard.

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3.4 Inspection of System on Level 1

A visual inspection of the hydraulic system "power plant" on Level 1  ;

of the reactor building was conducted.

The NRC inspector questioned why there were three wet pressure gauges

and one dry pressure gauge. The licensee subsequently confirmed that

the dry gauge was temporary and they were waiting for a replacement.

In response to NRC inspector questions regarding the spread of

pressure indication, the licensee verified by calibration that all

but the temporary gauge (dry) were within calibration specifications.

A nonconformance report (NCR) was issued for that gauge.

An NRC inspector's observation (not a ccmmitment or regulatory +

requirement) was that a number of plastic handwheels remain on this  !

portion of the system which represent a limitation in system  !

operation during or after a fire and, in the event of a high energy  ;

line break, it was also pointed out that the Group 1 header label is

missing on both loops and some deficiency tags remain in place.

4.0 NRR Review of Specific Issues Related to Fort St. Vrain Fire

A review of specific issues related to the Fort St. Vrain fire was

completed by NRR on December 7,1987, and a safety evaluation has been

issued (see Appendix B). NRR has concluded that the licensee's short term

corrective actions provide an acceptable basis for plant restart. The

licensee is, however, committed to complete their post-restart evaluation

within 60 to 90 days after restart.

5.0 Exit Interview

The NRC inspectors met with the NRC senior resident inspector and licensee

representatives identified in paragraph 1.0 on November 16-20 and

November 31 through December 4, 1987, and summarized the scope and

findings of the inspection as presented in this report.

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.' .. . APPENDIX B

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5= je cao

4 'o, UNITED STATES

l' v., 'i NUCLEAR REGULATORY COMMISSION _ _ . _ _ . _

' wassiNotoN. o. c. 20sss ,}

$ )  ;

k ,/ December 7, 1987 y'-

DECi4887

MEMORANDUM FOR: James L. Milhoan, Director .

h,

Division of Reactor Safety

Region IV

FROM: Dennis M. Crutchfield, Director

Division of Reactor Projects - I!!, IV,

V and Special Projects

Office of Nuclear Reactor Regulation

SUBJECT: ISSUES RELATED TO THE FORT ST. VRAIN

FIRE, DOCKET NO. 50-267 (TAC NO. 66365

TIA NO. IV-2-87)

References: 1. Memo dated October 21, 1987 from J. L. Milhoan,

Region IV to D. M. Crutchfield, NRR, concerning

Issues Related to Fort St. Vrain Fire.

2. Letter dated November 23, 1987 (P-87414) from PSC

to the NRC concerning Response to Second Request

for Additional Information Concerning Recovery

from Turbine Building Fire 4

In response to your request in Reference 1, we are providing our Safety Evalua-

, tion of the specific issues you have identified. These responses are provided

, in the Enclosure. Our conclusion is that the licensee's short term corrective

actions prcvide an acceptable basis for your approval of plant restart. However,

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it should be made clear to the licensee that his commitments to continue certain

post-restart evaluations must be honored. These evaluations must be submitted

to the staff within 60 to 90 days after restart.

Af ter the plant is restarted, it is our intention ti rapidly complete the

Appendix R related reviews. We appreciate your continued support in the review

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of Fort St. Vrain's safe shutdown systems for Appendix R.

We consider our responsibilities under TIA IV-2-87 coeplete at this time.

J.

Deiiffsl.p..

6Lm:.n

CrutcrifieV0irector

Division of Reactor Projects - !!!, IV,

V and Special Projects

Office of Nuclear Reactor Regulation

Enclosure:

As stated

cc: See next page

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CONTACT:

K. Heitner, NRR/PD-IV

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492-7592

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cc w/ enclosure:

M. Caruso

J. Callan, RIV

A. Beach, RIV

T. Westerman, RIV

R. Farrell, RIV

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1. p# C849'o, UNITED STATES

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i' y., '% NUCLEAR REGULATORY COMMISSION

r, wAsmNcToN. o. c. rosos

\, .... . )l Enclosure

SAFETY EVALVATION CONCERNING SPECIFIC ISSUES

RAISED BY OCTOBER 2-3, 1987

TURBINE BUILDING FIRE

FORT ST. VRAIN NUCLEAR GENERATING STATION

PUBLIC SERVICE COMPANY OF COLORADO

DOCKET NO. 50-267

1.0 INTRODUCTION AND BACKGROUND

On October 2 and 3,1987 a fire occurred in the Turbine Building at the

Fort. St. Vrain Nuclear Generating Station. On October 30, 1987, a meeting

was held with the licensee to discuss the fire and the licensee's plans

for recovery f roci the fire. By letter dated November 23,1987,(P-87414),

the licensee provided additional infomation regarding the recovery plan.

This safety Evaluation addresses specific issues corcerning that recovery.

2.0 EVALUATION

2.1 Hydraulic System

The staf f was requested to review the design of the Fort St. Vrain

hydraulic system with respect to:

1. fire protection considerations including seasures which may be

needed to prevent hydraulic oil from reaching hot surfaces

which could promote ignition; and

2. the adequacy of the hydraulic system design with respect to

acco.todating impulse pressures without failure (e.g., the

ruptured filter canister which appeared to contribute to the

fire).

The current design of the plant already contains a number of features

to protect against hydraulic oil fires. Additional fire detectors

are placed in areas where the hydraulic system contributes to a

higher risk of fire. In the subject fire, the licensee had deactivated

the audible alarms for this system and thereby compromised its

effectiveness. During the October 30, 1987 meeting, the licensee

comitted to stronger controls over the fire alare system and to

take appropriate compensatory measures when the alars system is

partially disabled.

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Additionally, the licensee's post fire evaluation concluded that the

hydraulic oil system f ailure occurred in a thersal relief valve.

This thermal relief valve had failed because it did not have a

protective orifice on its inlet line. The licensee has verified the

role of the orifice in limiting flow to the therinal relief valve to

an acceptable level of about 1 gpm. Additionally, physical evidence

showed that the failed thermal relief valve had received repeated

damage from system pressure transients. However, the orifice would

have prevented tnis problem. The licensee has now verified the

correct installation of tre orifices in the system. This will

correct the deficiency that led to this particular fire. Other

existing design features, such as spray shields and hydraulic oil

drains appear to be adequate and functioning correctly, to prevent

fires. The subject fire was well within our assumptions under

Appendix R for fires in the Turbine Building. hence we conclude that

the fire protection consiocratio% for the hydraalic sy, tem are

acceptable for plant restart.

Additionally, the licensee has comitted to corLct an evaluation of

exposed hot surfaces in proximity (10 foot radius) to the 30

hydraulic vahes. The licensee will evaluate surfaces with a

temperature of over 500*f, which are potential ignition sources.

This report will evaluate additional protective teasures and will be

s m itted by January 30, 1988.

The literset bn also perforted a detailed rett'1urgical analysis of

tw f ailed ',dra;lic cil filter cannister. Our original cancern was

that this fa'iled cennister was the source of hydraulic oil that

initiated the fire. The licensee perforeed a retallurgical annlysis

of the f ailed cannister and concluded that f ailgre occurrred at

temperatures above 650*F. Tests of another filter assembly showed

that failure in the absence of elevated temperabres was over 4 times

the hydraulic system's normal pressure. We find that the licensee's

analysis shews that the hydraulic system appe3rs adequately designed

to withstand actual system operating conditions. (This includes the

licensee's 3:tions to replace the tnermal relie' valve orifices

discussed at:ve.)

In addition, the licensee has comitted to perform quarterly sur-

veillance on the hydraulic system to verify component operability prior

to restart. The licensee has also comitted to testing of hydraulic

system design features such as flow limiting valves and differential

pressure alarms. We believe this additional attention to hydraulic

system operation will also reduce the risk of system failure and

potential fire hazards, it will also assure rapid detection by

the operators of system leakage.

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2.2 Piping and Affected Components

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The staff was requested to review piping and other components that

may have been affected by exposure to cold fire water, giving

consideration to impurities such as chlorides.

The major components affected by the fire are as follows:

- Hot reheat steam piping t

- Hot reheat relief valves, and  !

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Main steam relief valves l

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The evaluation of each component is provided as follows: [

Hot Reheat Steam Piping

The hot reheat steam piping in the fire area is fabricated from  !

plate, rolled and welded to form a 34 inch diameter pipe. Pipe r

stubs for the hot reheat relief valves are welded to the main pipe. [

The pipe, pipe stub and relief valve body are 21: percent Cr, 1  !

percent Mo. This alloy is suitable for such high temperature service.  !

This piping was subjected to potential damage from impingement of -

cold fire water during operation at temperature. .

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In order to perform a post-event assessment of damage to this i

piping, the licensee removed all insulation down to the main pipe.  !

This exposed the relief valve body, stub pipe and main pipe. This  !

area was examined by magnetic particle testing as was done for the j

initial installation. No evidence of any damage was found. It is l

not anticipated that the materials involved would be damaged by i

thermal stresses from this event. f

Concern was raised that the combustion products from the fire could f

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contain potentially corrosive chemical species. The licensee's

analyses of the fire soot indicated that these were not highly

! corrosive. However, the licensee has used s detergent water spray

j to remove most of this material. [

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Based on the above, we find the licensee program for evaluating

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potential damage to the rehett steam piping and the results thereof,  !

are acceptable. [

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Hot Reheat Safety Valves  !

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l The hot reheat line safety valves were damaged in the fire area.  !

Although the valves did not lift, they were sub ected to high ambient

temperature and cold fire water. The most crit cal component is the

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j valve calibrated spring which is adjusted to provide the proper

relief settings. The licensee's current program includes refurbish- I

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ment of these valves and replacement or refurbishment of the calibrated  !

i springs as required, j

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There art six hot reheat safety valves affected by the fire, and  !

they are the same model 6-1706 RWE-1-103-05150 and manufactured by  :

Dresser Industries. At the October 30, 1987 meeting in Bethesda  !

between NRC staff and the licensee, the licensee stated that all six l

valves were being repaired in accordance with the plant normal  !

repair procedure, MP-1010, which includes disassembly, examination , l

replacement, and testing after reassembly. The staff has

acknowledged the licensee's approach for refurbishment of the hot  !

reheat safety valves and finds it acceptable.

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Main Steam Safety Valves

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During the cooldown following the fire, the main steam system was l

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overpressurized. This happened because control circuits for the j

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main feed water pump were lost, and the economizer inlet pressure  !

, was used to estimate the main steam pressure in Loop !. Since the .

I pump discharge pressure could not be accurately regulated by the  !

i economizer inlet pressure, the steam generator was overpressurized.  ;

j One of the Loop ! main steam safety valves opened to relieve the l

1 steam system pressure. The valve that opened that has the lowest 3

setpoint pressure and is set to open at 2720 psig. The other valves l

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with higher pressure settings did not open.  !

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The licensee later reported the valve that opened did not reseat

! properly and water leakage was found from the bonnets of the once f

I opened valve. No details of the seat leakage and possible reasons  ;

i for causing the leakage was provided by the licensee. However, in '

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assessing the damage to these valves, the licensee noted that they [

) are designed to operate with both water and steam. The valves  !

j normally experience a wide range of temperature, f rom cold water at  !

' start up to 1000' F steam at power. The valves are sized to [

acco modate full flow with solid water, as well as steam.

I Thus, the valve did not experience any unusual temperature

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transients during the cooldown following the fire. The seat leakage

) experienced by the safety valve, especially following multiple i

I openings to release two phase fluid or water, is considered normal  !'

by the state of art safety valve design. There are three sain steam

safety valves on Loop I; they are the same model 3-3740WE-103 RT-21 ,

and man.factured by Dresser Industries. The licensee verbally (

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proposed to repair the leaking valve according to its normal repair ,

procudure, MP-1010, which include disassembly, visual examination and t

I testing after reassembly. The staff finds that the licensee's

proposals for repair of the leaking main steam safety valve are  ;

j acceptable. l

) 2. 3 Effects of Combustion Products

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1 The staff was requested to review the short- and long-term implications ,

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of exposure of plant equipment to combustion products which may (

persist and be corrosive.  !

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The licensee has perforced an analysis of the combustion products

deposited throughout the piant. An analysis of these products found

them to consist of less than one percent chlorine and less than two

percent sulfur. The licensee stated that such products are geneTaily

not highly corrosive. However, the licensee has initiated a washdown

program utilizing detergent and high pressure water in an effort to

remove these products throughout the plant. Additionally, the licensee

has taken "wipe" samples in electrical control and junction boxes

throughout the plant. As the result of these samples, the licensee

has concluded that there will be no electrical problems with electrical

equipment contacts as a result of the fire,

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We have reviewed the licensee's efforts to assess and mitigate the

effects of potentially corrosive combustion products. The licensee

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has taken what we believe are effective measures and we conclude his

program is acceptable.

2.4 Control Room Ventilation System

The staff was requested to review the design basis for the control

room ventilation system to determine whether any specific short- or

long-term measures should be taken to preclude the chance that smoke

might enter the control room in the event of a future turbine building

fire.

The licensee is taking a number of corrective actions to improve the

performance of this system prior to plant restart. These include:

- Testing and/or replacement of system filters

- System modification (pressure sensing line relocation)

- Preventive shhitenance and,

- Functional testing

These are discussed below.

The licensee's plans for testing or replacing the system filters are

as follo s:

One of the two charcoal filters (F-7502) has been tested per the

guidelines of RG 1.52 and found acceptable. The other charcoal

filter (F-7504) has not been tested, but rather, the charcoal will be

replaced upon delivery 4 to 6 weeks after plant restart. The system

particulate filter (F-7503) has been replaced, Filter F-7504 does

not function unless the control room fire detectors are actuated

(purge mode). The availability of the breathing air system in the

control room compensates for the possible unavailability of of

F-7504 during this period. The licensee has also committed to

installing additional air masks in the control room prior to rise to

power.

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The licensee proposed a modification to the system to move the

pressure sensing line from the auxiliary equipment room to the

control room. This will permit a more direct control of the pressure

within the control room envelope. This modification will be completed

prior to system functional testing. In addition, the licensee has

performed preventive and corrective maintenance on the system to

assure that it is ready for further operation.

Prior to startup, the licensee will conduct functional tests of the

system to assure that the correct positive pressure is maintained in

the control room in all modes of operation. Of special interest is

the differential pressure relative to Building 10, which also operates

at a positive pressure. The licensee will evaluate the results to

these tests to determine if additional system modifications are

necessary to satisfy the system design basis. (The licensee reported

to the staff by telephone on November 25, 1987 that the control room

differential pressure tests were successful.)

We have reviewed the design basis for the Fort St. Vrain control room

ventilation system and the items described by the licensee at the

October 30, 1987 meeting. We have determined that the licensee's

evaluation to date and current plans for actions to be taken prior to

and subsequent to power operation are acceptable. We understand

that, in addition, as a condition of acceptance, the licensee will

provide by January 30, 1988:

(1) an evaluation of the existing air pressure differential between

the control room and the adjacent space in Building 10;

(2) a complete list of corrective actions taken;

(3) a commitment to taking further actions, as necessary; and

(4) approved emergency operating and surveillance procedures for

the system during specific conditions including smoke in

adjacentareas.

3.C CONCLUSIONS

Based on the above, we conclude that the licensee has taken appropriate

corrective actions following the October 2-3, 1987 fire in the Turbine

Building. The licensee's actions have addressed the following:

- Hydraulic System

- Piping and Affected Components, and

- Effects of Combustion Products, and

- Control Room Ventitation System

- Stronger controls over fire detection systems and compensatory

measures when fire detection capability is compromised.

Based on the above, we find the licensee's proposals for plant restart is

acceptable.