ML20236U180
| ML20236U180 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/23/1987 |
| From: | Robert Williams PUBLIC SERVICE CO. OF COLORADO |
| To: | Calvo J NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| References | |
| P-87414, TAC-66365, NUDOCS 8712020337 | |
| Download: ML20236U180 (40) | |
Text
- _ _ _ _ - _ _ _ _ _ _ _
.,,- /
.o hPublic Service-l@.%.
P.O. Box 840 Denver, CO.. 80201 0840 November 23, 1987 R.O. WILLIAMS, JR.
Fort St. Vrain VICE PRESIDENT Unit No. 1 NUCLEAR OPERATIONS P-87414 1
1 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk H
Washington, D.C. 20555 1
ATTN:
Mr. Jose A. Calvo
-Director, Project Directorate IV Docket No. 50-267
SUBJECT:
Response to Second Request for Additional Information Concerning' Recovery from Turbine Building Fire i
.(G-87402)
REFERENCES:
- 1) NRC Meeting Minutes, dated 11/9/87 (G-87402)
- 2) Preliminary Report on the Impact of the October 2nd Fire, i
dated 10/30/87
Dear Mr. Calvo:
Public Service Company of Colorado (PSC) herein provides formal responses to the Second Request for Additional Information Concerning Recovery from the Turbine Building Fire (Reference 1).
These items were discussed during a teleconference involving PSC, NRC-Region IV, and the NRR Project Manager on November 20, 1987.
PSC responses reflecting questions clarified during that teleconference are included in Attachments 1,'2, and 3.
l Additionally, PSC has updated Section 10 of the Preliminary Report on the Impact of the October 2nd Fire.that was submitted to NRC at the October 30, 1987 meeting (Reference 2). This s'ection represented Action Items to be completed before restart and also identified long term enhancements.
This update is included as. Attachment 4.
PDR ADOCK 05000267 Ob6b 8712O20337 871123I PDR l'.Il
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P-87414 November 23, 1987 All equipment required by the Technical Specifications for startup will be operable. The only compensatory measures that will possibly be in place are related to the fire detection system; fire watches L
will be posted according to the new Fire Protection Operability 1
Requirements.
Equipment not required by the Technical Specifications will be repaired in a manner that will not adversely affect the safe
)
operation of Fort St. Vrain.
i The Final ' Report on the Impact of the October 2nd Fire will be I
submitted to the NRC on or before January 15, 1988.
If there are any questions, please contact Mr. M. H. Holmes at (303) 480-6960.
Very truly yours, j
s R. O. Williams, Jr.
Vice President Nuclear Operations R0W/WMD/sem Attachments cc:
Regional Administrator, Region IV ATTN:
Mr. T. F. Westerman, Chief Projects Section B Mr. Robert Farrell Senior Resident Inspector Fort St. Vrain
D.
. Attachment 1 to P-87414 November 23, 1987 Page 1 of 24 i
PSC's Response to NRC's Questions i
Based on November 20, 1987 Phone Call 1
1 1
1.0 GENERAL Item 1:
What is PSC's overall schedule for full recovery from the fire?
Specify when corrective action will be complete on major systems and components.
3
Response
PSC expects that initial criticality may occur as early as November 30, 1987, but does not anticipate problems that will extend initiating criticality past December 5, 1987. A "best estimate",
therefore, is December 3, 1987.
PSC will be utilizing the start-up sign off book to coordinate bringing the plant to operational readiness.
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Corrective action for safety systems and comoonents will be completed I
prior to criticality, as required by the Fort St.
Vrain Technical Specifications.
I 2.0 HYDRAULIC SYSTEM l
Item 2.1:
Provide a survey of hydraulic system components that are'in proximity to exposed hot surfaces.
Provide a r, evaluation of additional shielding to provide separation between the hydraulic system components and the exposed hot surfaces.
Response
PSC walked-down each of the 30 valves served by the hydraulic system on November 14, 1987, in an effort to ascertain what exposed hot surfaces (i.e.,
above 500 F) are in the vicinity of these components. The " vicinity" was defined as a sphere around the valve l
with a 10-foot radius.
Walk-down of the valves includes the five valve manifolds containing the oil filter canisters, which are within a few feet of their respective valves.
During this walk-down potentially hot surfaces were noted.
Valve yokes were the majority of surfaces identified, along with other piping components.
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. to P-87414 November 23, 1987 Page 2 of 24 PSC is still in the process of analyzing each valve on a case-by-case basis to determine the feasibility of shielding / shrouding either the hydraulic valve or the hot surface itself.
Physical interferences and equipment qualification parameters are being factored into :this evaluation.
This is considered an enhancement and will not be completed pr.ior to plant restart. Appendix A of the Fire Hazards Analysis for Fort St.
Vrain addresses the hydraulic system and the associated fire _ hazards.
This is part of the licensing basis of FSV, and based on its analyses, operation of FSV can be performed in a safe manner.
Item 2.2:
l Provide either test data or analysis data to substantiate the role of the 1/32 inch orifice in limiting flow to the thermal relief valves.
Also provide the results of the inspection of the other'1/32 inch orifices.
Response
PSC has confirmed through calculations. ( Attachment 2) that the installed orifice will perform its design function of limiting flow to the thermal relief valve. With the orifice' installed, the flow rate is approximately 1.3 gpm. Without the orifice in place, the i
flow rate is approximately 16.4 gpm.
(These flow rates reflect assumptions made in Attachment 2).
l Inspections performed on the remaining 5 valves that utilize the i
l orifice / thermal relief valve configuration revealed that all orifices i
)
are in place with the exception of HV-2254.
It was found that on the I
cap side, there was no orifice or thermal relief valve installed:
NCR 87-527 was issued.
It has been determined that the relief i
I valve / orifice is not necessary on this valve because. the hydraulic l
operator fluid subject to thermal expansion is not trapped.
Therefore, HV-2254 is considered fully operational.
l Item 2.3:
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Provide surveillance results for tests on the hydraulic system protective features, including the 6 gpm flow limiting valves to each I
header and the associated differential alarm pressure switches.
Response
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November 23,~1987-
.Page 3 of.24 SR.
5.3'.5-A
. performs the fan'nual ' calibration.for the. pressure i
indicators and low pres'sure alarms on the hydraulic' oil accumulators'
'l pressurizing. gas and on the. hydraulic power supply liries. 'This was last performed on April l 16, 1987...SR.5.3.5-Q.is the-quarterly >
functional test for the pressure. indicators and low pressure alarms j
on the hydraulic accumulators', pressurizing gas and on the hydraulic' 4
power lines. This. test was last performed on Sep_tember 17, 1987.
U
'l There is no calibrationior functional test presently conducted.on the l
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6 gpm flow limiting valves ~. However, PSC has concluded-that the 1
l valves adequately performed their design function.of 1imitingLflow;
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out of.the damaged headers in order for'the rest of the valves-in the loop to. stroke.
This was evidenced by the fact that all-the; valves not directly affected by the fire were able to correctly position themselves during the period of hydraulic fluid expulsion.
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PSC will. perform the quarterly functional testing. prior to restart.
Test results will be available during the week of November. 30,-
1987.
Additionally, PSC is procuring appropriate: instrumentation that'will facilitate flow limiting valve testing, although ~'this will not be-accomplished prior to restart.
s 3.0 CONTROL ROOM VENTILATION SYSTEM Item 3.1:
Provide the results of PSC's tests on the ventilation system filters.
4
Response
As discussed 'on November 20, 19?? PSC has_ conducted. testing on F-l 7502 (the control room HEPA filter), which is.the filter'that will be I
included in the Technical Specification Upgrade Program. :The testing was in accordance with Reg. Guide 1.52, according to:-the: testing i
organization, NCS Corporation, and the.results indicate'that this-i filter arrangement is acceptable (See' excerpts as Attachment 3).
PSC has replaced the.'12 pre-and 12 Lpost-: filters for F-7503, although no testing was' performed.
Replacement charcoal for F-7504
-i has been ordered, although delivery cannot be expected for 4-6 weeks.
However, F-7504 is not considered a start-up_ impairment since F-7504-does not function unless the control rcom fire-detectors are actuated.
With the breathing-air system installed 1..
the control-i
- room, PSC is confident that operations can commence without F-7504 being. replaced.
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Attachment I to P-87414 November 23, 1987 Page 4 of 24 Items 3.2 and 3.3:
3.2:
Provide a complete list of corrective actions taken for this system.
3.3:
Provide an evaluation of the existing pressure differential between the control room and the adjacent space in Building 10.
(Describe any corrective actions planned for this problem under 3.2 above.)
Responses:
PSC has initiated Change Notice No. 2713 to relocate the differential pressure sensing line from the auxiliary electric room to the control room.
Additionally, PSC has conducted preventive maintenance throughout the' control room HVAC system in preparation for the j
conducting of a functional test scheduled for the week of November i
23, 1987.
This test will determine whether the control room remains at a positive pressure relative to the turbine building during normal, economy, and high radiation modes.
Transients and purge modes-will also be investigated.
.j The control room and Building 10 are both positive pressure areas.
There are two possible flow paths from Building 10 into the control room.
PSC will be conducting testing during the week of November 23,
-l 1987, with the NRC Senior Resident Inspector observing, and will i
better understand the actual performance of the control room HVAC system with the sensor relocated at that time. Aaditional corrective 1
actions will be determined, if needed, based upon those test results.
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4.0 SYSTEMS AND COMPONENTS
'i Item 4.1:
I provide the valve manufacturers and design information on all valves affected by fire or cocidown transients.
Response
I I
Hot Reheat Safeties V-5225 through V-5230:
Dresser 6-1706 RWE-1-103-5-05150 Maxiflow Safety Valve (6" inlet, 8" outlet) 1 Body and Bonnet Material: ASTM A217, 1
Gr. WC 9 ( 2 1/4 Cr, 1 Mo)
Set Points:
V-5225:
700 21 psig V-5226:
705 21 psig V-6227:
710 21 psig V-5228:
715 21 psig V-5229:
720 21 psig V-5230:
725 21 psig
'I Design:
150-610 psig 1002 F
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Main Steam Safeties V-2214, V-2215, V-2216:
Dresser 3-3740WE-103-RT-21 Maxiflow Safety Valve (3" inlet, 6" outlet).
Body and Bonnet Material: 2 1/4 Cr, 1 Mo l
Set Points:
V-2214:
.2850 2% psig V-2215:
2790 2's psig V-2216:
2720 2% psig l
Design:
2440 psig 1005 F FV-2206 Masoneilan 57-20721.
8" 2500#
Body and Bonnet Material: ASTM A216, Gr. WCB i
E_.____________________________._____.___
. to P-87414 November 23, 1987 Page 6 of 24 Globe Electric / Hydraulic Valve Design:
3280 psig 403 F HV-2292 1
i Rockwell 4414 WC 9 JMMY.
6" 2500#
l Body and Bonnet Material: ASTM A217 Gr. WC 9 Y Globe Electric / Hydraulic Valve Design:
2500 psig 1000 F Item 4.2:
Provide a sequence of events for the lifting of the main steam relief j
valves. Were the valves subjected to higher than normal temperature changes (thermal stresses)?
Response
Secuence of Events:
1 Note:
Main Steam Pressure indication in Loop I failed high during 1
the fire.
Economizer inlet pressure is used to estimate the main steam pressure in Loop I.
1 2348:35 i
HV-2292, the Loop II startup bypass block valve is closed.
Loop II l
is transferred to the main steam bypass pressure control.
0000:02 Economizer inlet pressure is 2587 psi (this is normal). Main steam temperature is 802 F.
Feedwater inlet temperature is 233 F.
0006:23 Economizer inlet pressure begins to drop due to loss of steam to drive A boiler feed pump.
0006:59 Economizer inlet pressure is 706 psi.
Main steam temperature is 801 U.
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November 23,.1987 j
Page.7 of 24
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l 0008:22 Economizer inlet pressure returns to normal.
l 0008:27 Reactor Operator scrams the reactor.
0008:32 Jconomizer inlet pressure.is 2081 psi.
Loop 1 main steam temperature J
has dropped to 792 F.
l 0010:40 i
'B' boiler feed pump (electric) is start'ed to assure a supply of l
feedwater to the steam generators.
Economizer inlet pressure is 2852 l
psi. The Reactor Operator maintains the economizer inlet pressure to l
assure feedwater flow through the steam generator.
Loop I main steam temperature is 771 F.
V-2216, the operable main steam safety relieves to control pressure at 2720 psi.
l 0011 to 0031 l
The economizer inlet pressure gradually decreases from 2800 to 2400 psi.
0016 l
Loop 1 main steam temperature drops to 624 F, below saturation temperature.
i 0031:48 l
The Reactor Operator determines from other indications that he has feedwater flow through the steam generator and throttles the discharge of
'B' boiler feed pump.
Economizer inlet pressure drops from 2346 to 1540 psi.
Main steam temperature is 615 F.
inlet temperature is 241 F.
Approximately 0200 Reactor Equipment Operator making-routine rounds discovers the Loop I main steam safeties leaking from the bonnets.
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Background:
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. to P-87414 November 23, 1987-Page 8 of 24 During start-up, main steam pressure is controlled at 1600 psi.
Feedwater in the main steam piping is bypassed to the cold reheat header through the startup bypass flash tanks. The pressure control valve on this system is designed for flashing conditions..The steam enters the cold reheat header and the water is returned to the condenser. As reactor power is increased, the pressure control on the main steam header is increased to 2400 psi.
The temperature of the feedwater in'the main steam header will increase as reactor power is increased.
When-the feedwater in the main steam header reaches 662 F,
the saturation temperature of water at 2400 psi, the temperature will remain constant as the reactor power increases.
During this time, there will be two phase flow in the main steam lines.
As the reactor power increases, the quality of the steam improves and eventually begins to superheat. At approximately 100 F=
superheat, when the main steam temperature exceeds 760 F the main steam pressure control is transferred from the startup bypass flash tank to the main steam bypass flash tank. At this time HV-2292 is closed.
During the shutdown of the reactor, the main steam header remains on the bypass pressure control until the temperature falls below 800 F.
At this time, the control is automatically transferred to the startup bypass control system. As the main steam temperature drops to the saturation temperature, the main steam header will begin to contain wet steam. As more heat is removed from the reactor, the main steam piping will go solid.
This occurs shortly after a reactor scram, i
The time depends on the reactor power and the rate at which the heat is removed from the reactor following the scram.
This is a normal occurrence.
Thermal Analysis:
l Following a reactor scram from full power, the main steam temperature peaks at 1030 F in 5 minutes. The temperature drcps below 1000 F-after 2 minutes after the peak.
Wet steam is emitted from the superheater after 11 minutes and saturated water after 14 minutes.
The system sees a transient from 1030 F to 622 F in a period of 9 minutes.
The temperature of the main steam piping dropped from 802 F to 615 F in 10 minutes following the scram from 27% power on 10/3/87.
The temperature transient was not unusual.
The temperature transient was less severe than a scram from full power.
The steam generator was l
supplied with feedwater of a constant temperature throughout the I
I
. to P-87414 November 23,,1987 Page 9aof 24 CONCLUSION The main steam safeties did not experience an excessive temperature transient.
The increase in pressure caused by 'B' boiler feed pump had no effect on the temperature transient seen by the main steam safeties.
Item 4.3:
1 i
Provide a description of the fabrication of the hot reheat steam-line in the fire zone.
Provide 'a description of the post fire inspection of this line and relate to the fabrication methods and previous acceptance criteria.
Response
Hot Reheat Line - Fabrication and Associated Acceptance Criteria The hot reheat line which was exposed to the fire zone is shown in Figure 1.
This section of hot reheat-line was insulated and covered up to the base of each of the six hot reheat' safety valves. The hot reheat safety valves are connected to the 34 inch diameter, 1.511 inch minimum wall thickness, hot reheat line by short, 8 1/4 inch 00, 1.511 inch wall thickness, minimum, attachment lines.
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1 November 23, 1987 j
'Page 10 of 24 c.;
The following is a-listing off the applicable required codes and j
2 specifications governing the hot.reheatLline and: attachment lines.
H 1
. Piping Component
' Code's"and Standards:
j
]h
- Hot Reheat Line
' - ANSI'B3.1'.1,=1967, Power-Fort St. Vrain Line
' Piping. Code 34" L 5216-D6.;
- Fort St..Vrain; Specification:
34" OD, 1.511 inch
'l-M-2, Piping. Design Class minimum wall thickness D-6
- Fort St.,Vrain Specification:
'91-M-50
- MaterialESpecification,. ASTM
-A387, GR22 and. ASTM.
' A691 (welded plate)J
- Hot Reheat Connection to
- ANSI B31.1,.1967, Power.
Hot Reheat Safety Valves Piping' Code 8 1/4 inch OD, 1.511 inch
- Fort St. Vrain Specification 1
minimum wall thickness 1-M-2, Piping Design Class
,D-6
- Fort St..Vrain Specif' cation i
91-M-50
- Material Specification, ASTM
.ASTMA182,F22(forgedfitting)
(
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m
-m__.u.m_m.
._-__mu
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Attachment l'to P-87414 November 23,~1987 Page 11 of 24 Based on the codes and standards governing these materials the following are the required tests and inspections for the~ hot reheat line and attachment lines.
')
Piping Component Tests / Inspections Required
- Hot Reheat Line
- Chemical check analysis of each a
Fort St. Vrain Line length per ASTM A155 i
34" L 5216-D6
. Tension test and bend test of
)
each length per ASTM A155
)
l
- Hardness test on each end of each
]
length and on the seam weld.
I Hardness shall not exceed Brinell 217 per ASTM A155
- Photomicrograph'of each length l
of pipe, from one end,'at 100 diameters.for grain size.
- 100% ultrasonic examination 4
per MIL-STD-271:
1
- Hydrostatic test per ASTM A530.
Test pressure shall not be less i
than 1 1/2 times design pressure.
- Hot Reheat Connection'to
- Chemical check analysis and tension Hot Reheat Safety Valves test per' ASTM A182.
8 1/4 L 52105-D6 l
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. to P-87414 November _ 23, 1987 Page 12 of 24 i
Piping Component (Con't)
Tests /InspectionsRequired_LCon't)
- Hot Reheat Connection to
- 100% ultrasoni: examination of pipe i
Hot Reheat Safety Valves per Mil-STD-2710 81/4 L 52105-06 i
Welding I
In addition to the tests and inspections outlined, all fabricated fitting welds and butt welds (including shop and field welds) shall be tested as follows:
- 100% radiographer per MSS-SP-54 and 91-M-50 and i
- 100% liquid penetrant or magnetic particle inspected per I
ASTM E165 or an approved manufacturer procedure
-)
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l November.23, 1987:
'Page:13 of 24
~H_ot Reheat Line Fabrication-1 The section'of the hot' reheat line wh'ich was' exposed to.the fire zone ei is the section which contains the-six-hot-reheat' safety valves. -This section of' the. hot. reheat.line was shop fabricated at the Stearns-
-l Roger Pipe Fabrication Plant _(see Figure.2). This shop fabrication d
included welding the line connections-which would ultimately be I
connected to the hot reheat safety. valves. When the shop. fabricated section of the hot reheat line was compl.eted and in place at Fort'St.
Vrain, the hot reheat' safety valves.were attached by. welding 'in.the.
field.
i l-Post Fire Inspection' of Hot Reheat Line ~
!a i
The section of the-hot reheat line which was exposed to the fire zonei was insulated and covered.
Since thermal shock:would be ' minimal in this area, it was determined that the cionnection to the hot reheat safety valve, V-5226, which appeared to have been subjected-to !the greatest temperature, would be examined for possible fire related-I damage. This section of the hot reheat-line, 8'1/4'L.52105-D6, would-H be the. boundary ~ where the insulation,had been discontinued.
.Any' I
damage which may have occurred from thel' fire, or the subsequent extinguishment, would most likely occur in this area.
Damage to the hot reheat line material as a result of the fire would:
1 be expected to be limited to material phase transformation or thermal'
- cracking, (quench cracking).
Cracking would~ be limited to the i
materials surf ace or possibly, slightly subsurface.
Defects - of the-1 material's surface and subs'urface would.be observed'by performing nondestructive examination in the. form of' flourescent. magnetic, particle inspection.
Damage to the' hot reheat line material,-by phase transformation, could be ruled out as'a result'of. the. testing l
performed on the hot reheat safety valves. These valves are attached ~
l to the hot reheat line.but are not insulated which 1would result in i
their material being subjected to the seve-est environment.
The hot.
reheat safety valves were examined by fluorescent magnetic. particle inspection, hardness tests and nondestructive metallographic l
replication, and were found to have no appare_nt fire related. damage.
l-Therefore, the hot reheat line was tested by fluorescent magnetic l'
particle exam 1 nation.
The hot reheat' line, 81/4 L 52105-D6, was stripped of 1.ts; insulation to the connection weld of Line L5216 D-6, and cleaned by~ hand buffing:
to remove scale and surface debris. The entire surface was then fluorescent magnetic particle inspected per Fort LSt.
Vrain Quality-Control Inspection Manual, QCIM-24. This type tof inspection is very i
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- to P-87414 November 23, 1987' Page 14 of 24 l
sensitive to any surface, and shallow subsurface, defects which would l
be expected if damage had occurred as a result of the fire. This i
inspection would have greater' sensitivity to surface defects as compared to the original inspection which was applied to the material, during original fabrication.
This inspection found no rejectable -
j defects on the material.
I Based on the inspection performed on the hot reheat line, 8 1/4 j
L52105-D6, it has been determined that the hot reheat line. material 1
l is acceptable with no deficiencies which could be attributed to the
{
fire.
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Attachment'1 to P-87414 November 23,'1987 Page 18 of 24 Items 4.4 and 4.5:
1 S$
Prov'ide 'a list of all qualified equipment affected'by the fire, and identify those whose continued operation is justified by analysis.*
I i
4.5 j
i Provide a written summary of the analysis method (s) used to justify l
continued operation of affected qualified equipment.*
-)
In subsequent. discussions, the staff has deleted questions 4.4 and 4.5.
The staff has stated that. these' items should.be 1
i covered by PSC's program for equipment qualification under 10 CFR Part 50.59.
Response
PSC acknowledges that environmentally qualified equipment repairs and
.i replacements will be controlled through the Fort St. Vrain EQ Program
)
per 10 CFR 50.49. Additionally, Engineering Evaluation EE-EQ-0065 will contain the information required by the EQ Program.
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. to P-87414 j
November 23, 1987 Page 19 of 24 j
Item 4.6:
l 1
Describe the methods used to systematically cleanup combustion products throughout the plant and assure that there are no continued problems with safety related systems or components.
Response
The soot deposited was analyzed for chloride and sulfur content.
')
Chloride was found to be less than 1% and sulfur was less than 2%.
Per the Fire Protection Handbook, these amounts are not considered i
to be deleterious to plant components, since they are below the. 4 level described as mildly corrosive.
Clean-up was accomplished with high pressure water with detergents added and physical equipment wipe downs. No degrading effects are anticipated, i
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. to P-87414 November 23, 1987 Page 20 of 24 1
1 I
1 Item 4.7:
l Specifically describe how exposed electrical contacts will be checked i
for adverse impacts from combustion product deposits.
1
Response
)
Control Room The Fort St.
Vrain Turbine Building Fi re of October 2nd,.1987 provided the source of certain combustion by products which were pulled into the control room environment.
There exists no geantitative measure of the amount or concentration of by products pulled into this environment and either deposited or. exhausted to the l
outside.
Discussion with operating personnel has provided some visual measure of this concentration, This measure may be. described 1
by the following observations:
1.
Smoke in the control room was described as a light haze, greyish brown in color near the ceiling.
The thickness of this layer I
was estimated as "down to the top of the control boards" (or i
approximately 24 inches).
I 2.
The sensory effect was described as acrid and a clear air irritant.
3.
A black material (soot) formed a fan shaped deposit, three (3) feet wide on the carpet at the center joining of the double doors.
A rectangular shaped area one (1) foot wide was i
deposited along the bottom of the west door.
I An investigation has been completed to determine, first, if there are any combustion by products visible on or in the control room electrical components and second, if any materials were found, are these materials detrimental to contacts and other exposed surfaces.
l A visual inspection of various main control room. components and a series of swipe samples on November 11, 1987 provides the following:
1.
Protected horizontal surfaces have an accumulation of dust particles which appear to be light grey to brown in color.
(This material appears to be similar to materials found on office furniture remote to the control room and turbine building.)
Note:
Nineteen (19) wipe samples were taken from horizontal surfaces in an "S" pattern of approximately twelve (12) 1
-l i
1 1
. to P-87414 j
November 23, 1987 i
Page 21 of 24 inches long.
These samples were examined using a ten (10) power optical comparator.
q l
2.
Material concentrations 'on wipe samples measure approximately j
2 200 cm.
.This concentration represents a surface area of l
approximately 30,000 cm.
Several wipe samples have black
{
2 smears or spots 0.2 to 0.4 centimeters across.
The' area-ratio-of black material to the lighter materials is conservatively 6 j
to 1 million.
j i
3.
An inspection of relay contacts (to the extent visible and accessible) showed no visible foreign material.
All terminals and terminal board wiring inspected appeared to be free of-foreign materials.
Control board mounted switches, controllers, indicators, and i
recorders are by virtue of physical construction protected against I
falling particulate.
The various modules comprising the plant' protective system are housed in enclosed NIM bins.
The safety-related relays associated with this protective system are located in the lower portion of I-9310 and protected from falling or materials by the enclosed NIM bins mounted above.
settling CONCLUSION 1.
The negative differential pressure condition of the control room during operation of the HVAC system in the purge mode resulted j
in the ingress of combustion by products.
l 2.
Some part of these by products was in the form of a black particulate.
I 3.
The major portion of this material was deposited on the carpet in the immediate vicinity of the south double doors and the west door.
4.
The amount of material which settled on component surfaces is relatively insignificant in terms of the total surface exposed.
5.
The protective enclosure surrounding electrical and electronic equipment provides protection against the ingress of materials falling or settling.
No credible mechanism for forcing materials into component housings was identified.
I 1
l
Attachment'1-to P-87414 November 23, 1987 Page 22 of 24 6.
An inspection of component surfaces shows no visible evidence of foreign materials, visible surface discoloration or degradation.
Turbine Building On November 17,. 1981 an inspection was made of certain electrical-components located in the Fort St.
Vrain Turbine Building.
The purpose of this. inspection was to determine if any combustion by-products-(or other foreign material) were evident on electrical components, circuit contacts and terminals.
A total of 10 wipe samples was taken.
Each wipe sample was inspected under a 10 power optical comparator.
. Table A delineates the observations made relative to each sample.
Note:
For comparison purposes, base-samples.of dust from office furniture were taken. These samples appear as light grey to dark brown under ten. power magnification.
A base sample
(#26) was taken from~the exterior of a junction box located i
immediately west of the fire zone and at an elevation of 4821 feet.
This sample appears black in color.
l All samples taken from inside component enclosures with'the exception of #22, appear to be grey or brown in color and free of the black material contained on sample #26.
The components within the instrument air compressor auxiliary control panel (the source of sample #22) were inspected. A light film of grey material (dust) was noted, however, there was no visible deposit of materials on the relay contacts.
The switch located on the front of this panel (a SBM type) is structured such that material ingress around the exposed.
shaft is restricted.
CONCLUSION Electrical component contacts and terminals within enclosurec which are mounted within the turbine building and.outside the fire zone show no visible deposits of black materials (soot).
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Form tBI 344 24-4309
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' Attachment'2 to=P-87414
)
November 23c 1987 PubilC.
F RT ST. VRAIN NUCLEAR GENERATING STATIOiB Page 2 of 3 l
Services PUBLIC SERVICE COMPANY OF COLORADO ~
1 CALCULATION WORKSHEET l
CALCULATION FOR T'im MW Tim ou T 0F. SYSTEM 9 (-tr-A THemAL. CALCULATION NUMBER R.sLt G t-VA L.v G 15 Stut bt ops tJ - DR.t fite McT i4 ST%Lt GD PREPARED BY /A $4. W oo,.)(n lDATE l t 9 9 "")
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Form t BI 344 24-4309
1 i to P-87414
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Public FORT ST. VRAIN NUCLEAR GENERATING STATION o n mber 2 1987 service
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CALCULATION WORKSHEET CALCULATION FOR %C. FLo u) 12 Ave ouT~ 01-tt4 6 F Aits.3 Fttygg CALCULAT60N NUMBER c.AmGrt=/L'S Rumas, q
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. Attachment'3 to P-87414 November 23, 1987 4
'Page 1 of 6'-
I t
EPORT FOR IN4UE TESTIM 0-a NJCLEAR AIR QI.ANim SYSTEMS J
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1 FORT ST. VRAIN NJQfAR STATIM PLATTEVILLE, 010RADO t
i TESTS PERFORfD BY:
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ - - - - - - - " ~
. to P-87414' i
November 23, 1987 n C 5 CORPORATION 4555 Groves Rd. No. 41 Columbus Ohio 43232 614 864 7613 RADIOIODINE PENETRATION AND RETENTION TEST REPORT Public Service of Colorado PURCHASE ORDER NO. N7808 CLIENT Fort St. Vrain Nuclear Station TEST REPORT NO.
N 04 Platteville, Colorado 80651 SAMPLE NO.
87-2080 SAMPLE IDENTITY Control Room Date Sampled: 10/14/87 Date Tested: 10/26/87 TEST CONDITIONS:
4 Temperature 130'C Duration of Post Sweep 240 min.
Pressure 399 KPa Pre Equilibration Time Thermal Only Relative Humidity 95%
i"' Content 1.7 uci Face Velocity 12.2 m/ min.
Chemical Form I"'
METHYL IODIDE 1
Adsorbate Concentration 1.75 mR/L3 Standard Deviation of l
Total Counting Error 0.002 Duration of Loading 60 min.
RADIOIODINE TEST RESULTS AT:
% RETENTION
% EFFICIENCY 1"
0.10 99 90 2"
4" All test procedures and methods are in accordance with RDT M161T June 1972, October 1973, December 1977; ASTM-D 38031979 and Reg. Guide 1.52 and ANSI N509/N510.
l Standard Deviation NCS Distribution 1st 2" Canister 196.875 (1) Lab File 2nd 2" Canister 6.567 (1)OA File 3rd 2" Canister i.619 4th 2" Canister
./ Qv#
,/
TEST APpa0Vio ev V
TESTPERFOped[
f to P-87414--
l November 23, 1987 l
Page 3 of 6 l
NUCLEAR CONTAINMENT SYSTEMS, INC.
F-0Cl VISUAL INSPECTION OF FILTER. SYSTEM l
PUBLIC SERVICE COMPANY OF COLORADO CUSTOMER 16805'RD. 19i PLATTEVILLE, CO 80651 i
PURCHASE ORDER NO. N7808 NCS JOB NO.
T356 i
LOCATION FT. ST. VRAIN SYSTEM CONTROL ROOM
{/
NUMBER OF FILTERS 2 HEPA 4 CARBON 2000 CFM FLOW l
FILTER TYPE FLANDERS BC DATE OF INSPECTION 10/14/87 CHECXLIST FOR VISUAL INSPECTION
?
4 sf
.+
9 Housing
/ X/
//
/ /
a.
Adequate space for personnel and equipment for maintenance and testing, y
/ X/
//
/ /
b.
Reasonable access to housing.
j
/ X/
//
/ /
c.
Doors or access opening are adequately closed sealed.
/ X/
//
/ /
d.
Housekeeping in and around housing.
l
/ X/
//
/ /
e.
Condition of flexible connection between housi l
and fan external to. housing.
j
/ X/
//
/ /
f.
Fan-shaft coal.
l
/ X/
//
/ /
g.
Dampers' are in position to prevent any. bypassi of system.
Remarks (a).
to P-87414 November 23, 1987 Page 4 of 6.
/
/.//
8 Dampers'
/ /
/ /
/n/
a.
Damage to any frames or operating mechanisms.
X
/ /
/ /
//
b.
Missing seats o-blade edging.
j X
//
/ / ' //
c.
Condition of rest 11ent. seats or edging Remarks Drains
/X/
/ /
//
a.
All drains are closed to provent any bypassing of filters.
Cemarks 1
Pressure Gauges X
//
/ /
//
a.
All gauges are operating.
/X /
/ /. //
b.
No excessive AP appears.across the,prefilters.
//
/ /
//
c.
No excessive AP appears across the HEPA filters.
/X /
/ /
//
d.
No excessive dP appears across the carbon adsorbe j l
Remarks COMMENTS:
NO APPAhlENT DAMAGE DUE TO RECENT FIRE.
p TEST PERFORMED BY:
Wl:M d /TL_
o (b).
~
' ' 'to P-87414:
November 23,l'1987
.F-002 Page 5 of 6 PARTICULATE TIL20t IN-PLACE LEAK MST RDORT c23:QgR
-PUBLIC SERVICE CO. OF COLORADO 18605 RD., 19i PLATTEVILLE,'CO 80651.
PCRGASE CRCER NO.. N7808 pas h m,T-356 mqcy FT. ST. VRAIN m
CONTROL ROOM ICMBER G. FILTERS 2
M 2000 CFM M HEPA E M 0.2" WG mg FLANDERS DATC 7 TEST 10/14/87 L
TEST R EJLTS:
o o4
..w m m 99 95
. m,wIm mI-(CCP ADOSCL)
RD! ARKS I
E M ICUIPMCfT TDA-2EN SN/5536 l
(
CALIEFATICN CATE 6/18/87 CALIIR,L""ACN CCE CATE 6/18/88 T!s: PERrtssa sy % //e j I k y --- -
ses mmecu <
Ril. 0 7/8/86 to P-87414' November ~23,-1987 i
Page 6 of 6 i
CUGEN AESCREDt FILTDL IN-PLACE IDK MST RII:PtstT i
m PUBLIC SERV ~CE CO. OF COLORADO
)
16805 RD. 19i 1
PLATTEVILLE,.
l CO 80651-I l
m a yo,N7808 g a g, T-356 FT. ST. VRAIN g,y m ____ CONTROL ROOM 1
N 7 FILTDU 2'X 2 CARBON l
Ff.G 2000 CFM l
AP C;UUICN hq 0.4" WG FILIDI T PE BC I
CATE & ':EST 10/14/87 I
i.
1 TEST RESCL'IS:
<0.01
%.MEICAL IDJGGE
> 99. 99 %.MNICAL DTICIENCY l
l 3EMAR5.3 l
NCS TEST EUIPMcN LMP-10 SN/0006 cAI.:seATIm cx:s 8/16/87 CALIsaAT:m ace cats
'8/16/88
'IssT rzRrtRMED sr 'Y3 M T/L -
l ucs c:arca m m i M.O F-003 7/8/86
l 2
}
L Attachment 4Lto P-87414
' November 23; 1987-J
.PageL1. of?6 a
1 ai ATTACHMENT 4-J c
10.0 UPDATED ACTION' ITEMS-
.i
-; \\
10.1 Prior to Rise to Power Restore Pla'nt Systems i
This item indicates the-return-of.
fire affected components / systems to operation.
STATUS:
In progress j
Fix HV-2292 Hydraulic's '(flow valve / orifice)'
This item includes the repair of :the hydraulic actuator 'for' HV -
I 2292,-involving the replacement of both'the thermal relief valve and the orifice.
-i STATUS:
In progress Verify Orifice Installations
.l This item refers to' a' ph'sical inspection of the other 5-y 1
hydraulic valves which utilize the thermal. relief _ ' valve / orifice configuration.to positively determine: that these items.aret j
installed.
j STATUS:
Complete Procedure to Ban use of Pipe Wrenches on Oil Filter Canisters This item originally required'the issuance of a new maintenance ~
procedure to prohibit pipe wrenches to alleviate marring'of the filter ' canisters.
This has been changed to revise SMAP-29A to-ben pipe wrenches and require a strap wrench.
STATUS:
Complete Install Oil Bleed Line for Canisters This item was originally intended to eliminate the need for a pipe wrench to overcome system pressure'. when' removing..a canister.
This.has been-determined to be unnecessary because; the capability exists to vent the system through the' installed l
l 4
f I
w
i to P-87414 November 23, 1987 Page 2 of.6 5-valve manifold.
The standard clearance points for system 91 maintenance has been updated to reflect this capability.
STATUS:
Complete Cleanup Hydraulic System
+
This item indicates the evaluation of the hydraulic fluid in the system and restoration to within specifications.
The ' hydraulic j
fluid was subsequently tested and is within specifications.
STATUS: Complete Modify Control Room Pressure Tap This item refers to the relocation of the differential pressure l
sensing line from the auxiliary electric room to the control l
room.
l STATUS:
In progress Verify Control Room HVAC Performance l
This item refers to a functional test verifying positive pressure in the control room.
STATUS:
In progress Perform Control Room Filter Surveillance l
This item refers to the prassure drop testing of F-7502.
STATUS:
Complete Perform Reactor Building Filter Surveillance
+
This item refers to the pressure drop check per SR 5.5.3d w, and the check of the filters by an off-site agency.
S"ATUS.
Complete Initiate Management Directive on Fire Alarms Issue New Procecure on Fire Protection Operability
+
I
. to P-87414 November 23, 1987 Page 3 of 6 Determine Compensatory Actions Associated with the Fire Detection Systems I
These items serve to provide instructions for operators relative to the fire detection systems and, particularly, nuisance alarms,
~
STATUS:
Complete - (Operatians Order 87-14 Issued)
Partially Complete - (Fire Protection Operability Requirements -
Including Compensatory Measures - received PORC approval and l
currently being trained on).
Install Hydraulic 011 Storage Lockers i
This item will alleviate undesirable storage practices for i
hydraulic oil.
i STATUS:
In progress Review SSR(s) for Missing Handwheels i
This item was established to determine missing handwheels in System 91 that were included in the SSR process.
j l
STATUS: Complete j
Evaluate / Replace Plastic Valve Handles This item will replace plastic valve handles from the accumulators in System 91.
STATUS:
In progress Install Additional Masks in the Control Room l
This item will and two additional breathing air system masks and hoses in the control room.
(although there are 6 ports, the system will only support 5 users).
STATUS:
Complete Submit LER This item satisfies the 30 day reporting requirements of 10 CFR 50.73.
STATUS:
Complete l
l 1
l
1
]
' to P-87414 November 23, 1987 Page 4 of 6 New Items i
Perform. Hydraulic System Functional Test l
This refers to the performance of the quarterly test, SR 5.3.5-Q.
STATUS:
In progress Verify System 91 Handwheels are Installed I
This item will be accomplished -during pre-startup valve lineups / system walkdowns.
1 STATUS:
To be scheduled.
j l
f
d i
.1
' to P-87414 November 23, 1987 Page 5 of 6 j
10.2 Next Steps
{
l Replace Oil Filter Canisters
]
This item refers to the procurement of new filter canisters to replace the ones with pipe wrench marks.
STATUS: On order Evaluate Removal of Thermal Relief Valve i
This item indicates an on going analysis suggested by,.PSC's l
hydraulic system consultant to alleviate possible hydraulic j
leakage.
STATUS:
To be performed Enhance Pre-fire Plans I
l This items refers to upgrading fire brigade effectiveness.
l l
STATUS:
Ongoing Evaluate Fire Detection System Enhancements 1
This item refers to the complete analysis of the present system to determine improvements.
)
STATUS: Ongoing I
Evaluate Suppression Needs for Hydraulic 011 Hazard Areas i
This item indicates a " lessons-learned" approach beyond the Appendix A analysis.
l STATUS: To be performed l
Evaluate Catch Basin Enhancements This item refers to the evaluation of the catch basins for possible inclusion of flame arrestors.
STATUS:
To be performed
. to.P-87414' i
November 23, 1987 Page 6 of 6-Issue Final Fire Report This item indicates the intent to summarize all issues in one l
final report.
i i
STATUS: To be submitted on or before January 15, 1988.
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