ML19225B368
| ML19225B368 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 06/22/1979 |
| From: | Trammell C Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TAC-11299, NUDOCS 7907250071 | |
| Download: ML19225B368 (13) | |
Text
b/ N g
o, UNITED STATES y
),, ( ',c, NUCLEAR REGULATORY COMMISSION
- E W ASHINGTON. D. C. 20555
^
June 22, 1979 Docket No. 50-344 LICENSEE:
PORTLAND GENERAL ELECTRIC COMPANY FACILITY:
TROJAN NUCLEAR PLANT
SUBJECT:
SUMMARY
OF MEETING HELD ON JUNE 8, 1979 TO DISCUSS ENVIRONMENTAL QUALIFICATION OF PRESSURE AND DIFFERENTIAL PRESSURE TRANSMITTERS INSIDE CONTAINMENT I nt roducti on On June 8,1979, the NRC staff met with representatives of Portland General Electric Company (PGE) and Westinghouse Electric Corporation (W) to discuss the environmental qualification of pressure and differentiaT pressure transmitters inside containment at Trojan.
See attachment 1 for a list of atter. dees. Highlights of the meeting are summarized below.
Background
Beginning in 1975, W initiated a " Supplemental Seismic and Environmental Qualification Program" for certain instruments located inside containment.
TI.e 32 plants referaving this program are:
DOR PLANTS (8)
DPM PLANTS (24)
T roj an Beaver Valley 2 Indian Point 3 Diablo Canyon 1 & 2 Beaver Valley 1 Salem 2 Salen 1 Sequoyah 1 & 2 North Anna 1 North Anna 2 Farley 1 McGuire 1 & 2 D. C. Cook 1 & 2 Farley 2 Watts Bar 1 & 2 Summer Catawba 1 & 2 Shearon Harris 1-4 Vogtle 1 & 2 Millstone 3 Jamesport 1 & 2 m
7 907250 D114 i fa 02!
Meeting Summary For Trojan Nuclear Plant The program was a retesting progral; for instrunent transmitters used by plants in the program done at the request of NRC to demonstrate their capability to perfom required functions - either reattor trip or long-tem post-accident nonitoring - under more severe environmental conditions than previously employed.
The retesting of the original models of these transmitters proved unsuccessful for long-tem nonitoring functions and, as a consequence, Westinghouse issued ansnitter replace-ment recommendations to plaids within the scope of this program.
Further, on nore recent license applications (D. C. Cook anu Noi n Anna),
the staff introduced the additional requirements that sequential testing be employed and that a minimum of one hour oper ability be denonsti mted for transmitters employed for short-tem reactor trip and/or safety injection automatic protective functions in a high energy liae break environnent. Prior to this change in NRC requirements, Westinghouse had considered that transmitters qualified for shorter periods than one hour were capable of perfoming short-term functions.
As a consequence of these additional staff requirements, the infomation contained in WCAP-7410L* justifying the qualification of Barton, Foxboro and Fisher-Porter pressure and differential pressure transmitters for short-tem safety related high energy line break applications was with-drawn by W fron consideration within the Suypplemental Program.
Westing-house is issuing additional transmitter replacement recommendations to the plants within this program to reflect this change.
The transmitter qualification test results contained in WCAP-7410L will not be relied upon for instrument qualification for the plants in this program, incl uding Trojan.
As a result of these developments, PGE was requested to meet with the NRC staff to present the exact nature of the environmental qualification of these instruments presently inside the Trojan containment, and to describe the backup instruments which are available outside containment to perform long-tern post-accident monitoring functions.
- WCAP-7410L is the proprietary version of WCAP-774..
416 022
Meeting Summary for Trojan Nuclear Plant,iscussion Trojan has 29 pressure / level transmitters inside containnent with less environmental qualification than required today.* When W initiated the Supplemental Qualification Program in 1975, PGE committed to follow this program and replace any transmitters that subsequently proved to be less than fully qualified.
Based on }[ reconnendations, PGE is in the process of doing this now.
"GE has ordered Barton 763/764 Lot 2 transmitters which are expected to
, ass all qualification tests by July 1979.
Testing is presently under-way.
PGE expects to take delivery of these new transmitters about July 15, 1979, and plans to install them (following NRC approval of the qualification tests now underway) during the next refueling outage (January / February 1980).** The full environmental qualification test report is scheduled to be canpleted in September 1979.
PGE will be visiting the Barton nanufacturing facility to confina the present delivery schedule.
The 29 instruments involved are as follows:
Process Variable Measured No. of Instruments Steam flow 8
Stean generator level 12 Pressurizer level 3
Pressurizer pressure (narrow range) 4 Reactor coolant loop pressure (wide range) 2 Total 29
- Current requirenents include steam-break environment and long-term post-accident monitoring capability.
- PGE could also schedule a shutdown prior to this date to change the transmitters, and in facto offered to do so.
An outag of about 10 days would be needed.
416 023
Meeting Summary for Trojan Nuclear Plant 4-For all these instruments, there is not a high degree of confidence that they will function for any substantial length of time in a post-LOCA environment.
These instruments, their qualification, function and backup instruments are discussed in turn below.
1.
Steam flow.
Barton 384.
These differential pressure instruments are used to measure steam line flow and furnish a signal to initiate safety injection (SI) at 1107, of nomal steam flow in coincidence with either low-lew T
or low steam generator pressure (steam line break protection) signal s.
This, as do all other SI actuations, also causes a direct reactor trip and containnent isolation.
(The sensors neasuring T
are qualified for the LOCA enviroment.
Steam pressure sensors aN910cated outside of containment; there 7 qualification for a LOCA environment is not necessary.)
These steam flow transnitters were origir. ally qualified by a Franklin Institute test (Report F-C-2623 dated Septenber 1969*).
The test conditions were 286 F and 60 psia.
These instruments were not tested in a radiation environment.
However, this is not an important consideration, since their function is to cause an SI actuation on high stean flow (steam line break), a condi-tion which should not produce a significant radiation environment at least for short-tern protective action.
Therefore sequential,
testing, involving radiation plus temperature / pressure / stean environments is not an impartant consideration for these instrunents.
At the time Trojan was licensed (Decenber 1975), it was assumed tha.t these instrunents would acceptably detect high steam flow (in a matter of seconds) and actuate well before being subjected to severe steam envi nnent.
This assumption is probably a good one today.
Nevertheless, for plants now being reviewed for operating licenses the staff conservatively requires that such instrunents successfully withstand a full steam-break environment for one hour and still be capable of initiating protective action (SI signal).
The Franklin test report states that one Barton 384 started to deteriorate in slightly more than one minute; another
- This test was conducted for Westinghouse, who furnished the cited report to PGE as proof of environnental qualification of these i nst rument s.
Neither the report nor the Barton 384 transmitters are a part of WCAP-7410L.
This report by Franklin Institute has not been reviewed by the NRC Staff.
416 024
Meeting Summary for Trojan Nuclear Plant operated sucessfully for 15 minutes.
Thus, op' Jation for a mini-mum of one minute at 286 F and 60 psia was dersnstrated by test.
The following backup instruments for protective action are avail-ahle for these steam flow instruments (in the form of diverse signals to detect a steam line break): high containment pressure and steam line differential pressure.
These instruments are located outside of containment.
The steam flow instrunents play no significant role in post-LOCA l ong-term monitoring.
2.
Steam Generator Level.
Barton 384.
These are the same Instruments discussed above except that they are used to neasure steam generator water level.
These instruments are used, in conjunction with other instruments, to initiate a reactor trip on (1) low steam generator water level in coincidence with steam flow / feed flow mismatch, and (2) low-low steam generator water level.
They also initiate a turbine trip on high stean generator water level.
None of these plant conditions would be expected to subject these instruments to a severe environ-ment with the exception of a feedwater line break.
Again, it is expected that they would perform their short-term pr +ective function well prior to being subjectd to a hostile steam environment or submergence.
Again, backup protective instrunentation located outside containment is available for feedwater line break in the form of high containment tressure (5 psig) SI actuation and feed-water line isolation.
As to post-LOCA long-tern monitoring, these steam generator water level instrwnents co"1d be of significant value to the operator for certain accident conditions as an aid for decay heat removal.
Backup instruments for this purpose could be made available outside conu in-nent.
Steam generator level indication would be available by in-stalling a differential pressure instrument between the steam sanple and blowdown sample lines in the auxiliary building.
By operating valves, one steam generator level at a time could be read outside containment.
PGE will confirm the feasibility of this system.
Additional backup instruments outside containment exist in the form of steam generator pressure transmitters by which it could be crudely determined whether or not the steam generator contained water ( rough go-no-go).
Also, the steam generators could be filled with water under energency conditions.
416 025
Meeting Summary for Trojan Nuclear flant 3.
P ressurizer Level.
Foxboro E13DH.
These transmitters were designed for a short-term protective function - SI actuation on coincidence of low pressurizer level and low pressurizer pressure.
However, PGE has conmitted to place these channels in the tripped mode during power operation in response tc IE Bulletin 79-06A Revision 1, and has further requested a l' ise amendner' which would renove pressurizer level as an input to SI actuation.
Therefore, pressurizer level has no short-term protective function at present or in the foreseeable furture.
However, pressurizer level indication could be of significant value to the operator for post-LOCA long-term monitoring.
These instruments have been environmentally tested per WCAP-8541 at 286 F and 60 psig for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. There is no proof, however, of longer-term post-accident operability.
Backup instruments are available.
See which was handed out by PGE at the meeting.
ThJ pref-erable backup instrument, which could be installed, would be measure-ment of differential pressure between the pressurizer steam space and liquid space by which pressurizer level could be inferred.
PGE will confirm that this instrument could be installed in a short period of time following a LOCA and is otherwise a feasible backup i nst rume nt.
The presence of a steam bubble could also be inferred (crudely) from a conparison of pressurizer surge lir.e and pressurizer stean space temperature.
See attachment 2.
PGE was uncertain as to the environmental qualification of these instruments, but stated that the basic "nstrument (RTD) was rugged and any needed upgrading of electricas connections would be straightforward and could be read ily accomplished.
4.
P ressurizer P ressure.
Barton 393.
These are narrow-range pressure transmitters used to initiate both reactor trip and SI on low pressure.
They are included in WCAP-7*,10L, and were tested successfully in 1972 at 286 F and 75 psia for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and also subjected to a radiation environment of about 2x10exp8 rads in separate effects testing.*
- The instrunent actually tested was a Barton 386 differential pressure transmitter, which is very similar to the Barton 393 pressure transnitter.
416 026
Meeting Summary for Trojan Nuclear Plant The Barton 393 was later subjected to sequential tests as part of the W supplemental qualification program.
Test conditions were 320 F, 90 psia for 25 minutes ai,d 4x10exp7 rads.
The instrument failed this test, but did last greater than one minute.
This instrument could be useful for long-tem post-accident moni-t oring, but because of its narrow range, the wide-range reactor coolant pressure instruments would be of greater value, as discussed bel ow.
5.
Reactor Coolant System (Loop) Pressure.
Barton 389 (wide range).
These instruments (2) serve no protective functions for either SI actuation or reactor trip. However, they would be of significant value fur long-term pos accident monitoring.
In this case, though, numerous backup instrune.s are available in the control roon to indicate RCS pressure.
See attachment 2.
In addition to those items listed, indication of pressurizer temperature could allow the operator to establish F.CS pressure with steam tables aided that pressurizer level indication and pressurizer heaters re available.
The Barton 389 transmit:ers have not been tested for a LOCA environ-ment.
Their post-accident monitoring capability is unknown.
- Further, these instruments are located at elev. 48' in containment and would be submerged under post-accident LOCA conditions.
No submergence testing has been conducted.
PGE stated that the containment iso-lation values which would need to be opened to read RCS pressure via the sample lines are environmentally qualified motor-operated valves.
The staff suggested that the feasibility of using the dead-weight tester line as additional pressure indication be pursued.
PGE said that they would do so.
Conclusion At the conclusion of the meeting there was a general consensus that the present instruments are adequate for the present for short-tem protec-tive action (until they can be replaced with the Barton 763/764 mod el s),
but some concern remained about the pressurizer pressure instrunents (Barton 393 narrow range).
As for long-term monitoring, none of these instronents can be considered reliable.
416 027
Meeting Sunnary for Trojan Nuclear Plant PGE was ercouragM to replace these instruments as soon as possible -
with Barton 763/764 Lot 1 instruments if available.*
PGE was requested to docnent the infomation provided at the neeting in their upconing response to IE Bulletin 79-01, and to test backup diff-erential pressure measurenents that may be relied on ( pressurizer level, stean generator level) to assure the feasibility of these alternate devices.
Charles M. Trannell, Project Manager Operating Reactors Eranch #1 Division of Operating Reactors Attachments:
1.
List of Attendees 2.
Responses to Questions Oc: w/attachnents See next page
- These instruments have passed more stringent tests than existing Trojan
'ostrurents, and the lot l's can be changed to Lot 2's (which ar<> axpected os all tests) by changing an electronic card.
416 028
Meeting Summary for Trojan Nuclear Plant Docket Files D. Vassallo NRC POR D. Ross Local PDR Oregon Dept. of Energy ORB 1 Reading NRR Reading Mr. H. H. Phillips H. Denton Portland General Elec tric Company E. Case 121 S.W. Salmon Street V. Stello Portland, Oregon 97204 D. Eisenhut B. Grimes Warren Hastings, Esquire R. Vollmer Counsel for Portland General A. Schwencer Electric Company D. Ziemann 121 S.W. Salmon Street P. Check Portland, Oregon 97204 G. Lainas D. Davis Mr. J. L. Frewing, Manager B. Grimes Generation Licensing and Analysis T. Ippolito Portland General Electric Company R. Reid 121 S.W. Salmon Street V. Noonan Portland, Oregon 97204 G. Knighton D. Brinkman Columbia County Courthouse Project Manager Law Library, Circuit Court Room OELD St. Helens, Oregon 97501 OI&E (3)
C. Parrish Director, Oregon Department of Energy ACRS (16)
Labor and Industries Building, Room 111 NRC Participants Salem, Oregon 97310 J. Buchanan TERA Richard M. Sandvik, Esquire Licensee Counsel for Oregon Energy Facility L. Olshan Siting Counsel and Oregon Department D. Wigginton of Energy G. Zech 500 Pacific Building E. Reeves 520 S.W. Yamhill J. Angelo Portland, Oregon 97204 B. Buckley A. Dromerick Michael Malmrose C. Stable U. S. Nuclear Regulatory Commission R. Birkel Trojan Nuclear Plant D. Tibbitts P. O. Box 0 H. Silver Rainier, Oregon 97048 S. Miner S. Burwell J. Stolz R. Baer
- 0. Parr O'9 S. Varga 4 i fa 2
Attachrent 1 LIST OF ATTEf4 DEES TROJAN MEETING _
JUNE 8, 1979 NRC PGE C. Trannell D. Broehl J. Gray H. Schnidt M. Chi ranal L. Erickson F. Rosa R. Barkhurst
- 0. Chopra G. Nakayana P. Rich D. Tondi W
T. Dunning W. Sugnet R. Satterfield C. Faust D. Mcdonald V. Moore E. Butcher D. Lasher F. Orr 416 030
%dw1 &
PGE response to questions 1.3 b & c from NRC to Westinghouse
Reference:
W-NRC letter NS-TMA-2049, 5/14/79 Question I.3, b.
Identify those instruments inside Containment required to follow the course of Condition III and IV events.
Verify the capability of each instrument identified to perform this function and recommend a replacement instrument model for those not capable of long term monitoring.
Response
Instruments inside Containment installed for operational an/or safety function initiation that will provide post accident monitoring for Condition III and IV events are among those listed in FSAR tables 7.5.-1 and 7.5-2.
These are the pressurizer water level, steam generator level RCS (wide range) pressure transmitters, A-n d R cA RTDs.
R c4Moe coetaw4 % s b RT03 in yh (le.1 ese. Roseaca~+
f76 W 4
(RTO s Lave b<en +es +e d (ce e&
l'N K S.
T he se -
Ram to fsy, H b)
LocA ce4ba,
we WC Ap 9157 (320*- 2 20" e
Pressurizer water level transmitters (LT459, 460 & 461) installed are Foxboro Model E13Dil post LOCA environment rested per WCAP 8541 (286', 60 psig for 2 hrs).
Steam Generator water level transmitters (LT 517, 518, 519, 527,528, 529, 337, 538, 539, 547, 548, 549) installed are Barton Model 384.
These transmitters have been tested for LOCA conditions and beqc. m 4, de.t<n e,,J.
m s). 4 l3 4
over one m+e. (Franklin Institute report F-C-2623 dated 9/69).
In the event of an accident will perform their safety functinn ini ciarinn hur mau no t o ces s da.
+.n, C hast 1,n$ krm monde %g c.spdi h+y.
Reactor Coolant system pressure (wide range) transmitters (PT403,405) are Barton Model 389.
This model transmitter has not been tested for LOCA environment, Pod - aced 4d monitoring capability es unbom n, The steam generator water level transmitters and RCS pressure (wide range) transmitters are located at elevation 48' in the Containment or about 5' H low the maximum water level that would result from Condition Ilt
'" breaks.
It is estimated that for a 1" break submergence would occur in about 3 hrs, one hour for a 2" break and lesser times for larger breaks.
These instruments are pressure tight but submergence testing has not been conducted and the instruments mq ot p ova meadeeq c@ldy once submerged.
The above transmitters will be replaced with Earton Models 763 and 764 which will be qualified for longer term post accident environment monitoring capability, including submergence,under the Westinghouse Supplemental Qualification Program.
416 031
4
~
Question Westinghouse has indicated (Appendix A, of letter (NS-CE-1179)
I.3, c.
to J. F. Stolz of NRC from C. Eicheldinger of Westinghouse.
dated 8-26-76) that additional instrumentation located outside of Containment is available to the operator to follow the course of Condition III and IV events.
Identify the instrumentation and describe its capability for each specific event and plant.
Response
Outside of Containment instrumentation available to the operator and ith use is described in the attached table.
416 032 HPS/irf
TROJAN NUCLEAR PLANT POST-ACCIDENT BACKUP MONITORING CAPABILITY Normal Instrument Backup Instrument Use of Backup Which Mav Be Lost Outside Containment Instrument Przr Level 1.
DP cell in hot lab Read Locally - cay need to valve in.
2.
Comparison of przr If Twater is approaching Tsto, przr Tstm and przr may be emptying.
If Tsts approaches Twater Twater, przr is going solid.
3.
RCS pressure Rapid increase with decreasing behavior charging flow icplies going solid.
4 RCP ochavior If running, amps fluctuating or low indicates onset of two phase flow.
5.
Nuclear Increasing countrate indicates level instruments reaching core.
S/G Level 1.
DP cell in coli Read Locally - cay need to valve in lab for various S/G's (1 instrument for 4 S/G's) 2.
S/G pressure Periodically open atmospheric relief.
If S/G is nearly dry, it will rapidly depressurize. A ls o, if using Shm J u a.p s (to co &w-) bden b Ucw No1(
day v alve.A el % fir m s /(,
9,..s m,, i, shtak nae-no-6 L w L.4_
t loop T T4 g3 c RCS Pressure 1.
CCP discharge C12 and plant co=puter.
pressure 2.
CCP flow C12 and plant computer.
With CCP running, look at pump curves to estimate discharge pressure.
3.
S1 pump discharge C19.
Only accurate af ter flow starts.
pressure If no SI flow, this ensures RCS pressure greater than SI pump head.
4.
RHR discharge Cl2, C13, plant computer.
Only accurate after flow starts.
If no pressure flow, this ensures RCS pressure
- 5. % %n Indiators '
greater than PJIR pump head.4to o i,
% acm% vatu., a n. p H pe
.a r com s Ecs t a~e e liu s 9 5 ip awl
,m1 loc.6 m k.4 8 '-
l RCS Loop Tc's, S/G pressure Calculate RCS te=perature by use of Loop Th's, steam tables.
Incore T/C's
- Loq 7w,Tc ave infilm Tempecabces 56\\1 h sIch\\ 38d (ce drogpI9 egyd -
\\( coel6*,d* M u L tb t.T s M.
A Gdam n.d us a.* e N
416 0a; a4c.,, Jim.,asos, a,. A T - w u, u..., sls i..-t tc<r 7
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