ML19309G852
| ML19309G852 | |
| Person / Time | |
|---|---|
| Issue date: | 02/21/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19309G851 | List: |
| References | |
| ACRS-1707, NUDOCS 8005070684 | |
| Download: ML19309G852 (84) | |
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I ISSUE DATE: Feb. 21, 1980 pg '
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ACRS BABCOCK & WILCOX WATER REACTORS SUBCOMMITTEE MEETING Room 1046,1717 H Street, N.W.
Washington, D.C. 20555 January 8, 1980 The ACRS Babcock & Wilcox (B&W) Water Reactors Subcomittee met in Washingto January 8, 1980 to discuss the sensitivity to transients of once-through steam Notice of generators (OTSG) and other features of B&W designed nuclear plants.
this meeting, including a proposed agenda, was published in the Federal Register, Volume 44 No. 247 Friday, December 21, 1979 (Attachment A). A copy of the A list of attendees is Tentative Detailed Schedule is attached (Attachment B).
Copies of presentation slides and supporting documents included as Attachment C.
No written statements were received by the Subcom-are included as noted below.
mittee and no requests to make oral statements were received.
l EXECUTIVE SESSION The Chairman called the meeting to order at 8:30 A.M., and made an opening state-ment reviewing the purpose of the meeting and noting that Mr. Ragnwald Muller The Chairman felt there was no need for pre-was the Designated Federal Employ 6a.
liminary executive discussion and, noting that the Staff, B&W, and the concerned utilities were all present, with the concorrence of the remainder of the Sub-comittee, he elected to proceed with the, Staff presentations.
MEETING WITH NRC STAFF Dr. Thomas Novak opened the Staff presentation by introducing his colle Rubin and Norman Wagner.
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. Dr. Novak indicated that the 10 CFR 50.54 request (Attachment D) sent to the utilities by Harold Denton on October 25, 1979, asked for additional infor-mation to determine whether it was necessary to order a halt to all construction pending completion of a study of the risks of continuing, using Crystal River as the reference facility to be analyzed.
The Staff had selected ten minutes from the onset of a transient as the time during which no credit would be given for any operator action, either positive or negative. The Staff is doing analyses independent of those done by the licensees and vendor. Though the latter were asked to study sensitivities to operator error, the Staff has not covered this.
The Staff hopes to get an ACRS letter on this subject at the March meeting.
The state of construction is such that a delay until then would not matter.
It is not decided yet whether operation would be delayed or retrofit required.
Dr. Novak's presentation outline is Attachment E hereto.
Mr. Mark Rubin discussed the small steam generator sensitivity to secondary i
system changes using the slides in Attachment F.
Slide F-2 covered the ele-ments of system response. Slide F-3 covered recent operating experience, with trips experienced at C ystal River (6), ANO-1 (1), Oconee 1 (3), 2 (2), and 3 (1), and Rancho Seco (2). Slide F-4 listed the "directly attributed causes" as well as " fundamental causes".
Mr. Michelson observed that the Staff's schedule for those plants under con-struction (2-177 Units at Midland, and 205 Units each at Bellefonte and WPPSS) was tight since the ECCS work was not done. Dr. Novak observed that the schedule
may vary from plant to plant. Dr. Mark indicated that ACRS comment may well too, since there are plant differences (e.g. the 205's don't have a sparger at the top of the steam generator, and Midland, which has a sparger on top has a low steam generator).
Mr. Rubin noted that post-TMI actions nay have caused the larger number of overcooling events (See Slide F-2).
Norman Wagner discussed the B&W overfeed transient event sequence with Slide'G-1 and the analyses done by Brookhaven National Laboratory. The Staff and B&W disagree on whether flashing will occur in the upper head. With normal flow it takes about 100 seconds to fill the steam generator to the level of the steam line. Five variations of the base case were studied as indicated on Slides G-2 and G-3 (which shows Upper Head Quality vs, Time for the six cases studied).
As a result of preliminary analyses, no fuel damage and no violation of DNBR limits is anticipated.
I Mr. Wagner listed the modifications with potential to reduce plant sensitivity to feedwater supply changes in Slide G-5.
These include modifications to both the main and auxiliary feedwater supply systems.
Mr. Michelson observed that, with pumps off, the lowest pressure should be in the candy cane whereas the curves show the upper head flashing before the candy cane.
Dr. Novak agreed to determine why.
Mr. Ebersole observed that this was not the worst overcooling transient. Mr.
Wagner agreed but noted that there had been no cases of pressure regulator failure in ?aW plants, so, in an effort to be realistic this worse overcooling event was Sc: r.onsidered. Mr. Ebersole noted that there were four worse cooling
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J transients for an unanticipated trar.sient - (this one was anticipated).
Novak agreed to put analyses of those, such as steam line breaks, into context later.
The IRT Code was used by BNL whereas the applicants used the TRAPP Code.
In response to Mr. Ebersole's question, Mr. Wagner indicated that some plants,
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Midland, for example, did use bypass in their analyses.
Mr. Michelson observed that, by this analysis, before pressurizer level returns, the steam generator is already (at 100 seconds) filling the steam lines (which could lead to safety valves sticking open from water hamer effects.)
1 Dr. Novak stated t. hat the Consumers Power Co. submittal showed an SFAS signal isolates main feedwater and goes to emergency feedwater at 41.5 seconds which would prevent the overfill at 100 seconds.
Mr. Michelson was still concerned that operator action might be required prior to 100 seconds. Why was there no high level cut-off?
Dr. Catton noted that with the larger pressurizer (Curve 6) there was an increase in time, from 45 to 60 seconds, prior to voiding the pressurizer.
In response to Mr. Ebersole's question re the reliability of the ICS, Dr. Novak stated that a failure mode and effects analysis indicated that the IG worked fairly well. He indicated the Staff would brief the Comittee or. this.
Dr. Novak noted that no upper head flashing occurred in the applicant's analysis _
whereas it did in the Staff's analysis. The Staff took credit for RCP trip and SFAS initiation.
5-STATUS OF CONSTRUCTION Anthony Bournia reviewed the status of construction of the dual plants at WPPSS, Bellefonte, and Midland with Attachment H.
Only major equipment, not piping, was listed.
Foreseeable possible changes such as getting a larger pressurizer were noted.
Dr. Novak indicated that the review focussed on anticipated transients, that is those that were expected at least once in the lifetime of the plant. He stated that for a turbine trip followed by a reactor trip with continued main feedwater, there was no core uncovery and no fuel damage if the operator takes no corrective action before ten minutes.
If the period were any shorter than ten minutes in which action had to be taken the Staff indicated they might resort to a dedicated operator for just that function, such as the required dedicated operator to initiate AFW in CE plants.
This, however, in an interim solution, because the constant alertness of a 1
dedicated operator cannot be guaranteed and for the longer term a system change obviating the need would be preferred.
Mr. Michelson wondered if the operator could be trusted to shut off main feedwater if it were overfilling the steam generator.
Dr. Novak pointed out that turning on an AFW pump cannot get the operator inte any serious trouble, such as, fer example, turning on the poison system in a. OWR which is a " point of no return."
He indicated that if the action had to be taken within two minutes for safety
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reasons, it would have to be automated.
If the operator knows he has in over-filling event, it is relatively simple to terminate it.
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Mr. Michelson pointed out that one can't know where the water level is in a OTSG steaming at fairly high rates. Dr. Novak admitted that the Staff was still studying the problem and had made no decisions on needed design changes.
Mr. Ebersole called the Staff's attention to the provisions of ANSI Standards 4.1-1978 and N660 (Draft). The latter covers criteria for safety-related 1
operator actions (copies of these standards have been distributed to the ACRS).
Mr. Ebersole also stated that this study should include what the operator can do incorrectly, that will have to be corrected later by him or negated auto-matically.
Dr. Novak used Attachment I to show the immediate actions the operator must take following reactor trip. These actions were gleaned from the Rancho Seco emer-gency procedures.
Mr. Ebersole pointed out that the procedure, when it says " verify", continues with the assumption that the condition is verified but makes no provision for what happens if it can't be verified.
Mr. Michelson pointed out that if the main steam isolation valves close and electric driven feedwater pumps are not used, feedwater injection is auto-matica11y terminated, at that point.
The Staff does not review non-safety grade systems and therefore no credit is i
given for their presence. Dr. Lipinski pointed out that while this may be unrec'-itically conservative in some instances, it is quite possible that pneuma!!c systems that have only a single source can be flooded and all fail e,
simultaneously. These systems do not necessarily have the redundancy and separation of safety-grade systems.
Dr. Novak reminded the Subcomittee that the traditional criteria for eval-uating anticipated transients are that there is no local overheating, no core uncovery and no fuel damage, and the reactor coolant system pressure should not exceed 110% of design pressure.
IREP STUDY - CRYSTAL RIVER 3. - Joseph Murphy (PAS)
Studies subsequent to the TMI accident demonstrated that core damage was more orobable than WASH-1400 indicated because of sequences involving equipment and human interactions.
Plant-to-plant differences seemed more important following post-TMI studies (AttachmentJ).
Mr. Murphy discussed the IREP Study which involves two NRC people, about 15 people altogether for an estimated four month effort.
l Preliminary indications are that the cause of the TMI accident is a combination of AFW system design and operator action - the intertwining of the closed cooling water systems and human error. Timing for corrective action is 2-4 times shorter for B&W systems than for others.
There is a variation of three orders of magnitude in the reliability of auxiliary feedwater systems between different plants. Susceptibility to ATWS varies widely. This discovery led to the Integrated Reliability Evaluation Program (IREP)
(See Attachment J). Crystal River 3 is the subject for a pilot study.
This is to address the B&W sensitivity issue. LER's and AOR's are being ex-f aminej for precursers and system interactions. Event trees and fault trees are l
being developed. Preliminary findings and reccanendations are expected by March.
Detailed analyses will follow.
Mr. Murphy predicted that a significant finding may be that one of the largest risk contributors will come as a result of the common dependency of two closed cycle cooling water systems, the nuclear service water closed cycle cooling system and the decay heat closed cycle cooling water system, which provide cooling to a number of ESFs.
Balance of plant rather than NSSS, considerations will probably dominate risk and core melt probabilities. This is based on very preliminary analysis.
The Study will look at compounding human errors.
The prime contractors are Sandia Labs. and INEL with SAI a principal subcontractor along with Energy Inc. and Evaluation Associates.
Another preliminary result reported by Mr. Murphy was that the ICS will not have a significant effect on the overall results.
Mr. Etherington noted that the B&W system capability to vent at the top of the
" candy cane" was a plus.
Dr. Novak pointed out that the obfectives of the work included reducing the potential for voiding and mitigating the consequences of voiding. The latter is not a direct part of the analysis.
UTILITIES' PRESENTATIONS Mr. Hosler of WPPSS acted as spokesman for the three utilities who hoped for near-term OL's for B&W plants (See Attachment K).
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l The principal requests of Denton's Oct. 25,10 CFR 50.54 letter are shown on Slide K-2 (attached) and the utilities agenda is shown on Slide K-3 (attached).
The three utilities worked together on identifying the most severe overcooling events, anticipated transients and accidents, and detemining'of ECCS or RPS action was necessary for those transients. Consumers' Midland plant being of older design differed somewhat from the WPPSS' and TVA's submittal, Scheduling of completion of installations was, of course, done separately by the utilities.
The coments or OTSG sensitivity were generated with the help of B&W.
Mr. Hosler indicated that the utilities felt that the twenty-six Pebble Springs questions which they had been requested to address by the Subcomittee were related mostly to unique Pebble Springs balance-of-plant features and they felt they were not appropriate to address here. Nevertheless, he indicated a willing-ness to try to answer any Subcomittee questions.
OVERC00 LING EVENTS - Daniel LaBelle (B&W)
The Staff had requested additional infomation (in the 10 CFR 50.54 letter) concerning non-LOCA overcooling events.
Mr. LaBelle used the Slides in Attachment L hereto.
Mr. LaBelle confirmed for Mr. Michelson that whenever HPI is actuated, main steam feedwater valves are automatically closed on all plants. Then subse-quent to ESFAS actuations, the plants are designed with a FOGG (Feed Only Good Generator) logic.
Mr. LaBelle confirmed for Mr. Michelson that no plants or plait designs cur-rently have a high-level cut-off of main fee.dwater.
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The two transients analyzed so far were:
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Main FW overfeed following a reactor and turbine trip (with different assumptions'than used by the Staff).
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A single failure, double ended steam line break of a main steam line relief valve on the steam generator side.
All main FW pumps are turbine driven in B&W plants. Mr. LaBelle admitted to Mr. Michelson that no study had bera made of stuck open safety valves pre-cipitated by an overfilled steam generator. Mr. LaBelle thought this was not much of a problem.
Dr. Catton observed that the model used probably had too few nodes for accurate modeling.
Mr. LaBelle referred the Subcommittee to B&W 1Q128 for heat transfer data in the TRAP-2 code.
Dr. Catton felt the heat transfer calculations were inadequate.
If the AFW in-jection point is above the level in the steam generator, the heat transfer will be high. He felt this was not adequately taken into account.
Mr. LaBelle reviewed the curves in Attachment L.
For the overfeed case in the Midland plants, the reactor and turbine are tripped and subsequently the steam generators are overfed with main feedwater reducing the RCS temperatures and pressures.
In the 200 second range, the ESFAS reaches its low aux feed pressure set point, isolating the main feedwater. This will not actuate the AFW until the inventory in the generators has boiled back down to a low level, bringing the transient effectively under control.
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Mr. Ebersole pointed out that there was no redundant main feed isolation valve and failure to isolate could overpressure the containment.
Mr. LaBelle indicated this was with the RCPs running. Lack of time pre-vented examining the pump trip case. He pointed out tripping the pumps reduces the overcooling break and improves the pumping results. Dr. Catton was surprised that the analysis did not include pump trip at the plant pump trip conditions. Mr. LaBelle pointed out that from the outset B&W did not assume the RCPs would be tripped for overcooling events on low RC pressure.
The pump trip feature on plants under construction is limited to LOCA con-ditions. He explained to Dr. Catton that an attempt would always be made to distinguish between overcooling and a small break.
Dr. Womack (B&W) pointed out that keeping the RCPs running would result in the greatest cooldown and would therefore represent the most conservative case.
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In response to Mr. Ebersole's question, Mr. Kane (B&W) stated that only Davis Besse lacks high head HPI pumps and must rely on AFW. Bleed and feed repre-sents an alternate mode of emergency core cooling ic case adequate cooling on the secondary side is unobtainable. Mr. Ebersole pointed out that this in effect would create a LOCA because fluid would go to the pump and be recircu-lated through the residual heat removal system as in a LOCA. One could not reach cold shutdown with only safety grade equipment for a long time.
All B&W plants have only one PORV and two code safety valves.
Mr. LaBelle noted that because of the larger hot legs in the 205 plants they have a higher degree of voiding than the 177 plants.
- 12 e Mr. Michelson pointed out that the 205 plants were not adequately modeled.
The model covers bottom flooding but not spraying the upper portion of the tubes. Dr. Womack (B&W) indicated that perhaps the model underestimated the heat removal in the 177 case.
He would check it.
Dr. Womack showed a diagram of a OTSG (Attachment M) and reviewed the heat transfer process. The generator is intentionally designed to be sensitive to feedwater flow rate and level changes, and, like all steam generators, is sensitive to variations in secondary side pressure. Heat balance control is achieved via the ICS in power operation. The possibility exists, due to abnormal equipment operation, to over or undercool the system to trip.
Cer-tain post-TMI changes in operation may have increased the likelihood of this occurrence. Dr. Womack used slide M-4 to show post-TMI changes affecting plant response.
Dr. Womack discussed the defense-in-depth approach to maintain adequate core cooling in overcooling events (Slide M-3).
j All B&W plants under construction will have safety grade AFW systems, with im-proved reliability and control.
B&W feels that anticipatory reactor trip should be eliminated for disturbances such as turbine trip.
Early termination of core heat generation might lead to a larger transient. The plant control system can handle these events with-out relying on the reactor protection system.
Raising the PORV set point and lowering of the high pressure set point has increased the number of reactor trips in the B&W operating plants. B&W would prepose to restore the previous settings.
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A range of small breaks was discovered somewhat larger than TNI-2 and smaller 2han half a square foot which could result in core conditions that violate 10 CFR 50 Appendix K.
Operating instructions were issued to trip RCPs under these conditions. B&W has proposed to its licensees that this be automated, based on the sensitivity of the RCPs themselves to point conditions in their pump chambers, and the reduction of the RCP motor current which results from void formation in the pump's system.
Mr. Michelson asked if the motor current logic, would be auctioneered between the four pumps. Dr. Womack indicated that that design consideration had not yet been decided.
Mr. Michelson pointed out the desirability of not tripping the pumps unless it is really necessary. He asked if tests would be run. B&W had not yet identi-fied a need for a test.
Mr. Michelson expressed concern that it would be quite some time before this motor current trip method could be developed.
Dr. Womack pointed out that the pump trip could be set so that lower void fractions would not trip the rumps but higher void fractions (-20%) would.
If there is no small break the pumps would be kept running. One of B&W's analytical studies is currently determining how the void fraction changes around the loop. The relationship between pump current and void fraction will be charted. Voids must form because the water is densified.
Mr. Michelson pointed out that pumps which are subjected to pumping liquids with voids operate erratically not smoothly and therefore pump motor current l
may also vary erratically. Tests,f the feasibility of the system will cer-l tainly be required.
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- 14 Dr. Womack reminded the Subcommittee that loss of indicated pressurizer level on the B&W plants is not synonymous with loss of liquid in the pressurizer.
A typical case, Davis-Besse, has more than 40 inches of capacity below the zero indication.
B&W plans to install automatic equipment to eliminate RCP trip associated with low reactor coolant pressure only, in the plants under construction.
Main feedwater limiting equipment will also be added to prevent excessive filling.
The AFW system and its circuitry are being studied for upgrading to minimize the excessive addition of cold water possibly leading to emptying of the pres-surizer.
Mr. Michelson asked about pressurizer heater failure modes. Dr. Womack replied that if they fail they will burn out but the pressure boundary will not be vio-The heaters burn out before the sheath and in doing so they turn them-lated.
selves off.
Mr. Ebersole asked about the vulnerability of Davis _Besse with its two turbine driven AFW pumps. Mr. Dunn pointed out he felt that feed and bleed could be used at Davis-Besse and a slight amount of core unctvery would not be serious.
Dr. Novak indicated that a third pump of diverse power sources had been recom-mended by the Staff. With proper operator action the Staff had determined the Davis 4 esse core could be cooled without any feedwater.
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B&W indicated the operators had been trained on these procedures.
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Davis-Besse has a raised loop configuration which makes available a larger than normal inventory of primary water for drain back into the reactor vessel.
TVA PRESENTATION - Mark Linn Mr. Linn stated TVA's position that construction should not be halted. He cited the improvements made in 205 plants over the older 177s.
Dr. Novak asked B&W and WPPSS why the auxiliary feedwater injection point was shifted from the top to the bottom of the steam generator., B&W cited two reasons:
(1) The elevated loop did not require high injection for natural circulation.
(2) High injection caused difficulty in getting the ring indicator on top of the superheat support rack.
CONSUMERS POWER PRESENTATION - Mike Salerno Mr. Salerno indicated that any necessary modifications could be made during the construction phase at Midland. He cited the Class IE AFW system initiated on ESFAS, and safety-grade pressurizer heaters. He proposed to add a safety-grade trip on loss of feedwater but to keep the original PORV and high RCS Pressure Trip set points. Also reliable safety grade PORV position indication will be pro-vided. Also dual block valves will be installed. A test program will be initi-ated. Consumers is working on better methods to identify transients. Mr.
Salerno referred to the Consumers response to the Denton 50.54 letter.
WPPSS P4ESENTATION - A. Hosler Mr. Hosler reviewed the WPPSS submission (Dec. 3,1979 letter Serial G01-79-580, Dockets Nos: 50-460,50-513) with the aid of slides in Attachment N.
Dr. Novak thanked the Committee and indicated he had learned a great deal which the Staff will have to study before the next Subconnittee meeting.
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noted that Bellefonte and WPPSS were not on quite as innediate a schedule as Midland and it may be desirable to handle them separately.
There were no further questions. The meeting was adjourned at 4:20.
NOTE: A complete transcript of the meeting is on file at the NRC Public Document Room at 1717 H Street, N.W., Washington, D.C. or can be obtained from ACE Federal Reporters, Inc., 415 Second Street, N.E.
Washington, D.C.
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Federal Regleter / Vol 44, No. 247 / Friday. December E1.1979 / Notices 75f57 1
-.,wediender,femmery0 2ses asts.m.
At the conclusion of the Executive-1
'%s agenda for subject
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.be se follows:
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.n. e hea==itterwemeetin closed P"stadone by and hold escassione- ' y,,g,y,7,,,,,7 7,,,,,
., seesloos to hold discussions with d & sit Undflee Conclusloe ofJosinese y(
repneestatives of the NRC StaE and maadtmu udobrkkrwhd
' ' %e hhea==lthe may med la i
their consultana, mgareng items persons.
Executive Session, with any ofite :-
Portinen to & A{RS mview.
In addition,it may be necessary for consultants who may be present, to In denL
= use g b Subcommittee to hold one or more r & lore and -d== their preliminary mt in cland Faecudn Sess!m 2 cloud sessions for the purpose of op.nons regarding inatters which should
""haat" &*ir pm o
na agardas papara
, report exploring matters involving proprietary be considered during the meeting, be submined to &e ACRS.
information. I have detar *=A_ in At the conclusion of the Executive Ihan ed. h accadance with accordance with Subsection 10(d) of the Sesson, the Subcommittee will hear Federal Advisory Committee Act (Puh.
presentations by and hold 'H=@-
Sob gg gg g p, Advhwy Comh Act (Puk L 93 -
L 93-463), that, should such sessions be with tatives of &e NRC Staff.. -
4eaI' that it is nec""#Y to close these required. itis - y to close these the nu earladustry,variouantilities.
sessions to protect proprietary and their consultants, and other
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al t information. See 5 U.S.C. 352b(c)(4).
laterested persons.
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'I "IEcantiY hustrate hePlesentation of Farewinformados agardag 2 pics In adadon. it may be necouary k '"
8 to be discussed, whether the mee
- PNPond sedom Sm 5 U.S.C.552b(c)(4 has been canceHed or roscheduled..
. the Subcommittee to hold one or mese " ' ' 7 closed sessions for the purpose of M "
.. ' Deasd: December 17,lors' Chairman's ruling on regnests for the
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exploring matters lavolving proprietary 'W i
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opportunity to t oral stata=an'=
Informados.Ihave determined,in
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and the time ' otted therefor can be ~
"accordance with Subsecdos to(d) of b -
pa n.oe.asu rew u.wm ees obtafned by a prepaid alephone caH to Federal Advisory Commdttee Act(Puh. #
the cogntsant Designated Federal L 92-483), that, should such sessicas be
su.sma enes reew+e..
Employee.Mr.Ragnwald Muller
-required. It is nar==== y to cices thess L
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(telephone 202/834-1413) between 3:15 sessions to protecdonr.,W iy
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Addeory Committu on Reactof r.m. and 520 p.m EST.
Infonnadon. Sm 5 U.S.C. 552b(c)(4). '
Wuco W De emeber 17.1sm.
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-.. The ACRS Subcommittee on Babcock W Cimm/ase Mangeaient ogem.
has been canceDed or r=d-M-d e
l Chairman's ru!!ng on requests for ths "Tand Wilcox Water Reactors will hold a
"""d"'******l opportvedty to present oral statements 7 2 t meeting on January 8.1980 la Room and b time allotted therefcr can be a-104e.1717 H St. NW.,Weehington. DC obtained by a prepaid telephone caH to-20555 to discuss the sensitivity to Addeory Committee on Reactor b m- !==? Designated Federal transients of once-through steam Sefeguertfe: Ad Hoc Sutscommittee ort Employes. Mr. Richard K. Major ponerators (CTIllC) and other featume of Three afue leiend. Unet 2 Accident (telephone 202/s34-1414) between 4.15-Esbcock and W11ccx designed nuclear Acdon Plan;neeeting a.m. and Scop EST.
t plants. Notice of this meeting was pubtlehed n.c h.,20.1979.
De Acts Ad Hoc Subca==1ttee on
=d.,
informados concerning
.!a accordance with b procedures the Three Mile Island. Unit 2 Accident ihas to k decaud at eis madng -
can b food in hk on Bk d ['..
D ' ~ outlined in the Federal Registar on Action Plan win hold a meedag on
. October 1.19:11 (44 FR,58400), oral or
}anuary 7.19e0 in Roosa 104a,1717 H St*
available for public laspectico at the.
i wrttten statements may be presented by NW.Washingtor. DC 20555 to discuss NRC Public Docurnent Room.171y H members of the blic, recordings wiu the NRC " Draft Action Plans for Street. N.W., Washington. DC 20585 and be tted during those portions haplementing Recommendadoes of the et b ht Pubucadas Saden, ' '
State !.ibrary of
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el meeting w a transcript is being Pmeident's Cemndselon and Other e.
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kept, and questions may be asked only Studies of the Three Mile Island. Unit 2 by members of the thea==ittee. Its Accident." NUREC 0880. December 10
. Walnut Street. Harrisburg, PA 17125.
esasultants, and StaE Persons desiring 19"s. Notice of this meeting was Deted: December 17.tsrs.
W to make oral statements should notify Published December 20,1979.
leha C.Heyte, the Designated Federal Employee as far in accordance with the procedures AhmaryComentasse Aa r ---*opasr.
la aemacs as practiceble so that outlined in the Federal Register on.
pensowsms red m.m.en meaus
< appenciate arrangements can be made October L 1879 (44 FR 56408). oral or saa.sse caos rien.ei.e
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. to allow the - aaary.thne during b written statements may be presented by
- l meeting for such statements.
members of the public, recordings wtR 1
. The agenda for subject mating shan be permitted only during those portions CFFICR OF PUTS 00 GEL -
be as follows:
of the meetiig when a transaint is being RAANAGEt8DfT
. Leeder./oamary & JN ass, kept, and questions may be asked only flatilthe Coac/usson ofBusmese by members of the Subcommittee,its Adoption of OfHolei Seal try the OfRoe -
consultants. and StaE Persons desirtag of PersonnelIsanagement The Subcommittee may meet in to make oral statements should notify asesev:CSco of Personnel Executive Session, wita any ofits the Designated Federal Employee as far Management.
consultants who may be present, to la advance as practicable so that
, srplore and exchange theft preH=1aary appropriate arrangements can be made 4,.rieme Nodos.
opuuona regarding matters which should to allow the necessary time during the aussssaarvt his docmaent, under the he raadA= red dunas the meeting.
meeting for such stat==aa'n-auther:ry of Reorganisanon Plan 2. > -
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12-27-79 3
TENTATIVE Dt?IRIED SCHEDU2 m...
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ACRS BABCOCK & WIICOX WATER REACICRS SUBCCMMITTEE
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s-yy ICCM 1046, 1717 H. Street, N. W.
Washington, D. C. 70555 January 8, 1980 Times are Approximate 8:30 A.M. I. EXECITTIVE SESSION (OPEN)
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Opening Statement - Mr. H. Etherington, 91 % fttee.
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A.
Chairman B.
Discussion of Agenda and Objectives of Meeting
. g,,..
u.....
C.
Preliminary Opinions of Consultants 7,
p,
II. MErrING WPIH NRC STAFF - (T. Novak, S. Varga, J.MJrphy#.,e
~' ~~~
,y
~~~
,y) 9:00 A.M.
A.
BACKGROUND
- t.. $.
(1) Recent B&W Operating Experience 7
g.,.3
~~
~"~
~
(2) 10 CFR 50.54 Request Regarding Design Adequacy of 1.s.
B&W Nuclear Steam Systems (NSS)
{~_. _
'~
10:00 A.M.
B.
DISCUSSION
~.. :.,...,--~~
(1) Staff Review of Plant Construction Status and
...e.
~ " ' - ~
Suggested Halt in Construction
- a'g e u- '.~e
' :.z'.
s
'..1'. :s.
(2) Review of overcooling Transient Analysis
..- r. e.
u
. :.1 c e:
(3) Suggested Hardware and Procedural Changes g
_y
.n (4) Integrated Reliability Evaluation Program (IREP),
~ ' ' ' '
3
- -u :
>::as-. i:-
(a) Lessons Imarned from 'IMI-2 Accident for 5.::? "".u-Nuclear Risk Assessment (b) IREP
..y.
(c) Pilot Study - (Crystal River 3)
(d) Results to date
---,--1 (5) Iogic for assumi.vg:
(a) that an anticipated transient _ (as contrasted with
, l 3-an accident) be handled with no operator action. '-
(b) that only safety systems be included in their analysis.
- C.
)
i 1
=-
January 8 Subcensaittee
. :.nc.s.,. ! s.:xe-: ::e i
11:30 A.M.
C.
C WCWSIONS Staff Position WNCHECN BREAK 12:00 - 1:00
- L
' :v:- M '__ I.-
MEETING WI'D1 VENDOR AND CCNS'IRtXTION PERMIT HOI.DERS n_......,..
III.
1:00 P.M.
A.
DrrRODUCTION A-l
\\
. ~ ~ ~.. - -
P:
-:.u.
Mr. A. Hosler, %@PSS
~r.r.J.t--
L:. ? '.. c. :.
Surtnary of Overcooling Acident and Transient Analysis 1:;. F..
E.
"s r c :f..
1:10 P.M.
B.
Mr. D. W. LaBelle, B&W
".:*A. #1
+.3 ; ;: : -- :: Tu.:
Chee '!hrough Steam Generator (O'ISG) Sensitivity to Feed
~
1;43 F '-
?M~~~~~
1:45 P.M.
C.
Water Transients
- 12 -
~'
Mr. E. A. temack, i.'W w:. 5.. F....:-
Design Changes smder Consideration D.
r--
1 2:25 P.M.
(1) Constzners Power - Mr. N. J. Salerno 2:45 P.M.
(2) WPPSS - Mr. A. Hosler l
/
3:05 P.M.
(3) TVA - Mr. M. A. Einn O'
'" i - M:
3:3 7.'.
m;-
re a B&W :'.1 Completion of Answers to ACRS questions re a B&W plant 9: : - 5: 3125r P'.M.'. f 1
- E.
raised during 1977 Review of ths. Pebble Sprirygs Nuclsarr'.a-::-
rnsei d.:; :
Plant.
F.c..
4:30 EXmMIVE SESSICN (OPEN) n......_
43 Discussion of Su W ittee Report D: 5.:.:r
- f L:>:3- -
1 I
5:00 P.M.
An:CUPN m e gm. -M.
O I
I I
4 enha
.e. e.
e e,.uu.
1
- :. :a..
ATTACHMENT - C ACRS BABCOCK & WILCOX WATER REACT 0ftS SUBC0tHITTEE MEETING Room 1046, 1717 H Street, N.W.
~.._. ~
Washington, D.C. 20555 January 8,1980 ATTENDEES LIST ACR'S Members ACRS Consultants WA 7
H'.: Ethe~ ring'tb'n, Chainnan C. Michelson
......... John Bau]j.s, ton E M.. Mathis...
W. C. Lipinski E,.. : John F. Cox Dennis C. Terrill 4; J. Ray.
I. Catton J. C, Ebersetle
- ,... Mark Linn D'. W. Wil son
-~ ;.,,.
L. A. Haack
.m Consumers Power ACRS-Staff '
NRC T. J.,Sullfyan...
l R. Hu11~e'r. Sr. Staff Engr. DFE T. M. Novak E. Abbot, ACRS Fellow R. M. Berners
. D. T. Perry-W. Kast'enberg, ACRS Fellow J. A. Murphy
- .....
- -M. J;.Salerno
~
J. Stampelos, ACRS Fellow Mark P. Rubin
.): _. !.
. '.. :.?.'
- , ?
~;...
G. Young, ACRS Fellow Anthony Bournia
...~;
Darl Hood Don Kirkpatrick Nonnan Wagner ACE-Federal Reporters, Inc.
WPPSS_,,
....s...:....
Gene Waldkoetter Jine: Beach i:_!!.
A6h,Riley..'
O?.U.
Newell S. Porter David Hoffinan 3
....._, Alan Hosler BABCOCK & WILCOX OTHERS E. A. Womack A. A. Garcia, sal R. W. Donnan R. E. Scheffsta11, XMC, Inc'..'
- 8. Dunn Andrew J. Rushnok, Ohio Edison J. D. Carlton W. B. Rodill, VEPC0 -
Roland L. Reed Bob Leyse EPRI
., m.
Ron Finnin Don Martin, NYWT Richard R. Steinke Don R. Swanson, PGE Co...
Edward R. Kane M. A. Spuds, Su111vam & Cromwell Danny W. LaBelle W. A. Coffman, GAO Consultant Carl Connell Daniel H. Williams, Ark. P&L Robert E. Ltghtle K. C. Fortine, LWRAT Jennifer Small, Small Newspapers p
o e
[
%g umTED STATES N..,.
,...-)
- x '.L' ' %
NUCLEAR REGULATORY COMMISSION
-j -
j wAsmNGTON. D. C. 20666
[
S Octcber 25, 1.979 Docket No.: 50-460 and 50-513 Pr. N. O. Strand Managing Director n-:-
+:
Washington Public Powr Supply System
.u--
F. O. Box 968 c
3000 George Washington Way Richland, Washington 99352
Dear Mr. Strand:
3-i
~
.. SUBdECT: 10 CFR 50.54 REQUEST REGARDING THE DESIGN ADEQUACY OF BABCOCK ~
& WILCOX NUCLEAR STEAM SUPPLY SYSTEMS UTILIZING ONCE THROUGH
~
I
~
~
STEAM GENERATORS (WPPSS 1 & 4) l Several. ha'rdware and procedural c'hanges have been made to ' operating B&W plants.
to reduce the likelihood of recurrence of a THI-type accident...These changes
.I ihave been in the area of auxiliary feedwater systems, integrated control.
, system, reactor protection system, small-break loss-of-coolant accident analysi's and operator training and procedures.
However,. at this time, we
{
a're beginning to look more deeply into additional desig'n features of B&W
'[lj
' ~;-
d' plants to consider if any further system modifications [arle.hecessaryy ;...
p
,,'The'i:se"of once-throt.gh-steam-generators (OTSG) in B&W plants, has an opera,.. :....
,' tional advantage in that it provides a small degree of steam superheat,
- as contrasted with the conventional saturated U-tube steam. generator.
"1n addition, it provides for less water inventory thus in~aking a' steam
~ i line ' break less severe. However, the relatively low water! inventory ~ with;_- ~
~~
~
~
',;,7,, th,e' ' presence of a liquid-vapor heat transfer interface in th'e. a'ctive heat
.J
, transfer zone closely couples the primary system to the steam. generator.
.; "._,,'._ conditions with a consequently high sensitivity to feedwater-flow rate
.. -, perturbations. Enclosure 1 to this letter addresses sy' stein problems and..
" ~ staff concerns in this area. At present, we are investigating whether 1 B&W plants are overlysensitive to feedwater transients,idue.to.the OTSG.
' concebt, as coupled with the pressurizer si:ing, ICS design, and PCRV/ reactor trip set points.
~ As part of the post TMI-2 effort, detailed analyses have been made of under-
, cooling transients for B&W plants. However, due to the sensitivity of the
~ 0TSG design, B&W plants have also been experiencirig a number of relatively severe overcooling events.
l
Mr. N. O. Strand.
For your infomation, NRC is initiating a research task to quantitatively assess B&W system designs, including the integrated control system, aimed at identifying obvious accident sequences leading to core damage having a high frequency as compared to the Reactor Safety Study, see Enclosure 2.
( A complete detemination of risk will not be attempted). The objective of this assessment is to identify high-risk accident sequences (including
,E [TMI implications) utilizing event tree and simplified fault tree analyses,
'. :, Included will be estimation of release categories, approximate quantif.1 c
, ~
~...r
- i..- - -
- studies for reliability of operator response. The study wilj focus.on g the risk implications of the sensitivity of the B&W des.ign and on the.
,..,..3 potential interactions arising from the integrated cont' ol,systen.: We:
~..
r
", estimate this study to be completed in about six monthsy ' We wil.1. us'e..
4
~~
the Crystal River, Unit 3 plant as the referenced facility to be analyzed.
Ne 'have been holding generic discussions with Babcock and Wilecx.Canpany.
concerning this matter. Howver, system sensitivity to feedwater transients
..i~nvolve's balance-of-plant equipment and systems as well'asithe nuclear steam -
t
~ n supply system, and such plant-specific characteristics pus, be'co' side, red.:.
~
~
We 'are also considering whether it is necessary to halt.portjons-of the..
construction of B&W plants, pending the outcome of the reliability assess -
~~
_ ment. As a preliminary consideration, we have identified those systems and components that may be impacted by possible design changes as.a. result of this study. Enclosure 3 is a preliminary listing of such systems and.
. 7 components.
11nd'er the authority of Section 182 of the Atomic Energy Act.of.]954, as..
m..
. amended, and Section 50.54(f) of 10 CFR Part 50, additi' n'aL.inferination-
'. ?'.
o
'is' requested to allow us to determine whether it is necessa'ry'to' halt all
. ~~ ..
~
or portions of the construction of your plant pending th'e results: of:our. '
.r.
study. We request you provide:
- j.,.
f ' ~ ~
[.ja')) den *.ify the most severe overcooling events (considering both. anti.
cipa ad transients and accidents) which could occur'at your. fac.flity..
~;.
Thev should be the events which causes the greatest inventory
.. shrinkage. Under the guidelines that no operator action occurs.
before 10 minutes, and only safety systems can be used to. mitigate
"-,:'the event, each licensee should show that the core remains adequately.
~.
~
cool ed.
~ b) Identify whether action of tne ECCS or RPS (or operator act. ion)
~
is necessary to protect the core following the most severe over-cooling transient identified. If these systems are required, you should show that its design criterion for the number of actuation cycles is adecuate, considering arrival rates for excessive ecoling transtents.
t
c)' Provide a schedule of completion of installation of the identified systems and components.
d) Identify the feasibility of halting installation of these systems and
,,,,,, components as compared to the feasibility of completing; installation and then effecting significant changes in these systems.and-components.-
~
'i
e) Comment on the OTSG sensitivity to feedwater transients.
- .3,
,'l'2', '! f)M P'rovTde recommendations on hardware and procedural changes related to.the : g :-
need for and methods for damping primary system sensitivity to 1,:-
l, perturbations in the OTSG. Include details on any design-adequacy.
studies you have done or have in progress.
We are s'ending similar letters to all utilities holding. construction permits
~
for plants with B&W nuclear steam supply systems.
-u-
' We' resest your reply by December 3,1979. We believe that.a meeting with
.you' and the other utilities together with the staff and the. Babcock and
'Wilcox Company to discuss this matter would be beneficial'to all parties.-
1
~
' At that time, we will provide further details on the Crysta1 River. Study.'
[-
- [, '-,
i 3
We ' re scheduling such a meeting for November 6,1979 at 10:00 a.m. in.
a g-Room P-422 at our offices in Rethesda, 7920 Norfolk Avenue, Bethesda, Maryland.
~
~
Plea'se call Dr. Anthony Bournia at (301) 492-7200 if you have any questions -
~
concerning this letter.
Sincerely, f
Harold R. Denten, Director 3-Office of Nuclear Reactor Regulation e-
Enclosures:
1: -
As stated cc: See next page
Mr. N. O. Strand
~cc:
Mr. B. C. Redd Jerome E. Sharfman
~-
~
~
United Engineers A Constructors, Inc.
Atomic Safety and 30 South 17th 5treet Licensing Appeal Board Philadelphia, Pennsly1vania 19101
.. S. Nuclear Regulatory Commission Nicholas S. Reynolds, Esq.
Washington, 0. C.
20555 DeBevoise & Liberman 1200 Seventeenth St., N. W.
Washington, D. C.
20036 Mr. E. G. Ward Senior Project Manager
.a-Babcock & Wilcox Company P. O. Box 1260 Lynchburg, Virginia 23505 Robert Lazo, Esq., Chainnan Atomic Safety and Licensing Board V. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. Donald P. deSylva Associate Professor of Marine Science Rosenthiel School of Marine and
- --- +
Atmospheric Science University of Miami Miami, Florida 33149 Dr. Marvin M. Mann Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 2
Richard S. Salaman, Chairman 3:,
Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission.
Washington, D. C.
20555 Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. 5. Nuclear Regulatory Commission Washington, D. C.
20555
ENCLOSURE 1 u.a.....
' Tr'imary System Perturbations Induced by Once-Through Steam. Generator:,
.,.g....
e.
I.
Introduction-
~ id 'p'lants employ a once through st. sa generator (OTSG) design, r,ather,...,....,.
B than 'U-tube st eam generators which are used in other pressurized water Each steam generator has approximately 15,000 vertical. straight reactors.
o 603 608 F and exiting,.
' ' tubes, with the primary coolant entering the top at 7
g
~.,.
..th~e' bottom at about 555 7.
Primary coolant flows down.inside.the steam i
. i...
" gen' era'tortubes,whilethesecondarycoolantflowsup.from.tbe ottom on.
3 th'eshell side of the OTSG. The secondary coolant turns to, steam.about.,.,....
' " hair'wa'y L'p, with the remaining length of the steam generator being used.to.,_,...
~.
superheat the steam.
3,.,..,.
i I
'"The se"cohd'afy-side heat trensfer coefficient, in the stea7 sp4pe.cf.the OT5G..,.....
.u.
i's much l'e'ss than that in the bottom liquid section. Tnis.results.in.a. heat....,
' transfer' ~ rate from the primarysysten which is quite g ensitive to.the. liquid..._,.
'Tevel in' the steam generators. If a feedwater increasg event occurs, _ the......
' ' i'
....v.,.
~ liquid-vapor interface rises, increasing the overall hes.. pransfer,..This 3
..,g
' deFr'eas~es the outlet temperature below 555'F and initjfpgg, ap.oyercooling.,,_,,,,,
....,.,1 eveh~ti which can lead to primary system depressurization., By. contrast.,,...
~~
~~
if'a feedwater decrease event occurs, the'overall heat. tra.nsfer decreases,
'the'o0tlet primary temperature increases, and a pressyrjzation. transient.,,
ensues.
In'aither of these cases, the response of tne primary. system, pressure and
~"
~
pressdrizer level to a change in main feed ater flow r.a,te (or temperature)
~ ~
'is com:arktively rapid. These rapid primary system p. essure chances due.to r
cr.anges ik feedwater conditions is knene Merein as syster " sensitivity" and f s
~
2-unique to the B&W OTSG design.
e- :..
e z.- -
- er :-
Following the incident at Three Mile Island, various a.ctions.wer.e.,ta.k.e.n, n.
3 to
.4 inchease the reliability of the auxil'ary feedwater systems and improve plant transient response. System modifications to increase.the reliability of the AFW may have resulted in more frequent AFW initiation. However. us,e of,AFW results,,
.in"in'troduction of cold (100*F vs. 400 F) feedwater into. the,more sensitive,
0 o
- a-
- 1. r t e
- 3 : :-
This may act to enhance system sensitivity.
upper section of the steam generators.
- e a+:: :
- .t 3:31- ; e t-:.
' F"u'r't'h'er system modifications provide control-grade reactor trips based on
,,.m.
i
- ; - ; s,*
- ; :
While secondary system malfunctions, such as turbine or feedwater pump trip.
sa:: u i :.4-
- .. i..
these rea'ctor trips do serve to reduce undercooling feedwater transients by s..
7,....,,..
~ reducing reactor power pranptly following LOMFW, they may: amplify: subsequent
- -t- : : :.
t:.
-s
.o'vercooling.
' A r' examination was made of small break and loss of feedwater events for B&W 3 :.
- t :< ;-:
t
-25
~~' plantIs'.
T'his resulted in a modification of operator procedures for dealing - r.
.c
~
a-3 it..e.
- e. :r e.4 with a small break, which include prompt RCP trip and raising the water level t:..
.:.:t inthe st'e'am generators to (95t) to promote natural ci.r.cu.l.a.ti.o.n.
B.o. t.h.th.e.s.e,.,.
' ' ' ' ~
s ez e:
- s actions are taken when a prescribed low pressure set I.oint is reached in the
- r :. r i :1 +-
e i : + ::--
- :' 'es: t:s-
- .:- 1: ' reactor coolant system and for anticipated transients such as loss of feedwater re::-
- a-
- ::r
- ':- e-these actions may amplify undesirable primary system respenses.
n--
t ' ::
u.5-In addition to the post-TMI changes discussed above, a,tions,were,also,taken _,.,
3:- -. -<->
c to reduce ths challenges to the power operated relief, valve (PORV) by raising,
the PORV set point and icwering the high pressure reactor trip. While these,,,,,
-+
actions have been succassful in recucing the frequency of PORY operation, they-z.
1 l
l E
3
..n.3
~~have,rWiulted in an increased number of reactor trips,, Jhi{ occurs,because.
...3 3.....
..s.the reactor will' now trip for transients it previously.would,have, ridden,.....
~
~ ~ "
I through by ICS and PORV operation.
~ The staff is con.carned by the inherent responsiveness.of B&W OT5G design. While some specific" instances are presented in the next sec,t.io,n of,this, paper, the staff.,,..
- 1-
"~ conc ~erWs'a're also of a general natare. It is felt that good design Practica:t +-=
r s-
.e-
" and inainYenanYe of the defense-in-depth concept, requggs a stableyel)-be,ag.
s...
h
' ' Yystem'.-g fo alarge part, meticulous operator attenti{n,and prompp manual. action
.. ~..........
3
~
~
~is 'use'd on "these plants to compensate for the system sensi,t1vi.ty,,rather,than,.
"~ '
any inherent design features.
~~
~~ ' ' Th'a 'st a'f'f' bel'ieles that the general stability of the B.&y p.lan.t, ontro,1. systems,,,,,,
~
c
sEculo b'e" imp' roved, and that plant response to OTSG feedwa,ter. perturbatiops.be.,,..,.
..a 7
'dampered.
II.
Recent Feedwater Transients
.s....
~ ~ ~ ~ 'O'n ALgUs't 23, 1979 the staff met with the B&W licensees.to discuss recent- : +
- =--
t:
7
~~feedwa'ter' t'ransients. One aspect which is of interest is the relationship of
~
- ~'
- rn
.t a::t:
et..:.t-
' th'e' ope'r'a' tor 'to the functioning of the m'ain feedwater,,syst.em.,,In.at, leas,t one,..,
.,.u..
t :--: :
sC s instanceanoperatormanuallyopenedablockvalvein..segi.e{w1,th,acontro,1yalve
- +i
. ' (pa'rtly' open but thought to' be closed). This resulted,,in an,oye.rf.ee.d co.nd.i,tjon...
In several recent events the feed flow was reduced to. the, pojnt whe.re the......,.
reacier trip'e'd on high pressure. Subsequeat overfeed reduced pressure,to,.below.
~~.,...
p
'1600 psi, where HPI was initiated, reactor coolant pumps trjpped, and auxiliary feedwater flow introduced into the top of the steam generators, which increased the severity of the cooldown transient.
-4 It appears that in many cases the main feedwater control system does not react
- re ::si.:: et:.
3-,3.;
. 3:
3..,.,:..,. :.
quickly enough or is not sufficiently stable to meet feedwater requirements.
t e.;
- . :.2 -
- ,. p -
,~...;
Rather, the system will often oscillate from underfeed to overfeed conditions, One causing a reactor trip and sometimes a high pressure injection initiation.
und'esirable element of this lack of stability is that overcooling transients
,a.g;
.3:.
.-.g.-
si.-
(decreasein
. on the primary side proceed very much like a small break LOCA
't-U e : e ::e Thus, for a certain period of time the operators pressurizer level and pressure).
- 3.,
3..
. 3 3.
g
--,33...
3 e.+--
t sa a The same may:eot know whethe-they are having a LOCA or an overcooling n
.a e.e-:
e 3.
5 :
This type of ~ behavior can be initiated by the normal reactor control system.
3 was de.onstrated by a December 1978 event at Oconee, where failure of "a cont
..-t 3_
i avg. recorder led to reactor trip, a feedwater transient, and E r-2; grade T A partial list of recent B&W transients and their eff'ects is' contained in the
.1 1 : 3 -r.
+:r. ii. : t :
Appendix to this report.
Role of the Pressurizer level Indicator
~ ~ ~ ~ ~ ~ ~
,131..
^
A major area of concern arising from the B&W OT5G sensTtfVi~ty,~is the respon 3,..
3...
....y.
3...
- - : t i e-:3 i
e: u Several B&W feedwater transients have led to of prg;;uri:er level indication.
-.y3,.....
,,3
- 3...
i s.e :e-
- 3-Most notatble was a November 1977 incident loss gf pressurizer level indication.
....3 e3-.
The arrival e1 -..:
at Davis Besse where level indication was lost for several minutes.....:.
3..
n 3
rate for this event appears to be 2n.the order of.1.2 per reactor year, but
.e.:..
r -et:::-
3 3..
- 3.. g.3 3..
could be on the increase due to the potential for more reactor trips and feed-
+e..
et: :-
......... 43.
3 :-
This is of water transients resulting from post-TMI-2 system modifications.
- :.. : e t:
- -u -
~
concern because an overcooling event could empty the pressurizar, thereby c
-e the potential for forming a steam bubble in the hot leg which may interrupt low pressure. The staff feels that
. n,a.tura.l. circulation, following RCP trip on the uncertainties associated with two phase natural circulation are somewhat high for an event with a recurrence interval of a few years..
i 1
5
.....; g.
~
~
h[ditEobally, the' staff believes that good design practica and adher.ence to.
th~e' def~en'se-in-depth concept, would require that plant. operators be, aware of.,..,
.a.the~ reactor's status during expected transients. A low-level off-4cale reading on pressurizer level makes it impossible for the operators to assess system
' ihventory and more difficult to differentiate between an accident and an
~
a... ~~...
excessive cooldown transient. The sta?f feels that the. frequency with wh.ich this situation occurs is undesirable.
' ' ~ Somi c6heerh's'alho exist with regard to the operation.of the.gr.essurize( heater.s.,,,
shen'lois 6f level takes place. Nonsafetygradeconttolcircuitry.tgips.jhe..,..
heatirs'6ff when pressurizer level is low.
If these.nonjafety. grade cutof,fs...,..,
~
~....
ih6uTd' fail, "t.he heaters would be kept on while uncov.er.ed,.This situatic.n pas..
~ ~ thi p6tehti$1*of overheating the pressurizer to the fyi,1ure, point,.as. happened.,...
.with a test reactor at Idaho Falls.
IV..
Role of ICS-MFW a
Jsag s.a.
-me-a-t,
- ~ ' e Thi ICS'ippiirs to paly a significant role in the plant's.feedwatet. t.esponse.......
e.
- a.
~ '
,,,,,i staff is Errently reviewing an FMEA study on the.ICS., Howevert. review cf.,,...
Th
" ' ~ ~ ~6periting'expirience suggests that the ICS often is a.contributet.to feedwater
~,.......
- : tr>an:s.ients.' : t :In some cases the ICS appeared inadequate to provide sufficient
.i n
' ' plant codtrol and stability. Some of the utility descripti.ons of fegdwater.
..a
trinsiints (as sunnarized in the minutes of a meeting.on. August 23.,192. 91........
. emphasizid the role of the operator in operating theJJW system..The.followi.ng,..
~secuence'illustrites the type of event and system response which the staff feels.
. m could potentially occur.
1.
Reactor at 100% power.
2.
Reactor trip, from arbitrary cause (does not matter).,.
.~'
3.
Plant stabilizes in hot shutdown, for a few minutes,. heat rejecti:n to condenser (and/or secondary dumo valves).
"#. 0verfeed transient (MW) (not uncomon to B&W) causes,overcoolingg 3 4
c,
~ pressurizer level shrinks, pressure reaches 160 psi, R1 actuatr.s;.,
'- " ~~3
....g, RCP tripped; A W cn.
(Possible RCP seal failure).
g.
'5.
Operat'or manually controls AN (possibly MFW instead or in addition, if.
~
2 MN n< t isolated such that OT5G level comes up to 95% of' operating range.
^ This massive additto6 9 co'd water.:ay lead to emptyirg of pressurizer
. 'and'interruptio* of natura' circulation (or, the hot 1,eg may flash due v.......,.,i....,.' to depressurization and incerrupt natural circulation.even.if pressurizer..
3 3
~~ ^
does not empty).
3,..
3....
~6."HPI delivers cold water, no heat transfer in OTSG; vapor from cor.e..
w....
~l leads to system repressurization; steam may condense or PORY.may lift...
~ #50 p5Yrestart criteria available, circulation may not be.reestabli,shed....
7.
'It' appears'thit an upgraded safety quality ICS, which is designed.}o balance.,
" power to OTSG 1evel in a better fashion, could reduce.the. sensitivity..
~~o.....
~
illustrated in the above sequence..
.V.
Role of ECCS and Auxiliary Feedwater y,.,...
'" ~~ 'Tisi Endwn'that some feedwater transients result in 9vercooling to the.txtenh,,3 3 g
that the HPI actuation setpoint is reached. Traditionally, the operatu isolates...
'letdo.....wn and' turns on an extra makeup pump following trip.so.as to. avert.this,...,.
" ~~
~ ~ ~'actuaSon.' 'If this manual action is not perfomed quickly.enough..or.if the g.
"cocidown trrnsient is too ' severe, the HPI set point will.be reached,and. the. pumps.
~'
- automatica1Ty started. Following procedures, the operator would then. trip all.mair.
~~
~ "
- ^'co'lanl p~ umps and utili:e recovery procedures based on the plant symptoms...If.,
^
c
'the 'in'cident was actually a feedwater event and not a.small LOCA, he would then
~
- ' presumably ge to the less of forced circulatien procedures. When pressure has recovered sucn tha* t.7e c:clant system has beceme 50 F,subcooled, the operator.
I can sec'u're "Mf! One :rcolem is the difficulty in differentiating between a.small
.h
brea k-LOCA....and an. excessive feedwater transient., The. operator.would be forced.,
.,,.y 3
3 e.
--~.....to assume a small LOCA until proven otherwise. However,3following.the small..
b'riak br'oc' dures and introducing cold auxiliary feedwater, cay incre~ase the --
e sever'ity of an overcooling event.
Initiation of AFW and delivery to the OSTG, especially if accompanied by filling to the high level required by new pro-cedures (955) will continue the cooldown and depressurjzation. Thus, the,AFW.
" '**s9Ethm acts to' increase the responsiveness of the reactor,to,feedwater t{ansients.33.
i~
where excessive cooldown is occurring.
..., g,n VI.
Conclusions
~*i "The' staff believes that the ctrrent B&W plants are over.ly. responsive to 3
I' feedkater transients because af the OTSG design, pressupt,:er sizing and.;...,;,...
' PORY and'high pressure trip set point. Some of the sensitivity also.arjses from 3..
" I' ~ U
". inadequacies in the ICS to deal with expected plant perturbatier.s..
" ";~
U Regar'dless of the reasons, B&W plants are currently experiencing.a numbeg_,
~
.,feedIater. transients which the staff feels are undesiegble,. The staff...
of
-believes that modifications should be considered to reduce.the.pl, ant. sensitivity :..,
' to th.....ese events and thereby improve the defense-in-deptn.which,wi11,,anhance
.3 the safety of the plant.
.., 33..
O s
APP [NDIX FEEDWATER TRANSIENT $UMMARY DESCRIPTION FACILITY TRANSIENT DA E CR-3 8/16/79(0259 Reactor Trip on liigh Pressure - 4 to' 3 RCP. A-S/G underfed -721 Pwr.
8/16/79 (1125)
Reactor Trip on liigli Pressure - 3 RCP - A-5/G underfed - 451 Pwr.
Reactor Trip on Higli Pressure - 3 RCP - A-5/G underfed - 48%Pwr.
8/t7/79 (0706) 8/17/19(1825)
Reactor Trip on liigh Pressure - 3 RCP - A S/G underfed - 261 Pwr.
8/02/79 (0202)
Reactor Trip on Low-Low Level in both S/G - 101 Pwr.
Turbine Trip - Antic. Trip did not work - Rx Trip on ill Press - 751 AND-1 8/13/79 (1749)
Oconee-1 6/11/79(0333)
Reactor Trip on Anti. Trip (L0fW) - 991 Pwr.
6/11/79 (0752)
Reactor Manually Tripped when FWPT "18" Tripped Oconee-2 5/07/79 (0346)
Reactor Trip on liigh Pressure - feedwater oscillations - 181 Pwr.
6/03/19 (2046)
- a. Reactor Trip on liigh Pressure - feedwater oscillations - 301 Pwr.
7/12/79 (1714).'a "
ReactoriTrip oniAntic. Trip (L0fW) - 1001 Pwr.
Rancleo Seco Davis-Besse
- NONE,
- *ffril'l(Off It' Af t i l !!! ItA ll
...vea
..n e.
4 lan : l's 8 ni l.
t,.l
/ / t g.<.
l a e gi
.n Ill.gti l'..
Pf it./ I's ( e s."
- I?
.' s
- . 16 i g, ir.
li e.gte t.-
in..
1 III.l*
P f,!
1*.E P 4.
It/18./ 7's (t1.".)
I..
t t t ill l'
't I
litil'..
't. I '/."a
( 0 / t se. )
I. s g..... I t ? i.
?s it. 4i.i l'i.gli f*,-
.nn I lit l' 8
i.!
.7 l'on It/ I 1/ 7's ( lfl.".)
l'..
.. t e i g.
..n l... l..
- 1. w l en 1.n l l. */t'.
- 't/ri.'/') i s t.'t s ? )
~.
.... - I,... s l'.1 li ip Ante.
le is.. flit niast wis t 6t.'l I.* s's ( l !1't)
'. t
.t.'
'. s ti (lati:s)
'rar 1
- /ItrIt in 'I t) l i gi.... ( ial i l
l ENCLOSURE t 1 REP - INITIAL PLANT STUDY l
l We have-ittempted to develop a general fram2 work for the conduct of a limited risk assessment of a B&W reactor aimed at identifying any unique risk-impacting An absolute determination of sequences relative to the Reactor Safety Study.
We have selected Crystal River 3 a plant owned and risk is not intenced.
operated by Florida Power Corporation, for analysis. The architect-engineer for this Babcock and Wilcox reactor was Gilbert Associates.
It began comercial i
operation in March 1977.
The project, as presented in Figure 1, will require the following tasks:
'. 3: :: ; n :-::
es 1
A surv ER files as now established in ORNL and A0 reports, as.
- .- +:.
r - welf :a.ey of the Ls-the Sandia and Flucr-Zion systems interactions' studies t
.:... : 3. ::
~
~"
zinteractions and common mode failures which have occurred in sTmilar plants.
-~
This'.,s'urvey should parallel construction of system logi_e M. odels; and event-trees since it will ensure that actual experience is incorporated into the'
- : -- :3:
assessments performed.
['
2e Event: trees.for loss-of-coolant accidents and transient conditions. Specific
.-., a.ttentton will be given to more frequent LOCAs and these wil-1 include a feed.:
- :,;i.-
.s : water.. transient tree which incorporates experience at B&W plants and 'will -
^~
+ i L
explorothe post-TMI modifications. Emphasis will be*giv'en toward. unde'r. L,
standing the human coupling interaction between systems at the event tree
- --
- :, : e e
~-~
5T
'~^
sequence level.
They will
- Fau.1.t trees for the key systems identified in the. event trees.
- 4.
- .. be. constructed to the component level and will include control, actuation.
and electric power considerations. Human errors will'be Tncluded as well Our as the ability of the operator to cope in the time span available.
- 1 a preliminary opinion is that simplified fault trees will be required for":
3.
auxiliary feedwater and secondary steam relief,'
n
- r.
sur e:
- .the. following systems:
3 : his.h pr. essure emergency core cooling in the injection'and'recircuTatioW~d 'rec d-r:
- c
.3-modes,- 19w pressure emergency core cooling in both infec U:
u study o.f. loss of AC power, considering the 480 and 4160 busses and 'the
- emergency diesel generators, with limited analysis of high voltage switcha -+ '
Separate fault trees will probably be required for ECCS and '
- i ::-:-yard f.aults.
- -- AFWS-initiation logic and the system trees must include'the contribution::'
- n from. auxil.iary systems such as instrument air, ventilation, component e--
Truncatiori of the fault trees
- 3. cooling, etc., and control-induced failures.
This basis will
.r wi.11 be permitted provided a written basis is provided.
present the rationale why no coupling of cutsets or event sequences is
~
expected from further development of the tree.
}'
An investigation of the adequacy of high pressure-low pressure interfaces..
4 Analysis of the physical phenomena associated with cominant secuences to 5.
This obtain estimates of the magnitude of releases fro. tne containment.
will aid in categori::irg releases into appros*iate reiease catescries.
-T6 ' conduct' a program of this magnitude in a short time per o, delays assoc
-s.
id lated with acquiring and transferring information must;be minimized. Optimally,, -
the event tree and fault tree analysts should share a common location during the initial portion of the project. As the fault trees progress below the top logic, however, the analysts should be located at or near the site with immediate access to as-built drawings and procedures as well as a representative of the plant operations staff. This will permit verification of engineering and procedural details and will minimize information transfer and print re-production. Access should also be arranged between the fault tree analysts
'at the site, the femaining tr.cm in Bethesda, the architect-engineer, and the vendor.
I.-?.?
.5fn:sddition'to basic plant data, deterministic calculations may be,This may - -
required to
- - understand the behavior of the plant under off-normal conditions.
~
- : also involve real-time siau14 tion a. an appropriate si.mulator to the extent
- possi_ble.!,'The arrangemer.ts with tha vendor should cover this possibility and it may'be' desirable to have confirma tory calculations made by one of.the NRC contractors on a selected basis i
e l
0 O
e 9
9
I 4
- q. y p s r.,
.' i t '. Y ' 1 *; eit j
.a ie-i. ' r e.
.p i..
30>
1 FICllRE 1 g
l r..
s eis t,
2 ANALYSIS OF ACCIDENT EVENT TREE 4
g PROCESSES C0tISTRUCTION AS NEEDED I.ER SUMYLI rof.L 10GElllER t.xtSTING INFO Y
=
ALREADY C0til'lLED V
=
CATECORI'.ATION g
LICENSING FAULT TREE 4$
quANTIFICATION OF EVENT REC 0ttlENDAT10HS C0:15TRUCTIOH 1REE SEQUENCES P
DETERHINISTIC CALCULATIONS BY VENIM)R AS NEEDED ANALYSIS OF I
"I li1Gil PRESSURE -
OF RESUt.TS 8
L0ld PRESSURE INT.
I.
i 1
.u u a..s. ens
)
a.
s i... :
e
't t's sin I)
. NRR AND PEER REV[l@li..,.,
)
<. - enin i..n I
A.
vi i i. :,
I l.
j' N
r-5:.
ENCLOSURE 3 PRELIMINARY IDENTIFICATION OF SYSTEMS AND COMPONENTS THAT MAY BE IMPACTE BY DESIGN CHANGES HPI System EFW System DHR System
+-
CFT System RCS Pressure Control System Makeup / Letdown System SG Pressure Control System
- ,3-3.
.g..
Steam Generator Pressurizer Quench Tank Control Room Layout
,3...
RCS Pipino o
T A/ F M O
~
llfl'fD h
STAFF LETTER TO CP ll0LDERS WITH BgW DESIGNS BACKGROUND REQUIREMENTS 3
e ANALYSIS OF OVERC00 LING EVENT o REQUIRED ASSUMPTIONS c
OPERATOR ACTION SIflGLE FAILURE SAFETY GRADE EQUIPMENT e
CONSTRUCTION
~
e DISCUSSION OF OTSG FEEDWATER TRAtlSIENTS,
e POSSIBLE DESIGN CilANGES a
e
e s
{
i I
1 1
1 i
STAFF CONCERNS RELATED TO OTSG i
RECEllT OPEPATING lilSTORY INDUSTRY RESPONSES IREP STUDY 4
O O
1 1
l I
I l
f l
. ~ ~
-4.
pfgQ W'
._ s # "~
RELEVANT DESIGN FEATURES OF B&W PLANTS i
OTSG 1.
SMALL SG WATER INVENTORY
~
~ 2.~
LIQUID VAPOR INTERFACE IN ACIIVE HEAT TRANSFER ZONE 3.
COMPLEX FEEDWATER CONTROL SYSTEM WHICH HAS -
_f.
- E:
n'
~
- 2i- -
A NUMBER OF FAILURE MODES WHICH RESULT IN
~~: :: -
~
OVERFEED AND UNDERFEED EVENTS I
1
ELFiFNTS OF SYSTEM RESPONSE _
PRIMARY SYSTEM HEAT TRANSFER RATE QUITE SENSITIVE TO LIOUID LEVEL Ill SG'S.
Z CONS $0VENTLY,PRIMARYSYSTEMPRESSUREANDPRESSURIZERLEVEL RESPOND OUICKLY TO CHAl!GES IN SG CONDITIONS.:- :' -
~
A NUMBER 0F FEEDWATER OVERFEED EVENTS HAVE OCCURRED WHICH HAVE : -.-
~~~EAUSEDRAPIDPRIMARYSYSTEMDEPRESSURIZAT10tl,SOMERESULTING 1
iNLOSSOFPRESSURIZERLEVELANDSAFETYINJECTI0NINITIATION.
'~'
PdST-Mi CTIONS TO IMPROVE AFW RELIABILITY MAY HAVE RESULTED IN --
MORE FREQUENT INTRODUCTION OF COLD AUXILIARY FEEDWATER INTO
~
-THE SGl OHICll SERVES TO INCREASE SEVERITY OF-OVERC00 LING EVENT
\\
~ ~
- NEW SG LEVEL SET POINT OF 95% FOLLOWING RCP TRIP.- AL_S_O AMPLIFIES. -:
OVERC00 LING TRANSIENTS FOLLOWING REACTOR TRIP.
- '. : 'l 7 ~2 ~~L0h'ERED NfGH PRESSURE TRIP SET POINT HAS INC_REASED FREQUENC. O Y
l hE5CTOR TRIPS, WHICH MAY LEAD TO INCREASED OVERC00 LING EVENTS.
~~ DUET 6LOSSOFPRESSURIZERLEVELINSOMEC00LD.0WtLTRANSIENTS, THERE' EXISTS INCREASED DIFFICULTY IN DIFFERENTIATING BETWEEN SMALL LOCA'S AND OVERC00 LING TRANSIENTS.
. i... ; i....,
- i
- i.
B&W OPERATING EXPERIENCE APRIL - NOVEL 1BER 1979 1
i..'
i i: ii-PLANT DATE POWER s
'CAUSE TRIP.; SIGNAL RESULTING OVERFEED CRYSTAL 08-16-79 72%
RCP SECURED llP YES 08-16-79 liS%
i:i 08-17-79 i 11 8 %
FW CONTROL PROBLEMS IIP.
i(1 i
. o
.+.
i l
08-17-79 26%
FW CONTROL PROBLEMS HP ot n:il i 08-02-79
!10%
?
SPURIOUS FW PUMP
..:,, i.,
i,,.
ia:i.
RUNBACK HP li.in i..
,,,; i.u; ANO 08-13-79 75%
RELAY FAILURE HP YES OCONEE 1 06-11-79 99%
LOSS OF B00 STER PUMPS
FWP TRIP
~
?
08-06-79
?
FWP TRI.P i.... i'.:....Ilede i.:
lia niI tilf\\ l i s.. o
.'s i
,B,EFORE TRIP.,
OCO, NEE 2 05-07-79 gg,8{
FW CONTRO,L, PROBLEMS
, g,,,IIPg,g,
.f.:W,.CONTR01 FAILURES IIP BEFORE TRIP 06-03-79, 30%
..i,Ili ta-1 u i:,il l e
i 0,CONEE 3 11-10-79 100%
'1
.n i i RhU,ik..af, RIP...i i i...'AkTS
'RAN 0 SEC0 07-12-79, IUU5 Tb.,.
i.
0Z1-22-79 10g%
, INVERTER FAILURE..
, g.,,,, y, i FW.MISMAT.CH HP j
1
CAUSES OF FEEDWATER EVENTS DIRECTLY ATTRIBUTED CAUSES FUNDAMENTAL CAUSES EQUIPMENT-FAILURE OTSGWHI_CH..CAUSESPRIli.ARYSYSTEM
~
~
~' ~
TO RESPOND QUICKLY TO CHANGES IN SG CONDITIONS.
~~~
OPERATOR ERRORS
- E. :
~~
~
II'-
~2' INABILITY OF ICS AND FEEDWATER
-NEED-FOR TUNING OF ICS SYSTEM TO~ MATCH FEED RATE WITH
~
~
~
~~
REACTOR POWER OUTPUT.
~
~
~
NECESSITY FOR PROMPT AND CORRECT
~ ~
~2
~~
OPERATOR ACTION TO PARTIALLY I ~
~ E' 0FFSET SENSITIVE SYSTEM
'~'
RESPONSE.
e O
f B & W OVERFEED TRANSIENT h)
],'
~
EVENT SEQUENCE EVENT TIME (S$CONDS)
POWER
=
100%
INITIAL CONDITIONS FLOW
=
100%
INITI L CONDITIONS
_.P.ZR RRESSURE
=
2157.8 PSIA
~NIfiAL-CONDITIONS;
~
I TURBINE TRIP INITIATED 0--
.. TURBINE FLOW RAliPS TO ZERO
~Y2i5SEC, 0
SCRAM 0;5SEC, ICS FAILS 5W!STAYSON AUX F.W. RAMPS TO 4.1%
1 - 4 SEC, PEAK PZR PRESS' RE
=
2232 PSIA
~ 5,4 SEC.
J PZR EMPTIES (PPZR
=
1589 PSIA) 42.7 SEC,
~
~
~
ECCS TRIP (AT P
=
1615)
~
- 41.5 SEC,
~~"
PUMP TRIP (ON ECCS TRIP)
'~
41.5 SEC,
.- ECCS DELIVERY (AFTER 6.4 SEC, DELAY)
EU,9-SEC,
~
~~E UPPER HEAD FLASHES 42.5 SEC, CANDY CANE FLASHES BETWEEN 45 AND 50 SEC, l
CANDY CANE QUALITY =
0.001 50 SEC, MAX. CANDY CANE QUALITY =
0.0088
+105 SEC, VOID
=
0,28 (HEM)
PRESSURE FALLS CONTIMUES
\\',\\
(,I
/
/;l
'3
+
\\
I 1, ',
'\\
,-l
-l..
i
&p
!h \\'
1/
('
i O
\\
\\
\\
pf h
i i
k 4
,1..
BASE, CASE f/',,
6 e;.2.. IIPI FLOW X 1.1 u.
- 3.,
T
~ ~
E
- 5. SCRAM R0D WORTH X 0.75 k
,h
.h k
fg PT. = 1715 PSIA
't
'g o.3
,:M
'g
- 6. PRESSURIZER AREA X 1.5 IV
/2 N
o N
\\
%e-l*
r%'
5 lp k
~
3
/
o.w ili as b
(
i /
e D
T 6
'\\
3 h
0$
f
~
Y a
~
\\
~
\\
l jQ i
\\
l!I 2
/
f
\\
o I
i 1
1 ii 1
1 1
I I
O 10 20 do 4O 80 60 70 00 90 /00 00 /20 /Jo & '/e Mo
/
g,
I WMEhC) t. v. :.I n
,s
~;
~
'j f
1,-l
. i}l' 3 i!-
W f,. ;i: i
.1 :
i/ l'. l'.::t s'
/
j L.
e
- s. i j
/
1
+
i Q
/,o -
'/ '
o.9
'1.$ASECASE.
- 2. HPI FLOW X 1.1
/
/[/
of
tt-)
,/
y g
II. HPI FLON X 1.1 & HPI SET C
/
b "7
PT. = 1715 PSIA
- 5. SCRAM R0D WORTH X 0.75 4/
o.6
- 6. PRESSURIZER AREA X 1.5
/
/
p o.5
/
g
/
q o.v
/
l' /
'/
N o.3
< A, R
o.2.
/
~
D
/
g) p o,i
/
r-i I
I I
i I
I l-1 I
I I
I I
I o
to ao ao 4to go &> 70 Bo 90 /m ilo /20 tao '{lo 'Soido 77ME (JEC.)
I I
!.'i.i l 11 i. I : i' ' ' i.I i
/'
g_
RESULTS OF ANALYSES
-2.n A 0 L
ya fasaq GM
~
NO FUEL DAMAGE ANTICIPATED; NO VIOLATI0fl:0F 2
DNBR LIMITS l
~
~
NO OVERPRESSURIZATION OF REACTOR C00LAtlT SYSTEM: ::--
PRESSURE BELOW 110% OF DESIGN VALUE e
i
AlWAGNd
~
1
~
~
g MdDIFICAT10flSWITHPOTENTIALTOREDUCEPL!\\NTSENSITIVITY TO CHANGES IN FEEDWATER SUPPLY -
~~
~
I.
MAIN FEEDWATER (11FW) SYSTEf1
~
~
SEPARATE SYSTEM TO LIMIT MFH ADDITION RESULTING.. _
n A.
FROM C0flTROL SYSTEM FAILURE.
-.i
~~~
l' B.
SYSTEM TO PROTECT AGAINST MFW OVERFILL AFTER
~
~
C.
ANTICIPATORY TRIP FOR LOSS OF MFN.
II.
AUXILIARY FEEDWATER (AFW) SYSTEM
~
~
~
A.
MODIFICATIONS TO LIMIT C00LDOWN FOLLOWING AFW-
~
INITIATION BY FLOW CONTROL AND/0R MULTIPLE SETP0ltlTS FOR FINAL LEVEL.
~
~5
~'
5.
AFW DIRECTED INTO LOWER PORTION OF STEAM. GENERATOR.-- -
~
C.
IMPROVED ALGORITHf1 FOR AFW PLANT CONTROL.
~
D.
AFW FEED ONLY GOOD GEllERATOR (F0GG).
1
M
~ -
/
.7 System / Components Equipment instamaten % Cornolete W NP WNP SNP SNP MP MP I
4 1
2 1
2 I
High Press. Ing. Sys.
100 0
100 100 100 100 Aua. Feedwater Sys.
100 0
100 100 100 100
.)
Cecay Heat Rem. Sys.
100 0
100 100 100 100 * ' " * '
~
l Core Food Tank Sys.
0 0
100 100 100 1' 0' ' ; " ' " '
0 RC Pres. Cont. Sys. IlCS) 0 0
50 10 20
'10 ' " "
'I j
Make Up' Letdown Sys.
100 0
100 100 100 100 SG Pres. Cont. Sys.
0 0
50 10 3
5
^^~E 3
Steam Generator 50 0
100 100 90
'90 ' ~ ~ '
E Pressurise' 0
0 100 100 90
'90* - '
Quench Tank 100 100 100 100 90
'90' J.
Control Room Layout 0
0 100 100 75
'3" ~
RC Sys. Pong 0
0 100 50 75 90 i
Main Feedwater Sys.
0 0
5 88 100 100 t
ww
T b/ M ::/e/n
~
IMFiEDIATE ACTI0tlS FOLLOWIt!G REAC10R TRIP
.-(
t
~
NOTE:
w-g
~
~-AfiYPARAM$TERMARKEDWITHANASTERISK(*)SHALLBEVER-IFIEDWHEN STEP 6.1 IS PERFORMED.
ll ~
EsIFYTHATALLCONTROLRODS,EXCEPTGROUP8,AREATTHEIR.
- ~ ~
_ ~
"IN LIMIT" AND flEUTRON LEVEL IS DECREASIflG.'
_E :
?
~
.2~
V$5IFi THAT THE IURBINE IIAS TRIPPED WITH ALL THROTTLE -STOPS AND GOVERl11NG VALVES CLOSED.
.3*
lERlFY Til$ TURBIflE BYPASS SYSTEM IS CONTROLLIflG HEADE.R PRE.SSURE-
~
AT t!0 LESS THAfl 1000 PSIG (TAVE v545 F).
.4 VERIFY THE SITE ELECTRICAL LOADS llAVE BEEN TRANSFERRED TO : __
STARTUP TRAllSFORMERS #1 AND #2,
- r_
~:..:
7:
.5 CLOSE THE LETDOWN ISOLATION VALVE SFV-22.009. _:.:
u _i~_.
'~ i6* -MONlTOR PRESSURIZER LEVEL AND f1AINTAIN 245"i - START 'AR HPI-PUMP IF NECESSARY.
~.7*
VERIFi'RC$I PRESSURE REMAINS ABOVE 1300 PSIG (SATURATION PRESSURE
~
0 FOR 580 F CORE TAVE). SEE ENCLOSURE 7.1 FOR ACCEPTABLE P/T RELATIONSHIP.
~-
.i.....
...,,..i4
/'/n/b fI ffu&MV i:1 Wl1:1 VA;c1h!i6lu.. I;i i I.
uill1 It !.
~'
NTNC4IfA#'
^
l.
LESSONS LEARNb'$ ROM THE YN"ACC'IDENT FOR NUCLEAR RISK
[
ASSESSMENT lin,.
- 1.
,y:,1iii:, !;n a A.
TMI DEMONSiTRATED AN EVERT SEQUEHCE LEADING TO CORE DAMAGE THAT WAS MORE PROBABLE THAN WASH-l'100 SEQUENCES DUE TO EQUIPMENT AND HUMAN INTERACTIONS B.
THI TRIGGERED A NEW LOOK AT PLANT-TO-PLANT DIFFERENCES I
1.
STUDY OF AUXILIARY FEEDWATER SYSTEM RELIABILITY IN CE AND WESTINGHOUSE PLANTS INDICATES WIDE PLANT-TO-PLANT VARIATION IN RELIABILITY 2.
OTHER EVIDENCE OF WIDE VARIATIONS IN RISK ATWS SUSCEPT!BILITY f-INTERFACING SYSTEMS LOCA t
!! I lld fll 1lilili'k' iris. 1..
l l i '.',,t.'! !. i. Eill i
A'....I '.i il 1; I l
/,!:i :,1.E!! !., I. I i N.li.;.
l.'
A.
I M ! lii i.ut':. i t..u.
- ( it;
!:.H.*ilij
. i;. ;... Ellll,i l'im!'li. 'l l ll!!:l; i.'. J l
. ' '! ; i r.t k i M:li imt;.r; I t l' r *.
',: i ri! i a ! '; r.. '
f I'l !
.i!: ; i Ib lill !.. IP2, tii i !:
1
~
INTEGRATEDRELTABILITYPil$,,Ahm.
R II.
OBJECTIVES
' I' A.
IDENTIFYANYHIbdR$$/EhEIDbhIE00ENCES l.
IN ORERATING.PLANTSvDSING;RSS METHODS 2.
BUILD FOUNDATIONS FOR FURTHER APPLICATIONS LONG RANGE DEVELOPMENT OF RISK-OR REllABillTY-BASED LICENSING ASSISTANCE i
TRAINING OF CADRE OF RISK ANALYSTS RELIABILITY MODELS USEFUL FOR:
~
o HUMAN ERROR RISK ASSESSMENTS o
COMMON MODE FAILURE STUDIES 1
p s.i i 9.a. SYSTEM.SJMESACTION STUDIES i1.
1:41t i.isiu. a n.
o:: it i i i./1 o OPERATIONS EVALUATION, ETC.
I II;I di1. i i.iu,11cli l(1:,K Ait il>l[il
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iii ili I
. !!;., i l'l!!', ll'.l :!,1 * '. Mi lillJii.
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- d II:,11 i 1ii i;tilli i 16 11 l I T'.1; I i i 1(I ',1 61:
n 1;; aill!'i li B.
METHODS 1.
EVENT TREE ANALYSIS 2.
SIMPLIFIED FAULT TREE ANALYSIS 3.
ESTIMATION OF RELEASE CATEGORIES FOR ACCIDENT SEQUENCES i
4.
QUANTIFICATION OF PROBABILITY OF INITIATING EVENTS AND EQUIPMENT FAULTS INCLUDING PRE-EXISTING HUMAN-CAUSED FAULTS 5.
SENSITIVITY STUDIES FOR RELIABILITY OF OPERATOR RESP 0flSE, AS REQUIRED l
~
45 B
.. i:. I:,
I.
1 ';i ;;l i
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lil.I i :, l ',
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a,tilli lil.in I
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. ~n ' '. I ol r!<iitil; III. PILOT STUDY
$YSTAL ilb$R n
a
,1
- n pi. N ii i. i. i t A.
OBJECTIVES PROTbbP$"0F'INNNIShNTbNI$S 1.
2.
ADDRESS B&W SENSITIVITY ISSUE 3.
ADDRESS INTEGRATED CONTROL SYSTEM INTERACTION ISSUE B.
PROGRAM PLAN l.
SURVEY BACKGROUND DATA:
LER'S AND A0'S FOR PRECURSORS, SYSTEMS INT $RACTIONS
,f
,I i
ICS FAILURE MODES AND EFFECTS ANALYSIS Ill. 111.i
.im.t-DMAMICRESfQNSEANALYSES(TMI&ATWSSTUDES)
A.
ihi l' I I IVi.
l.
Pl;tliti; ii
- s;i l' I:l',i. *.iili ll '.
i
n.n :,
i i is i, i,,.
a l i. :., i re i I m /t 2.
DEVELOP EVENT TREES
- TRANSIENTS
.O POWER CONVERSION SYSTEM OPERABLE O
POWER CONVERSION SYSTEM INTERRUPTED
- LOCA 0
>l FT.
(13.5 IN. DIAMETER) 0.4 FT.2 _ y p7,2 (8.5 IN DIAM - 13.5 IN DIA) 0 0.087 FT.2 _ 0.4 FT. 2 (4 IN DIA - 8.5 IN DIA) 0
'O
< 0.087 FT. 2 (13,,g,3)
- INTERFACING SYSTEMS LOCA h
,.1es' I i 841
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Il. d l l i I;l '
,.11tt it; & i
.8.! 1
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i
3.
DEVELOPSIMPLIFIEDFnULTTREES 11.
PRELIMINARY QUANTIFICATION AND RESULTS 5.
PREllMINARY FINDINGS AND RECOMENDATIONS -
MARCH 1980 i
6.
DETAILED ANALYSES, AS NECESSARY 9
(
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t!i i. il..i!/ii,
IV.
ANTICIPATED OUTCOME GIVENCE QUANTIFICATION IS IN PROGRESS n
PRELIMINARY ANALYSIS INDICATES THE Colfl0N DEPENDENCY o
OF TWO CLOSED CYCLE COOLING SYSTEMS WILL BE SIGNIFICANT THE PROBABILITY OF OCCURRENCE OF CORE MELT AND OF LARGER o
RELEASE CATEGORIES MAY BE IIIGilER THAN WASil-1400 o
ON A PRELIMINARY BASIS, IT DOES NOT APPEAR THAT THE EFFECTS OF Tile NSSS DESIGN WILL DOMINATE Tile RESULTS OF THE ANALYSIS I
i/,
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B.
ET/FT TECHNIQUES ARE ONE OF Tile BEST WAYS TO DISCOVER i
1! t.
ANDRANKACCIDENT'RiSKCONTRIBUTORSANDTOJUDGE IMPACTS OF IMPROVEMENTS 1.
SYSTEMATIC METHOD OF DISCOVERY 2.
AVOIDANCE OF TUNNEL VISION:
ISOLATED FIXES MADE OUT OF CONTEXT TilAT MIGHT INCREASE, NOT IMPROVE, RISK LEVEL i
3.
REALISTIC INTEGRATION OF " SAFETY" AND "NON-SAFETY" SYSTEMS li.
FOCUS ON FUNCTIONAL RELATIONSHIPS OF SYSTEMS 5.
IDENTIFIES IlUMAN INVOLVEMENTS IN MULTIPLE FAILURE EVENTS I ',.
I 1, i i 11 i llt.
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~
C.
WHERE RISK DOMINANT ACCIDENT SEQUENCES ARE FOUND, WE INTEND TO SUGGEST POSSIBLE OPTIONS FOR IMPROVING RELIABILITY AND REDUCING RISK D.
ASSIST NRR IN RESOLUTION OF CONCERNS RELATIVE TO STABILITY OF NSSS DESIGN i
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STATUS OF FAULT TREE ANALYSIS.
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12/18/79 l
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1
' SinFlified l Reduced 3
System Success System D
. System
, Schematic
' Criteria Fr
. Description l Evaluation Notes F
g I
I X4 Ac Power X
R F
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' Accumulators l
X AM I.
X CFCS I
CSIS
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X
'CSRS Containment Isolation X
X R
F "DC Power X
F X
.ESF Actuation X
X R
F X
HFIS F
X X
HFRS F
X X
LFIS
'F X
X LFRS X
X Nuc. Syc Water X
R F
X MS i
ICS.
X~
Instrument Air 8
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~
- ~ _.
CONSUMERS POWER COMPANY MIDLAND UNITS 1 AND 2 TENNESSEE VALLEY AUTHORITY BELLEFONTE
_ _i
~ : -
WPPSS WNP-1/4
~~
VEPC0 NORTH ANNA ~.2FJ siy
~
REQUESTS OF 10 CFR 50.54 LETTER tE. Z ~.:_.l. _,
1.
IDENTIFY MOST SEVERE OVERC00 LING EVENTS (ANTICIPATED TRANSIENTS AND ACCIDENTS).
STATE IF ECCS OR RPS ACTION IS NECESSARY FOR IDENTIFIED TRANSIENTS.
~'~
2.'
PROVIDE SCHEDULE FOR COMPLETION OF INSTALLATION 0F.
5.
.:._r :-
3'~
SYSTEMS AND COMPONENTS IDENTIFIED IN THE LET;TER.:. -
~~I 3.
FEASIBILITY OF HALTING INSTALLATION OF.~ SYSTEMS.VS.
EFFECTING CHANGES LATER.
E::i;-
..-- _. ?
4.
COMMENT ON OTSG SENSITIVITY TO FW TRANSIENT.
~
5.
RECOMMEND CHANGES TO DEAL WITH SENSITIV.ITY. -1;:
I' : :...,: -
- 5 l
~
AGENDA A.
INTRODUCTION AG HOSLER, WPPSS 0:__:. :B.
OVERC00 LING ANALYSES DW EABELLEEB&W _.. -
_ :.~..~
- C.
OTSG SENSITIVITY EA WOMACK; B&W
~~
D.
UTILITY RECOMMENDATIONS 1.
CPCo MJ SALERNO 2.
WPPSS AG H05LER 3.
TVA MA LINN :
E.
PEBBLE SPRINGS ITEMS
~_
~~.
l
Dwlale/Az tw l-l?Ak
/C/28 gg,- ps/
g+wb I OVERC00 LING EVENTS LIMITING CONSEQUENCE ANALYSIS OBJECTIVES 1.
DEFINE MOST SEVERE OVERC00 LING' EVENTS
- i ~.
- ~
~ ' ~
~:
(I.E., GREATEST INVENTORY SHRINKAGE)
~
2.
DEMONSTRATE ADEQUATE CORE COOLING
- NO OPERATOR ACTION FOR 10 MINUTES
~
- ONLY SAFETY SYSTEMS FOR MITIGATION
~ =~
~
~
~
3.
IDENTIFY NEED FOR RPS OR ECCS OR OPERATOR ACTION TO PROTECT CORE T.7.
4.
CONFIRM ADEQUACY OF DESIGN CRITERIA FOR -
2'~~
b L
CHARACTERISTICS OF SEVERE OVERC00 LING EVENTS
~ : " :. :. :
e REACTOR TRIPS ON HIGH FLUX OR LOW RC PRESSURE.
~
~~ ::
~
s ESFAS ACTUATED ON LOW RC PRESSURE OR LOW
~
~
SG PRESSURE (1)
HPI ACTUATED
~~
(2)
MSIV/MFIV CLOSURE G)
AFW ACTUATED
~
~
e ESFAS (VIA F0GG-TYPE SYSTEM LOGIC) ISOLATES E -
2:. ~
AFW FLOW TO DEPRESSURIZED OTSG
.. :- ~ -'
e OVERC00 LING. TERMINATED BY OTSG LEVEL CONTROL' ~~
~
OR OPERATOR ACTION
~'
~ ~ ~
e HPI CONTROLLED OR TERMINATED BY OPERATOR ACTION
~
1
RPS/ESFAS-frequency for overcooling events-at-B&W--plants Sunungy orsurrys AwALyzso INITIATING EVENT SINGLE FAILURE SENSITIVITY STUDIES I
\\
A.
ANTICIPATED EVENT MADE MORE
'Ii SEVERE BY SINGLE FAILURE REACTOR TRIP / TURBINE TRIP MAIN FEEDWATER OVERFEED B.
DESIGN BASIS OVERC00 LING DOUBLE ENDED STEAM LINE MAIN STEAM RELIEF VALVE e LOOP AT REACTOR TRIP BREAK STUCK OPEN e LOOP AT LOW RC PRESSURE ESFAS TRIP e DECAY HEAT e HPI SINGLE FAILURE e STEAM GENERATOR LEVEL CONTRO,L c
e BREAK ON DIFFERENT OTSGs j
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FIGNRE 4-4.
MFW OVERFEED, TURBINE TRIP, REACTOR TRIP CASE 1,.177 FA RCS TEMPERATURE VS TIME 800
(
590 1-580 570 O
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FIGURE 4-8.
MFW DVERFEEO, TURBINE TRIP, REACTOR TRIP-CASE 1, 177 FA PRESSURIZER LEVEL VS TINE..
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FIGURE 4-43.
SLB, CASE 3, 177 FA 33.2 i
27.67 22.14 i
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i FIGURE 4-41:
SLB, CASE 3, 177 FA l
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SLB CASE 3, 177 FA pi 300 ItI,\\
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Summar of events aliely-zed -
~
E ACTUAL DATA ALLOWED
~
[.hi. ~.'_
NLNBER FREQUENCY FREQUENCY
~
NO. OF REACTOR TRIPS (RPS) 228
~6 95/YR
~ 10/YR NO. OF AUTOMATIC ESFAS
~
~
'ACTUATIONS 27
.316/YR 1.0/YR NO. OF PLANTS INCLUDED 9
32.8 REACTCR-
~
YEARS O
e O
e 9
O m
a-
L-/3 CONCLUSIONS
~
~1) ' ADEQUATE. CORE COOLING CAN BE MAINTAINED FOR OVERC00 LING EVENTS...
.. ~: '2)'.RPS AIID ECCS ACTUATION IS REQUIRED AND ADEQUATE T0.MIIIGATE.r.
THE MORE SEVERE OVERC00 LING TRANSIENTS.
- 3) ~0PERATING EXPERIENCE CONFIRMS THAT THE ARRIVAL RATE DE TRAN= '. '
~
ISIENTS REQUIRING RPS AND ECCS ACTUATION IS WITHIN THE DESIGN :-' -
BASIS.
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F DEFENSE-IN-DEPTH FEATURES EXPECTED TO MITIGATE OR PREVENT e
LARGE VARIATIONS IN PRESSURE, PRESSURIZER LEVEL FOR FREQUENTLY OCCURRING TRANSIENTS.
e POTENTIALLY MORE SERIOUS UPSETS WILL THUS BE AVOIDED.
t e
LIMITING CASE ANALYSIS SHOWS CORE REMAINS SAFELY COOLED.
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t/lhlMILA-SOME POST-TMI CilANGES TO B&W OPERATIllG PLA!1TS AFFECTIi1G PLANT RESP 0ilSE e
BETTER AUXILIARY FEEDWATER SYSTEM RELIABILITY e
ADDITION OF ANTICIPATORY REACTOR TRIP FUNCTIONS e
CHANGE TO PORV AND HIGH-PRESSURE TRIP SETPOINTS e
RC PUMP TRIP ON LOW PRESSURE e
INCREASED SECONDARY WATER LEVEL FOR NATURAL CIRCULATION
- a i ':: :.
'+
i.
.iii...
. 1 ;-
3 6+i
A%k gy//T
~~
RECOMMENDATIONS FOR WNP-1/4 A.
CHANGES TO RETAIN THE ORIGINAL NSSS OPERATING CHARACTERISTICS 1.
QUALIFY PORV 2.
PROVIDE 1E CONTROL AND POWER TO PORV 3.
PROVIDE 1E PORV ISOLATION (BLOCK) VALVES ACTUATED BY ESFAS ON low (<1600 PSIG) RCS PRESSURE.
l RECOMMENDED CHANGES FOR WNP-1/4 (CON'T) i B.
CHANGES TO IMPROVE SECONDARY SYSTEM RELIABILITY I~
I 1.
INCREASE WATER MAKEUP CAPACITY TO CONDENSER HOTWELL DURING RUNBACK AFTER I-G TRIP.
2.
PREVENT A SINGLE FAILURE IN ICS FROM i
PROVIDING 15% STEAM DUMP CAPACITY.
3.
IMPROVE CONTROL RESPONSE OF THE ICS FOLLOWING SENSOR FAILURE.
\\
l
RECOMMENDED CHANGES FOR'WNP-1/4 (CON'T) 9 [ r~ h i
C.
CHANGES TO IMPROVE THE RESPONSE OF THE NSSS 1.
PROVIDE FOR RAPID MFW FLOW REDUCTION FOLLOWING REACTOR TRIP 2.
ADD 1E LOSS OF ALL FW TRIP N ~CCG 0M /""'
7 3.
ADD MFW OVERFILL PROTECTION
~~
4.
REDUCE POTENTIAL FOR AFW OVERFILL FOR _
SINGLE FAILURES 5.
IMPROVED ALGORITHM USED FOR AFW CONTROL
RECOMMENDED' CHANGES'FOR WNP-1/4 (CON'T)
D.
CHANGES TO IMPROVE THE CAPABILITY TO MITIGATE TRANSIENTS 1.
PROVIDE 1E LOW LEVEL CUT OFF OF HEATERS.
E 2.
IMPROVE AFWS CONTROL FOLLOWING ESFAS ACTUATION 3.
TRIP RCP ON LOW RCS PRESSURE AND VOID DETECTION 4.
PROVIDE IMPROVED F0GG LOGIC,
RECOMMENDED STUDIES FOR WNP-1/4
-(hf 1.
SECONDARY SYSTEM RELIABILITY 2.
CONTROL AIR SUPPLY SYSTEM 3.
MINIMUM FINAL FW RESPONSE STUDY 4.
AFW TURBINE RELIABILITY 5.
NNI/ICS POWER SUPPLY RELIABILITY 6.
HEATER DRAIN PUMP RELIABILITY STUDY
s n :..
R005070[db
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q NUCLEAR REGULATORY COMMIESION WMHIN r,TOf J, D. C. 205L5
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APR 2119E0 FCT:: RHO 71-0249 Nuc'. ear Fuel Services, Inc.
ATTS:
lic. C. J. li1chel P.O. Eox 218 Erwin, TN 37650 Gentlemen:
We " ave evaluated your Quality Assurance Program submitted with your March 17, 193: letter to satisfy the requirements of 10 CFR 571.51.
Our review indicates that additional information is required to satisfy the 3;piicable requirements of Appendix E to 10 CFR Part 71.
Please address the er. closed request for additional information and submit seven copies of the revised program within 60 days following receipt of this letter.
If you have any questions regarding this request, please feel free to contact Mr. Jim Conway at (301) 492-7741.
Sincerely,
?.t<@ '#
[,~,-CharlesE.MacDonald, Chief p
Transportation Certification Branch Division of Fuel Cycle and Material Safety, tiMSS Erci:sure:
2ewast for Additional Ir.fo rmation
r
'4 i;UCLEAR FUEL SERVICES (71-0249)
P.e_q_uest for Additional Information 1.
r: vide a stateT.ent that designated QA individuals have the responsibility and au:h:rity, delineated in writir.g, to stop unsatisfactory work.
2.
rovide a list of the QA procedures plus a matrix of these procedures cross eferenced to each criterion of Appendix E to 10 CFR Part 71.
3.
Ipplicable design control requirements should be in effect for modification and g air activities associated with existing shipping containers.
Accordingly,
- r: vide statements that:
2.
Quality standards are specified in the design documents, and deviations and changes from these quality standards are controlled.
Essigns are reviewed to assure that (1) design characteristics can be controlled, inspected, and tested and (2) inspection and test criteria are identified.
Proper selecticn and accomplishment of design verification or checking processes such as by design reviews, alternate calculations, or qualifi-cation testing are performed.
When a test program is used to verify the adequacy of a design, a qualification test of a prototype unit under casign conditions should be used.
d.
Individuals or groups responsible for design verification are other than the original designer and the designer's immediate aupervisor.
Cesign and specification changes are subject to the same design controls e.
and approvals that were applicable to the original design unless f4FS cesignates another qualified responsible organization.
The positions or groups responsible for design reviews and other design verification activities and their authority and responsibility are identi-fied and controlled by written procedures.
2 Frevide a statement that procurement documents contain or reference the design
- 2 sis technical requirements including the applicable regulatory requirements, aterial and component identification requirements, drawings, specifications,
- Mes and industrial standards, test and inspection requirements, and special
- rccess instructions.
5.
Frtvide a statement that procurement documents identify the documentation (e.g.,
crawings, specifications, procedures, inspection and fabrication plans, inspection ir.d test records, personnel and procedure qualifications, and chemical and physical
- est results of material) to be prepared, maintained, and submitted to l4FS for revies and approval.