ML16161B097
| ML16161B097 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 10/05/1979 |
| From: | Ross D NRC - TMI-2 BULLETINS & ORDERS TASK FORCE |
| To: | Gill R BABCOCK & WILCOX OPERATING PLANTS OWNERS GROUP |
| References | |
| IEB-79-05C, IEB-79-06C, IEB-79-5C, IEB-79-6C, NUDOCS 8002110224 | |
| Download: ML16161B097 (25) | |
Text
0 RUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Docket NOS.517 Docket No.
O
, 50-270, 50-287 OCT 51979 o4 50-289, 50-302, 50-312 50-313, 50-346 Robert Gill, Chairman TMI-2 Effects Subcommittee B&W Oner's Group P. 0. Box 2173 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Gill:
The NRC staff is reviewing the responses to bulletins79-05C and 79-06C on the subject of the need for tripping of reactor coolant pumps for certain small break loss-of-coolant accidents. The conclusions reached by the PWR vendors in those responses vary to a considerable.degree, and are highlighted in Table 1. To a degree these differences may be attributed to design differences. However, our present judgment is that the major differences are attributable to model differences, which are highlighted in Table 2. The various RCP trip criteria that have been proposed or are believed to be in place are presented in Table 3.
A variety of model features have been developed without the use of relevant experiments to support model justification, and no model has been demonstrated to be overall conservative by integral test comparison. There is apparent lack of agreement between vendors as to whether individual assumptions are conservative. Given these model variations, it is not surprising that the con clusions reached vary.
Thave done some calculations with RELAP-4, Mod 7, and these are summarized in Table 4. These calculations, while useful to us, are not definitive, and cannot cuide us as to what is a suitably conservative set of model assumptions.
Our concern is that during the pendency of traditional interactions between you and us (i.e., questions, positions, analyses, rebuttals, appeals, etc.) events are forcing a more prompt solution. I refer to the RCP trip at North Anna (a non-LOCA transient) the RCP trip at Prairie Island (a steam generator tube failure) and the Davis-Besse transient of September 27, 1979 which nearly set in motion the requirements of 79-05C.
The second half of the concern I have relates to HPI termination criteria. Table 5 lists several criteria proposed, in place, or thought to be in place. There no apparent reason why the same set of safety considerations would lead us into than one, uniform criterion for termination of HPI.
Your assistance in achieving this desired state of unifornity is needed.
If the NRC were to adopt a Q&A approach to solving the problem separately, some candidate questions are enclosed in Table 6.
800211o0Z214LI OCT 5 1979 Instead of withdrawing to the Table 6 approach, I suggest we meet soon and discuss, in an administrative sense, how the PWR regulated industry can close on these two issues.
For further details regarding the issues described herein, as well as the planned meeting, Mr. Brian Sheron (301-492-7588) is available for further discussion.
Sincerely, D. F.. oss, Jr., Director Bulletins and Orders Task Force Office of Nuclear Reactor Regulation
Enclosures:
Table 1 - Conclusions Reached by PWR Vendors in Response to Bulletins79-05C and 79-06C Table 2 - Differences During SBLOCA with Pumps Running Table 3 - RCP Criteria Proposed or In-Place at Plants Table 4 - Staff Calculations for PWR Vendors Table 5 - HPI Termination Criteria Proposed or In-Place at Plants Table 6 - Typical Questions on RCP Trip and HPI Termination cc:
Babcock & Wilcox Company B&W Licensees Identical Letters sent to:
Ed Scherer, CE Tom Anderson, W George Liebler, CE Owner's Group Cordell Reed, W Owner's Group James H. Taylor, B&W
TABLE I CONCLUSIONS REACHED BY PWR VENDORS IN RESPONSE TO BULLETINS79-05C AND 79-06C EFFECT OF EFFECT OF TRIPPING MAXIMUM AVAILABLE CONTINUOUS ONE PUMP BREAK LOCATION BREAK SIZE TIME FOR PUMP TRIP PUMP OPERATION IN EACH LOOP B&W RESULTS NOT LIMITING BREAK
-3 MINUTES ACCEPTABLE CORE NO EVALUATION SENSITIVE DUE SIZE ABOUT 2 (BASED ON PRELIMINARY COOLING TO HOMOGENEOUS 0.02 - 0.2FT CALCULATIONS)
MODELING ASSUMPTION CE*
FOUND HOT LEG LIMITING BREAK 6 MINUTES AFTER 0.1 FT2 BREAK ACCEPTABLE CORE BREAKS LIMITING/
SIZE ABOUT 2 TRIP + SIAS FOR IN HOT LEG LEABS COOLING FOR BE SOME COLD LEG
.02 -.1 FT EM ANALYSIS 10 TO PCT's,12200 F ANALYSIS PROVIDED BREAKS COULD MINUTES AFTER TWO PUMPS TRIPPED EXCEED 2200oF TRIP + SIAS FOR WITHIN 5 MINUTES BE ANALYSIS AFTER BREAK W
COLD LEG BREAKS LIMITING BREAK 10 MINUTES FOR ALL ACCEPTABLE CORE NO EVALUATION LIMITING, NO HOT SIZE.02-.05 PLANT TYPES (2, 3, COOLING LEG BREAKS FT 4 LOOPS)
ANALYZED RESULTED IN PCT'S> 2200oF
1200 PSI HPSI PUMPS. ANALYSES CONSIDERED CONSERVATIVE WRT PLANTS WITH 600 PSI SITS AND/OR HIGHER HEAD HPSI PUMPS.
TABLE 2 MODEL DIFFERENCES DURING SBLOCA WITH PUMPS RUNNING MODEL Item W
CE B&W RELAP/MOD-7 Cold Leg Stratified Flow Homogeneous Homogeneous Heterogeneous Pump Discharge flow flow flow Pipe Downcomer Heterogeneous Model switches Homogeneous Heterogeneous model from homogeneous flow flow to heterogeneous model when drift velocity criteria met.
Core Heterogeneous Heterogeneous Homogeneous Heterogeneous flow flow flow flow Hot.Leg Pipe Homogeneous Heterogeneous Homogeneous Heterogeneous flow flow CE represents flow flow. No counter the hot leg with two current flow flow paths. This allowed representation allows counter current flow in horizonal paths.
Steam Homogeneous Drift flux Homogeneous Heterogeneous flo Generator flow model - allows flow no vertical slip/
Hot Side liquid fallback fluid runback to Tubes to hot leg if hot leg possible Steam Homogeneous Homogeneous Homogeneous Heterogeneous flc Generator flow flow flow no vertical slip Cold Side Tubes Cold Leg Homogeneous Homogeneous Homogeneous Homogeneous Loop Seal flow flow flow flow (suction pipe)
Model/Method W
CE B&W EG&G Idaho ECC Injection No injection No spillage 30%
Consistent with assumed in assumed for spillage of vendor broken loop hot leg water assumptions for cold leg breaks -
injected in breaks no injection broken loop assumed in broken for cold loop for cold leg breaks leg breaks ECC Injection Downcomer/lower Downcomer Cold Leg W - upper downco.
Location plenum node (cold leg by (cold leg by CE -
cold leg (cold leg by design) design)
B&W -
cold leg design)
Quench No carryover No carryover No carryover No carryover Behavior during accounted for accounted for accounted for accounted for Recovery Steam Super-Superheating 12 axial coolant No superheat 3 axial coolant Heat Calcu-considered nodes in core, calculated due nodes in core.
iation (description Superheating to single Superheating prooretary) of each node control volume of each node allowed model of core. allowed.
All core. heat added to liquid phase.
Separate heat up model cal culates super heat but uses CRAFT mixture level.
Core fluid Thermodynamic Thermodynamic Thermodynamic Thermodynamic quality equilibrium equilibrium equilibrium equilibrium assumed -
assumed -
assumed -
assumed actual actual quality actual actual quality quality not not calculated quality not not calculated calculated calculated
TABLE 3 RCP CRITERIA PROPOSED OR IN-PLACE AT PLANTS NRC CRITERIA A. Upon reactor trip and initiation of HPI caused by low reactor coolant system pressure, immediately trip all operating reactor-coolant pumps.
B. Two licensed operators.
WESTINGHOUSE A. Westinghouse proposed criteria is that reactor coolant pumps be tripped on low reactor coolant pressure of 1250 psig after verification of high head safety injection.
B. Responses from licensees are under review - most adopt NRC criteria.
COMBUSTION ENGINEERING A. Stop two reactor coolant pumps, one in each loop after it has been verified that the rods have been inserted fully for five seconds. If this action has not been completed within the first five minutes after SIAS, all reactor coolant pumps must be stopped within 10 minutes after SIAS.
- i.
Responses from licensees are under review - some adopt NRC criteria.
BABCOCK & WILCOX A. B&W proposed criteria is that reactor coolant pumps be tripped on initiation of HPI caused by low reactor coolant system pressure.
Licensees are currently using NRC criteria*.
- Most recent proposals for future automatic pump trip have the signal from saety injection only (not low pressure or current).
TABLE 4 STAFF CALCULATIONS FOR PUR VENDORS PUR VENDOR BREAK SIZE BREAK LOCATION RC PUMPS TRIPPED FLUID MODEL ASSUPTION1S UESTINGHOUSE 4 INCH DIAMETER COLD LEG AT REACTOR SCRAM HETEROGENEOUS 4 INCH DIAMETER COLD LEG NOT TRIPPED HOMOGENEOUS (bEFORE TPIP / AFTER TRIP) 4 INCH DIAMETER COLD LEG AT 511 SECONDS WORST CASE)
HOMOGEMCOUS/HETEROGENKOUS 4 INCH DIAMETER COLD LEG AT 760 SECONDS HOMOGENEOUS/HETEROCENEOUS 1 INCH DIAMETER COLD LEO AT REACTOR SCRmM HETEROGENEOUS 1/2 INCH DIAMETER COLD LEG AT REACTOR SCRAM HETERUGEHEOUS I STUCK PORU PRESSURIZER AT REACTOR SCRAM HETEROGENEOUS COMBUSTION ENGINEERING 8.1 SQUARE FEET COLD LEG AT REACTOR SCRAM HETEROGENEOUS 6.02 SQUARE FEET COLD LEG AT REACTOR SCRAM HETEROGENEOUS I STUCK PORU PRESSURIZER AT REACTOR SCRAM HETEPOGENEOUS 2 STUCK PORU'S PRESSURIZER AT REACTOR SCRAM HETEROGENEOUS BASCOCK I UILCOX
- .-I SQUARE FEET COLD LEG AT REACTOR SCRAM HETEROGENEOUS 6.07 SQUARE FEET COLD LEG AT REACTOR SCRAM HETEROGENEOUS I STUCK PORU PRESSURIZER AT REACTOR SCRAM HETEROGENEOUS
TABLE 5 HPI TERMINATION CRITERIA PROPOSED OR INPLACE AT PLANTS NRC CRITERIA Operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been automatically actuated because of low pressure condition, it must remain in operation until either:
(1) Both low pressure injection (LPI) pumps are in operation and flowing at a rate.in excess of 1000 gpm each and the situation has been stable for 20 minutes, or (2) The HPI system has been in operation for 20.minutes*, and all hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure. If 50 degrees subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated.
The degree of subcooling beyond 50 degrees F and the length of time HPI is in operation shall be limited by the pressure/temperature considerations for the vessel.integrity.
BABCOCK & WILCOX
'A. The LPI System is in operation and flowing at a rate in excess of 1000 gpm in each line and the situation has been stable for 20 minutes.
or B. All hot and cold leg temperatures are at least 500 below the saturation temperature for the existing RCS pressure, the hot leg temperatures are not more than 500 hotter than the secondary side saturation temperature, and the action is necessary to prevent the indicated pressurizer level from going off-scale high. If 500 subcooling cannot be maintained,. the HPI shall be reactivated. The degree of subcooling beyond 500 and the length of time HPI is i.n operation shall be limited by the pressure/temperature considerations for the vessel integrity."
-rent taff view is that this time Deriod should be deleted (see Table 6)
Table 5 continued
-2 C. All licensees have adoptedethese criteria, for the most part.
(UNDER REVIEW)
The criteria proposed by W are as follows:
"In the event of a spurious safety injection signal, the sequence of reactor trip, turbine trip and safeguards actuation will c:cur.
The operator must assume that the safety injection signal is non-spurious unless the following are exhibited:
- a. Normal readings for containment temperature, pressure, radiation and recirculation sump level AND
- b. Normal readings for auxiliary building radiation and ventilation monitoring AND
- c. Normal readings for steam generator blowdown and condenser air ejector radiation AND
- d. All steam generators exhibit normal pressure and water level following reactor trip and safety injection actuation (similar to a reactor trip:freM normal condttions).
IF all of the symptoms a through d above are met, THEN secure Safety Injection when the following are exhibited:
- a. Reactor coolant pressure is greater than 2000 psig and increasing AND
- b. Pressurizer water level is greater than 50% of span AND
- c. Water level in at least one steam generator is in the narrow range span, or in the wide range span at a level sufficient to assure that the U-tubes are covered."
Table 5 continued
-3 COMSUSTION ENGINEERING.- -(UNDER REVIEW)
"After any SIAS, operate the SIS for at least 20 minutes and until RCS hot and cold temperatures are at least 500F belovi the saturation temperature for the RCS pressure unless the cause of the SIAS has been verified to be an inadvertent actuation. If 500 subcooling cannot be maintained after the system has been stopped, the high pressure injection system must be restarted."
TABLE 6 TYPICAL QUESTIONS ON RCP TRIP AND HPI TEWRTNATION We believe that if it is necessary. to trip the RCP for small break LOCA, then it should be done automatically. Further, such automatic action should be from circuits of high quality usually associated with the reactor protection system. This would provide a high likelihood that the action would take place when needed, and at the same time would reduce the likelihood of unneeded trips.
Since you have indicated that, for some breaks, RCP trip is necessary, it follows that RCP automatic trip circuits should be used. For this reason it is requested that you comply with the following:
- b. Support the choice of this alternate signal by showing that it will conservatively protect the core during the most limiting LOCA (in the critical break range) and that it will not be actuated for the most limiting non-LOCA event.
- c. Provide an auotmatic trip system that meets IEEE-279 requirements; particularly with respect to meeting single failure for tripping all the pumps and with minimum likelihood for spurious actuation.
Submit a proposed Technical Specification to include limiting conditions of operation and surveillance requirements
- d. Because of recent non-LOCA events which actuated the HPI, the staff believes that simple HPI termination or throttling criteria tied to subcooling of the fluid in the core and hot and cold legs would be appropriate for hypothesized events, regardless of pressure in the primary system. Our examination of pressurizer level traces obtained during recent transients indicates that this parameter should be considered
-2 as a coincident instruction to the operator. A simple RCS pressure criteria is not useful during SG tube failure, SLB, overcooling transients, especially if one must also consider potential pressure vessel integrity problems.
Provide arguments why subcooling should not be used or should be tempered by other indications.
- 2. Adding a protective feature such as RCP trip must have some overall effect on the risk characteristic of the reactor. For the general class of LOCAs, it seems reasonable that:
- a. for large LOCA there is no effect;
- b. for most small break LOCAs the pumps are expected to be powered (i.e.,
low likelihood of LOOP) and continued RCP running would offset some degraded ECCS conditions; and,
- c. for a small range of LOCAs continued pump running may be counterproductive.
However, for non-LOCA events, especially transients, it seems reasonable to postulate that spurious RCP trip increases risk.
Indicate whether adding the RCP auto trip feature poses a significant shift in the risk characteristic.
- 3. When RCPs are tripped due to a SI signal, there is (for some designs) no pressurizer spray. This, in combination with continuous HPI operation, may lead to PORV operation. Thus a transient (such as overcooling) may progress to a degree.not heretofore contemplated. Indicate the effect of no pressurizer spray on pressure control during the set of transients. How can auxiliary spray be supplied, if needed, in the presence of an SI signal?
- 4.
It seems like auto RCP trip, if based on a pressure signal in whole or in part, should be based on the lowest possible pressure.
Presumably this trip pressure would be below the SI setooint, but above the steam generator safety valve
-3 setpoint.
Please comment on the parameters needed to generate a RCP trip signal, and how they might respond to non-LOCA events.
- 5. HPI termination based on adequate subcooling would avoid undesirable formation of voids in the RCS. If, however, the pressurizer were low in level it would be desirable to continue adding coolant. Thus low pressurizer level would be a parameter that one would logically monitor before terminating HPI. If the steam generators have all boiled dry (from arbitrary causes) then the only available heat removal path will be through the PORV. This is generally an ICC* situation. In this instance the PORV will be opened and the parameters of subcooling, pressurizer level, and RCS pressure probably have no further influence on the courses of action.
Thus we believe that subcooling should be a permissive before reducing HPI flow that may have been automatically actuated. Common sense says that the makeup systems should continue in operation until pressurizer level is restored.
(See question 1-d).
Please consider why these parameters are not adequate for all situations except ICC.
- 6. Provide verification for the individual modeling assumptions used in your analyses of SBLOCAs with the pumps on; in particular the flow regime modeling.
What verification is proposed for integral model assessment?
Alternatively, explain why your present model is appropriately conservative to bound the analysis uncertainties or propose the additional conservatisms needed to assure the uncertainties are adequately bounded.
- 7.
Recent experience has shown that many non-LOCA events will exhibit front-end
- Inadeouate Core Cooling
-4 symptoms of a SBLOCA. In particular, the.events at North Anna land Prairie Island indicate that loss of the RCPs can increase the difficulty in bringing the plant to a hot standby condition. The staff therefore feels that many of the events analyzed in Chapter 15 of SARs need to be reanalyzed to account for RCP trip in the event of SI initiation (i.e., SG tube rupture, secondary side overcooling events, etc.).
Provide alternative criteria for SBLOCA pump trip which significantly decrease the probability of requiring the trip for non-LOCA events, or provide reanalyses showing the acceptability of affected Chapter 15 events assuming.pump trip.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of METROPOLITAN EDISON COMPANY,
(Three Mile Island, Unit 1)
CERTIFICATE OF SERVICE Ivan W. Smith, Esq.
E 11yn Weiss, Esq.
Atomic Safety & Licensing Board Panel Sheldon, Harmon, Roisman Weiss U.S. Nuclear Regulatory Commission Sute5
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20006 Dr. Walter H. Jordan Mr. Steven C. Sholly 881 W. Outer Drive 304 South Market Street Oak Ridge, Tennessee 37830 Mechanicsburg, Pennsylvania 17055 Dr. Linda W. Little 5000 Hermitage Drive Mr.
Thomas Gerusky Raleigh, North Carolina 27612 Bureau of Radiation Protection Department of Environmental Resources George F. Trowbridge, Esq.
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Washington, D. C.
20006 Mr. Marvin I.
Lewis 6504 Bradford Terrace Karin W. Carter, Esq.
Philadelphia, Pennsylvania 19149 505.Executive House P.- 0. Box 2357 Metropolitan Edison Company Harrisburg, Pennsylvania 17120 ALtn:
J. G. Kerbein, Vice.PresidtT-P.O. Box 542 Reading, Fennsylvania 19603 Honorable Mark Cohen 512 E-3 Main Capital Building M-s.
Jane Lee Harrisburg, Pennsylvania 17120 R.D.
3; Box 3521 Etters, Pennsylvania 17319 GopFpe~tn Walter W. Cohen, Consumer Advocate Holly S. Keck Department of Justice Anti-Nuclear gropp Representing Strawberry Square, 14th Floor York Harrisburg, Pennsyl\\vania 17127 245 W. Philadelphia Street York, Pennsylvania 17404 Robert L. Knupp, Esq.
Assistant Solicitor John Levin, Esq.
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P.O. Box P Box 3265 407 N. Front Street Harrisburg, Pennsylvania 17120 Harrisburg, Pennsylvania 17108 Jordan D. Cunningham, Esq.
John E. Minnich, Chairman Fox, Farr and Cunningham Dauphin Co.
Board of Conmnissioners 2320 North 2nd Street Dauphin County Courthouse Harrisburg, Pennsylvania 17110 Front and Market Sts.
Harrisburg, Pennsylvania.1.7101 Ms.
Kathy McCaughin Three Mile Island Alert, Inc.
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Marjorie M. Aamodt Atomic Safety and Licensing Board Panel
. M 5o o
U.S. Nuclear Regulatory Commission Coatesville, Pennsylvania 19320 Washington, D. C. 20555 Docketing and Service Section U.S. Nuclear Regulatory Commission Ms.
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Frieda Berryhill, Chairman Coalition for Nuclear Power Plant Postponement 2610 Grendon Drive Wilmington, Delaware 19808 Mrs.
Rhoda D. Carr 1402 Marene Drive Harrisburg, Pennsylvania 17109
DISTRIBUTION Docket File (*)
NRC PDR (*)
LPDR (*)
ORB 4 Reading NRR Reading B&O T/F Reading H Denton E Case V Stello R Vollmer W Russel B Grimes T Carter D Eisenhut R Reid D Garner M Fairtile C Nelson G Vissing D Dilanni H Silver S Lewis M Mulkey D Davis R Ingram P Kreutzer OI&E (6)
R Fraley, ACRS (16)
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Licensing Board Pu
,f;fce cf the Secretary U. S. Nuc ear ReCuLa:orV Comissic U. S.' Ncer Reula tcry Commissicn Dashirton, D.2.
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aashirc:on, D. C..
2055 U. 5.
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70 edo Edison Company Ivan W. Smith, Esq.
Atomic Safety and Licensing Soard Panel U. S. Nuclear Regulatory Comission Washington, D.C.
20555 Dr. Cadet H. Hand, Jr.
Director, Bodega Marine Laboratory University of. California P. 0. Box 247 Bodega Bay, California 94923 Dr. Walter H. Jordan 881 W. Outer Drive Oak Ridge, Tennessee 37830 Ms. Jean DeJuljak 381 East 272 Euclid, Ohio 44117 Mr. Rick Jagger Industrial Commission State of Ohio 2323 West 5th Avenue
Director of Health 450 East Town Street
- Columbus, Ohio 43216