ML17206A957
| ML17206A957 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Calvert Cliffs, Palisades, Saint Lucie, Arkansas Nuclear, Maine Yankee, Fort Calhoun  |
| Issue date: | 06/05/1979 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.), Maine Yankee, OMAHA PUBLIC POWER DISTRICT |
| References | |
| NUDOCS 7908130359 | |
| Download: ML17206A957 (21) | |
Text
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 June 5,
1979 r~a3C ALL OPERATING COMBUSTION ENGINEERING PLANTS Gentlemen:
We are generically reviewing operating plants using Combustion Engineering (CE) designed reactors in light of the Three Nile Island Unit No.
2 (TMI-2) incident.
Accordingly, the purpose of this letter is to advise you of several concerns involved with this generic review and to request your active participation in resolving these matters.
We have scheduled a meeting with the owners of all operating plants using CE designed reactors.
This meeting is scheduled for 9:30 a.m.,
June 12, 1979, in Room 6507, Maryland National Bank Building, 7735 Old Georgetown
- Road, Bethesda, Maryland.
You are expected to attend this meeting and to be pre-pared to discuss the matters identified below.
(1)
We are preparing a generic report on TNI-2 matters related to operating plants that use CE designed reactors.
Although this report is not yet
- complete, we expect that it will recommend further analyses of transients and small reactor coolant system breaks, the development of appropriate written procedural guidance to operators as indicated by these analyses, and further training of operators in the use of these new procedures.
A clear understanding of the remaining scope of work is expected to be developed during the course of the meeting using the staff's information
, needs shown in Enclosure 1 regarding small break LOCA analyses and models.
(2)
In certain instances, licensees are using fuel which was not provided by CE; therefore, these licensees may be relying on safety analyses which were not provided by CE.
Thus, it is not clear to us what the respective roles of the licensees, CE, the fuel suppliers or other parties'hould be in implementing such activities as the development of small break and transient analyses and the actual system guidance for the preparation of operator procedures.
(See item 1, above)
We therefore need a clear and concise definition of the respective roles of the above parties in these cases.
(3)
As a result of its review of the TMI-2 incident, the ACRS has issued five letters to the Comnission containing recommendations to preclude TMI-2 type events.
CE was previously requested to provide the staff with a concise discussion and its position on each of the recommendatidhs con-tained in the ACRS letters.
By letter dated May 25, 1979, attached heal'eto as Enclosure 2,
CE submitted only a general response which did not address individual recomnendations.
We believe that it would be mutually beneficial for utilities to provide specific comments on those recoranendations having a potential impact on plant design and/or operations.
A summary of the ACRS recommendations is provided in Enclosure 3.
You should be prepared to discuss the procedure and a tentative schedule re-garding the submittal of information needed by the NRC staff to complete its review.
It is clear that a number of technical issues will need to be resolved for plants using CE designed reactors.
The difficulty in performing the necessary work with the limited resources available within the NRC is intensified by the need to conduct similar and concurrent activities with owners of Babcock 8
'Wilcox, Westinghouse and General Electric designed operating plants.
At the same time, there is need to resolve these matters promptly.
To resolve the issues involving CE designed reactors in an expeditious, manner, we believe there is a compelling need to establish an owner's group for CE operating plants.
We expect that such a group would be needed for the remain-der of calendar year 1979.
Since owner's groups have been effective'in the past in resolving generic problems with a minimum of staff and industry re-
- sources, we strongly urge you to meet with other owners of CE operating plants to consider the formation of such a group prior to our meeting on Dune 12, 1979.
This matter will be one of the principal agenda items at that meeting.
Please note that the investigation of a number of areas related to the TMI-2
- accident, including the ACRS recommendations and action items from NUREG-0560, "Staff Report on the Generic Assessment of Feedwater Transients is Pressurized Water Reactors Designed by the Babcock 8 Wilcox Company," will be specifically included as a part of the "Lessons Learned" staff activity.
You can expect additional correspondence in the future on these items.
If you require any clarification of the matters discussed in this letter, please contact I. Villalva, the staff's assigned project manager for these generic activities involving CE designed reactors.
Mr. Villalva may be reached on (301) 492-7745.
Sincerely, Robert W. Reid, Chief Operating Reactors Branch k'4 Division of Operating Reactors
Enclosures:
1.
Small Break LOCA Analysis Information Needs 2.
CE May 25, 1979 General
Response
3.
Summary of ACRS Recommendations cc W'enclosures:
See next page
ENCLOSURE 1
O.
RE EST FOR ADDITIONAL INFORMATION REGARDING SMALL BREAK LOCA ANALYSIS The response of the primary system of a given plant to small break LOCA's will differ greatly depending upon the break size, the location of the break; mode of operation of the reactor coolant
- pumps, numbers of ECCS systems functioning, and the availability of secondary side cooling.
In addition, this res'ponse may differ for different plants designed by the same NSSS vendor, because of differences in loop configuration (f..e.',
2 loops,'-pumps; 3-loops,,3-pumps).
or different ECCS designs.
In order for the staff to complete its eval,uation of the response of currently operating
, c-E PWR designs to postulated h
small break LOCA's, the following information is needed:
{1)
Provide a qualitative description of expected system behavior for {a) a range of postulated small break LOCA's, including the zero break. case, and (b) feedwater-related limiting transients combined with a'tuck-open power operated relief valve.
These cases should include situations where t
1 auxiliary feedwater is both assumed available and not available.
The r
cases considered should also include breaks large enough to {a) depressurize the primary system, (b) maintain the primary system at some intermediate
- pressure, and (c) repressurize the primary system to the'safety and/or relief valve setpoint pressure Various break locations in the primary system, should be considered, including the pressurizer.
(2)
Provide a qualitative descrip%ion of the various natural circulation modes of expected system behavior following a small break LOCA.
Discuss any ways in which natural circulation can be interrupted.
In particular,
'iscuss the applicability of the concerns in t'e Hichelson reports (reports on BSW 205 FA plants and CE System 80 plants) identified in Annex 1 to this Enclosure.
Assess the possible effects of non-.condensjb]e gases contained in the primary system, The following questions pertain to your Small break LOCA analysis methods:
(3)
Demonstrate that your current small break LOCA analysis methods are appropriate for application to each of the cases identified in items (8) through 02)below,.
Thfs demonstration should include an assessment of the adequacy of the pressurizer and steam generator
- noding, and the pressurizer surge line representation.
This may be accomplished by,.verifying the methods with the use of data (e.g.,
comparison with experiments, TMI-2 evaluation).
If, as a result of the above assessment, you modify your analysis methods (e.g., pressurizer and steam generator noding), provide justification for any such modification.
(4)
Verify the break flow model used for each break flow location analyzed in the response to Item (8) below.
(5)
Verify the analytical model used to calculate natural circulation heat removal under two-phase flow conditions.
(6)
Provide justification fo'r your treatment of non-condensible gases following discharge of the safety injection tanks.
(7)
Verify your analytical calculation of fluid level in the reactor pressure vessel for small break LOCA's and feedwater transients.
For each of the.analyses requested in Items (8) through {12} below, (i)
Provide plots of the output parameters specified in Annex 2 to this Enclosure.
Indicate when the pressurizer safety 'and/or relief valves (iii')
(iv) would open.
Include appropriate information about the role of control systems in the course of the transient.
Describe how the system response'ould be affected by control systems.
If the scenario is different for different 'classes of plants
{two-loop, three-loop, four-loop, different ECCS designs),
provide an example of each kind.
(8)
P'rovide. the results of a sample analysis of each type of small break' behavior. discussed in the response to item (1) (e.g., depressurization, pressure hangup, repressurization).
Provide the results of an analysis of the worst break size and location in'erms of core uncovering.
This may be a break which does not result
'in HPI initiation.
This may require more than one calculation.
(10)
Provide the results of a complete analysis of feedwater-related limiting transients combined with a stuck-open power operated relief valve.
These cases should include situations where auxiliary feedwater is both I
assumed available and not available.
(ll)
Provide the results of 'a small break LOCA analysis assuming loss of feedwater and auxiliary feedwater.
The case with the worst break location which affords the least amount of time for operator action should be analyzed.
Single failure of the ECCS should be considered.
(12)
Provide the results of a small break LOCA analysis assuming that one steam generator is lost either due to isolation or due to loss of auxiliary feedwater.
(13)
Provide the results of an analysis of the effect of reactor coolant pump operation (tr ipping all RCP's, keeping all and some RCP's running) on the course of small break l.OCA's.
I (14)
Provide the results of an analysis.of the effects of djfferent HPI termination criteria on the course of small LOCA's.
Specifically, for each small break LOCA analyzed jn response to Item (8)..above,,'compare the effects of the NRC HPI termination criteria (as stated jn ISE Bulletjn Ho.79-063, Xtem 6(b) )
to those for the HPI termjnatjon criteria which C-8 has recommended to licensees with.
G"~ designed operating plants.
Provide plots of significant parameters pf 'jnterest, such as system pressures, temperatures, and subcooling, on a
common tjme axis.
Indicate on the plot when the operator would terminate HPI injection for both sets of criteria.
(15)
Provide a list of transients expected to lift the PORVs; identify the at which two-phase flow discharge would be experienced.
assumed steam and two-phase flow rates through the valves for these transients.
Provide justification for your assumptions, including the time
- , (16) provide guidelines for the preparation of operational procedures for the recov'ery of plants following Small LOCA's.
This should include both short-term and long-term situations and follow through to a stable condition.
The guidelines should include recognition o'f the event,
~ precautions,
- actions, and prohibited actions.
lf RC pump operation is assumed under two-phase conditions, a justification' of pump operability should be provided.
Discuss instrumentation available to the operator'nd any instrumentation that might not be relied upon r
during these events (e.g., pressurizer level).
What would be the effect I
of this instrumentation on automatic protection actions7
ANNEX 1
TVA C. Michelson Concerns 2.
3.
4.
5.
7'.
9.
10.
Pressurizer level is an incorrect measure of primary coolant inventory.
The isolation of small breaks {e.g., letdown line; PORV) not addressed or analyzed.
Pressure boundary damage due to loadings from a) bubble collapse in subcooled liquid and 2) injection of ECC water in steam-filled pipes.
In determining need for steam generators to remove decay heat, consider that break flow enthalpy is not core exit enthalpy.
Are sources of auxiliary feedwater adequate in the event of a delay in cooldown subsequent to a small LOCA?
Is the recirculation mode of operation of the HPSI pumps at hish Pressu<<
ah established design requirement?
Are the HPSI pumps and RHR pumps run simultaneously?
Do they share common piping?/suction?
If so, is the system properly designed to ace'omodate this mode df operation (i.e., are any HPSH requirements violated, etc...?)
Mechanical effects of slug flow on steam generator tubes needs to be addressed.
(transitioning from solid natural circulation to reflux boiling and back to solid natural 'circulation may cause slug flow in the hot leg pipes).
Is there minimum flow protection for the HPSI pumps during the Vecirculating mode of operation?
The effect of the accumulators dumping during small break LOCAs. is not taken.
into account.
11.
12.
13.
What is the impact of continued running of the RC pumps during a small LOCA?
During a small break LOCA io wpich offsite power is lost, the possibility and impact of pump seal damage and leakage has not been evaluated. or-analyzed.
During transitioning from solid natural circulation to reflux,boiling'nd back again, the vessel level will be unkn'own to the operators, and emergency procedures and operator training may be inadequate.
This needs to be addressed and evaluated.
NOTE:
Items 1 through 4 are taken from "Decay Heat Removal During A Very Small Break LOCA for a BRW 205-Fuel Assembly PWR," C. Hichelson, Draft Report, January 1978.
Items 5 through 15 are taken from "Decay Heat Removal Problem Associated with Recovery from a Very Small'reak LOCA for CE System 80 PWR,"
C. Hichelson, Draft Report, Hay 1977.
{continued.next page)
ANNEX l page 2
14.
The effect of non-condensible gas accumulation in the steam generators and its possible disruption of decay heat removal by natural-circulation needs to be addressed.
15.
Oelayed cooldown following a small break LOCA could raise the containment pressure and activate the containment,.spray system.
Impact and consequences need addressing.
ANNEX 2
Plotted Out ut Parameters CL Core:
L, X
T Reactor Vessel; Upper Head:
L, X
Downcomer:
L, X
~Pi in:
HotLeg:
X, T,M,L (Pressurizer Leg)
Cold Leg:
X, T, M, L, MM>>, J MM>>de (Break Leg)
Pressurizer:
Win, Xin, L, X, P, T
Steam Generator:
Primary:
X,
, T, h
Secondary:
P, L, X, T, WREL, WAFW h
Leak:
- PORV, W,
X or
- Break, W, X, Wdt
~PL 1
1:
X.L Nomenclative:
P - Pressure Lx Mixture Level X - Quality Ti
-, Temperature W - Mass Flow Rate 1
h - film heat transfer coefficient HPI - High Pressure Injection REL - Relief Valve AFW - Auxiliary Feedwater M<
C 6 Pcnaqr S~tems
( o~ti, non f:hgino~nng. frig.
le%) prospqcl HillAoprt
<<'inOSO>. COnneclical 06095 Ti.l. 2Q31MS 1931
>culex 9 <i297 ENGLosURE 2
~
~
SYSTEMS Hay 25, 1979 LD-79-0)i Dr. D..F. Ross't'.,
Deputg Director'iv)sion of Project Hanagement Off)ce of Nuclear Reactor Regular)on U-S Nuc1ear Regulatory Caamission Wash(rigton, D. C.
20555'ubject ACRS Rec~i ndytions Relating tq THE-g Accident References'A)
NRC letter from 0. F. Ross, Jr. to A. E. Scherer, dated Ray 17, 1979.
(8}.NRC letter from D. F. Ross; Jr. to A. E. Scherer, received &y 23, Ig79.
Dear Or. Ross:
In response to the subject letters.
you are no doubt aware that CorAbustion Engineering has been active]y involved in evaluating the concerns 1)sted therein and has pa) ticipated in several discussions on these suMects with the HRC staff as well as r'ecent ACRS a;ecting.
R have been concerned, however', that by s)INp)y evaluating the indfvidua> )tens lfstef
$ n the ACRS letters, m auld not adequately assess the total impact of these recoreendations on overall plant safety.'s a resu14 rather than addressing items on a case-by-case
- basks, Ne are addressing theta as part. qf an overall'1 design reviewer pr+rhfa, Csllbust1on Eng)neer)ng
{C-E) has recertHy folitled a task force to evaluate the safety implications for nUclear pomr plant design and operation of the II-P accident.
In parallH, this task force has begun.a revim of the sensitivity of the C-E design to abnormal operating conditions soch as iho 6 Hhfch occurrEM at TNI-2.
The ask force w)ll use the Three H>le Island-2 accident as an in-dication of areas for consideration and will not reStrict itself solely to the considerat)on of TBI-2 specific problems.
Furthermore, the task force is not restricting itself to traditional approaches to safety eva)uaf$ on (such as con-s)der)ng curtly f'ixed scenarios, single failure philosophy. plant protection and recovery us)ng only safety-grade sysums or past design decisions).
Me believe that this effort Mill present the best evaluation of 'W-2 related
- problems, and uf'll consider a11.aspects of the C-E design.
.I
/
Me have, nevertheless, reviewed the specific concerns listed in the enclosures to your letters..
Our kn)tial review has indicated that while many of these recmoendat)ons appear to be desirable with respect to pl.nt safety, the actual implmzntat)on of certain changes should await completion of our engineering evaluation.
On the other hand several of the recommendations are. being iL"ple-meoted.
A natural circulation test
$s already conducted during the power ascen".-)on test program in every C-E test plant and in fact a complete natura1 circulation cooldown ha" been conducted at a C-C plant.
Ccebu iion Engineering has also provided recomen4ations to the utilities operatirg 0-E designed plants in response to IAE Bu11etin 79 958 which you have air eady'een Other recermen4a-
- Qons, such as reactor vessel level indication are under active consideration but vfll require additional evaluation.
Combustion Engineering intends to continue its evaluation of THE-2 r'elated prob-lems-Mhen the evaluation of each individual design change has been completed ez wf)l provide recennenAtions to our overs.
If you should have any questions regarding this fiutter. please fee1 free to contact nie or Hr. R. R. Hi11s, Dr of; ap staff at (2O3)QS-l9$ 1, extension 4738.
"YerI truly yours, C(NaUSrIOH e'CINEmrHG.
rHC.
AES'dag A. E.
herer L)censing Nanhgqr
I ENCLOSURE ACRS RECOMt1ENDAT IONS A.
Letter N. Carbon to Chairman Hendrie dated A ril 7, 1979 Recommendation 1 - Perform fur ther analyses of small break transients and accidents.
Recommendation 2 - Provide operator additional irfor;..ation and means to follow the course of an accident; as a minimum, consider expeditiously:
(a) unambiguous RY level indication (b) remotely controlled vent,cr RCS high points Recommendation 3 - Item 4b of Bulletin 79-05A considered unduly pre-scriptive in view of uncertainties in predicting course of anomalous small bre'ak transients/accidents.
r B.
Letter R. Frale to Commissioners
.dated A ril 18, 1979 Recommendation 1 - Natural Circulation-related Items a.
Detailed analyses of natural circulation
- mode, supported as required by experiment, by licensees and NSSS vendors.
b.
C.
Develop procedures for initiating natural circulation.
Provide operator means for assurance that natural circulation has been established, e.g.,
by install-ation of instructions to indicate flow at low velocities.
d.
Expeditiously sur vey operating Pl'R's to determine whether suitable arrangements of PZR heaters and reliable on-site power distribution can'be provided to assure this vital aspect of natural circulation capability.
e.
Operator should be adequately informed concerning RCS conditions which affect natural circulation capa-bility, cog.,
(1) indication that RCS is approaching saturation condition by suitable d';splay to operator of T
Im T
and P2R press'=
'n "on...'unction witn s~
am tables (2) use of flow exit te:.per:-=.r=- indicator by'fuel assembly ther".iocouples
':. -". e a"ailable.
ACRS Reco;7enda tlons e Recommendation 2 - Thermocouples used to measure fuel assembly eiit temperatures to determine core 1ierfi>>mar>cr
~hou1d
'e
- used, where currently avaiiable, to guide ope>ator concerning core status (full range capability),
Recommendation 3 - Qperat'ing reactors should be given priority regarding definition and implementation of instrumentation to diagnose and follow the course of a serious'accident, including (a) improved sampling procedures under accident conditions (b) improved techniques to provide guidance'o offsite authorities.
Recommendation 4 - Reiterates previous recommendations that high priority be given to "research to improve reactor safety" (a )
research on behavior of LWR's during anomalous transients
\\
(b)
HRC to develop capability to simulate wide range of postulated transients and accident conditions.
Recommendation 5 - Consideration should be given to additional monitoring of ESF equipment status, and to supporting services, to help assure avai'1ability at all times.
C.
Letter M. Carbon to Actin Chairman Gilinsky dated A ril 20, 1979 Recommendation 1 - Initiate immediately a survey of operating procedures for achieving natural circulation, including:
'a) event involving loss of of site power (b) consideration of role of PZR heaters.
ADDITIONAL RECOMMENDATIONS RELATING TO TMI-2 ACCIDENT IN MAY 16 1979 ACRS LETTERS Interim Re ort No.
3 dated Ma 16, 1979 Recommendation 1
Recommendation 2
Recommendation 3
Recommendation 4
Recommendation 5
Examine operator qualifications, training and licensing, and requalification training and test'ing.
Establish formal procedures for the use of LER information:
(a) in training supervisory and maintenance personnel (b) in licensing and requalification of plant operating personnel (c)'n anticipating safety problems Consider formal review of operating procedures for
=
severe transients by inter-disciplinary team, and develop more standardized formats for such procedures, Re-examine comprehensively.the adequacy of design, testing and maintenance of offsite and onsite AC and DC power supplies with emphasis on:
(a) failure modes 8 effects analyses b) more systematic testing of power system reliability c) improved quality assurance and status monitoring of power supply systems Hake a detailed evaluation of current capability to-withstand station blackout, including:
(a) examination of natural circulation capability under such circumstances (b) continuing availability of components needed for long-term cooling under such circumstances (c) potential for improvement in capability to',
survive extended blackout Recommendation 6
Examine a wide range of anomalous transients and degraded accidents which might lead to water hammer, with emphasis on:
(a) controlling or preventing such conditions (b) research to provide a better basis for control or prevention of such conditions
Recommendation 7
Plan and define NRC role in emergencies, including consideration of:
r;: ~> emeroencv
- nlsns, procedures 8
>ss'ur9'ce CIA'c I~i@1 emergency plans',
procedures and organizations are in place (b) designation of emergency technical advisory teams (names and alternates)
(c) compilation of an inventory of equipment and materials needed in unusual conditions or situations Recoranendation 8
Review and revise within three months:
Recommendation 9
Recormendation 't0-(a) licensees'ases for obtaining offsite advice and assistance in emergencies from within and outside company (b) licensees'urrent bases for'otifying and providing information to offsite authorities in emergencies Examine the lessons learned at TMl-2, including con-sideration of the following:
(a) behavior, failure modes, survivability and other aspects of TMI-2 components and systems as part of the long-term recovery process (b) determine if design changes are necessary to facilitate decontamination and recovery of major nuclear power plant systems Expedite resolution of unresolved safety issues by the following means:
(a) suitable studies on a timely basis by licensees to augment NRC staff efforts (b) use of consultant and contractor support by NRC staff Recommendation ll -
Augment expeditously the NRC staff capability to deal with problems in reactor and fuel cycle chemistry in the following areas:
(a) behavior of PWR 8
BWR coolants and other materials under radiation conditions (b) qeneration, handling Il disposal of radiolytic (or other)
H at nuclear facilities (c) performance 3f chemical additives in containment sprays (d) processing and disposal techniques for high and low level radioactive wastes
(e) chemical operations in other parts of.n'uclear fuel cycle (f) chemical treatment operations involved in recovery, decontamination or decommissioning of nuclear facilities Recommendation 12 -
Reconsider whether or not use of the Single Failure
~
Criterion establishes an appropriate level of reliability for reactor safety systems Recommendation 13 -
Mith respect to safety research:
(a) consideration should be given to augmentation of the FY80 NRC safety research budget II
.(b) consider orienting a larger part of the safety research budget toward exploratory (as opposed to confirmatory) research I
Recommendation 14 -
Perform design'studies of a filtered venting or purging option 'for containments for possible use in the event of a serious accident Interim Re ort No.
2 dated Ma 16 1979 Amplified many of the recommendations included in earlier ACRS letters dated April 7, April 18, and April 20, 1979, including ACRS views on relative priorities to be assigned a number of those earlier recomnendations.
(Address amplifications and suggested priority assignments as appropriate.)
~ Florida Power 8 Li Company CC:
Robert Lowenstein, Esquire Lowenstein, Newman, Reis 5 Axelrad 1025 Connecticut
- Avenue, N.W.
Mashington 0.
C.
20036 Norman A. Coll., Esquire HcCarthy, Steel, Hector 8 Oavis 14th Floor, First National Bank Building Hiami, Florida 33131
~ Nr. Jack Shreve Office of the Public Counsel Room 4, Holland Bldg.
Tallahassee, Florida 32304 Indian River Junior College Library 3209 Virginia Avenue Fort Pierce, Florida 33450