ML20215L200
| ML20215L200 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 09/12/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2412, NUDOCS 8610280526 | |
| Download: ML20215L200 (33) | |
Text
___
c!
CERTIFIED MIhUTES
[y f
[
DATE ISSUED: Sept. 12, 1986 Gets-M a
' gggf e
MINUTES OF THE APRIL 2, 1986 g f /
MEETING OF THE ACRS SUBCOMMITTEE ON THE
/e FORT ST. VRAIN NUCLEAR GENERATING STATION gr/ - 2 6 +,
The ACRS Subcommittee on the Fort St. Vrain Nuclear Generating Station (FSV) met on April 2, 1986 at the Visitors' Center at the plant site near Longmont, Colorado.
The purpose of the meeting was to explore technical problems addressed during the recent extended outage, discuss management changes made as a result of the licensee's independent assessment of management controls, discuss regulatory issues and review the status of technical development and support available to Fort St. Vrain.
Notice of this meeting was published in the Federal Register on Wednesday, March 12, 1986 (Attachment A).
A list of persons attending the meeting is attached (Attachment B).
A copy of the schedule of discussions is also attached (Attachment C).
This meeting was planned to span two days but a heavy snowstorm forced cancellation of the second day's sessions.
Material distributed to the Subcommittee during the meeting is listed on Attachment D.
Copies of these documents are on file in the ACRS office.
The entire meeting was open to public attendance.
No written statements were received from members of the public and no member of the public requested an opportunity to make an oral statement.
About five members of the public attended portions of the meeting (one reporter and four State officials).
Attendees ACBS HEC Staff C.P. Siess. Subcom. Chairman H. Berkow, NRR D.A. Ward K.L. Heitner, NRR J.C.
McKinley, Staff C.S. Hinson, NRR P.M. Williams, NRR O.D.T. Lynch, NRR P.
Fortescue, NRR Consultant J.P. Jaudon, Reg. IV R.E.
Ireland, Reg. IV R.E. Farrell, Reg. IV SRI h blic B=rxic2 C2mDany bblic DL Dclorado R.E. Walker. Pres. & CEO M. Hanrahan, Colo. Dept. of H.L. Brey Health J.W. Gahm R.H. Hix, Colo. Office of D.W. Warembourg Consumer Counsel M.H. Holmes W. Wendling, Colo. Public F.J. Novachek Utilities Commission L.M. McBride M. Bender, Colo. Attorney M.E. Niehoff General's Office R.L. Craun H. Dubroff, Denver Post F.W. Tilson T.
Borst 8610280526 860712 2
PDR Ced' N->
1
__________________________J
Fort'St. Vrain Meeting, April 2, 1986 3
4 J. Williams, J. Reesy Kenne8h[4 J.
K. Owens D. Rodgers D. Goss J. Gramling M.J. Ferris T.D. McIntire N.
Linkon T. Prenger K. Reed (see Attachment B for complete list)
Dnening Statem2nt Dr. Siess, Subcommittee Chairman, opened the meeting at 8:30 a.m. on
)
Wednesday, April 2, 1986 with a statement regarding the conduct of the
(
meeting, introducing the Subcommittee members present, confirming the proposed schedule and topics to be addressed, and taking note of the threatening weather.
Introduction Mr. Walker reviewed some of the issues that were considered at the CP review of FSV.. These included secondary containment, fuel particle design, and tritium release.
The main helium circulators were extensively tested at the Vailmont Plant.
In retrospect, more effort should have been spent on testing the auxiliaies to the helium circulators.
He noted that Public Service Compay of Colorado (PSC) had prepared a book of materials for this meeting and asked that everyone have a copy.
This book contained the visual materials for the technical presentations.
1.
BBCJBBB Erezenta1 Lou Mr. K. Heitner, NRR Project Manager, reviewed the history of the transfer of this project to Region IV in December 1982 and then its transfer back to headquarters in October 1985.
Currently NRR has assigned two project engineers to FSV full time due to the large number of unresolved problems and the fact that it is a unique plant.
In addition, NRR has made available about 1/2 million dollars in technical assistance.
Oak Ridge calculates reactor systems performance.
Los Alamos assists with questions relating to fuel performance and materials behavior.
Idaho Engineering Laboratory is assisting with the technical specification upgrade program and the long term fuel surveillance program.
NRR is continuing to review both the fire protection features and the safe shutdown model for FSV.
The fire protection review is complete and draft evaluation has been prepared.
The draft currently grants the exemptions needed to be in compliance with Appendix R except in certain parts of the plant where there are problems.
PSC is making modifications to the plant to bring the balance of the plant into compliance with Appendix R.
NRR is also still trying to resolve problems with the safe shutdown model.
2 hw
Fort St. Vrain Meeting, April 2, 1986 knresponseto'aninquirybyDrSiess,Mr. Heitner explained that the NRCrequiyej,that, in the face of a fire in any other part of the plant, t,Mrpls.", must be capable of being shut down without any damage.
The alternate cooligg method may not be adequate in that fuel temperatures may exceed 2900 F and fuel damage must be assumed.
The second major problem that NRR is addressing is equipment qualification.
Plant operation is currently proceeding under an exemption that will expire May 31, 1986.
PSC faces a number of problems which include lack of retrievable documentation for components that must be quglified and the fact that their steam temperatures approach 1000 F which can produce an even more harsh environment than that postulated ~for the light water reactors.
PSC has proposed an automatic steam line isolation system to partially alleviate the problem.
Mr. Heitner conceded that communications between the licensee and the NRC have not been very effective.
The NRC staff position is that operator action cannot be relied upon in less than 10 minutes and therefore automatic actuation is required.
Some of the older cabling in the plant cannot be positively identified and therefore cannot be qualified by reference to vendor test data.
This continues to be a problem.
2.
HECZHeginn IV Presentation Mr. Jaudon, NRC Region IV, discussed the NRC's perception of the licensee's performance.
The last Systematic Appraisal of Licensee Performance (SALP) report found FSV to be in Category 3 with regard to plant operations.' A new SALP period ends on April 30, 1986 and the NRC Staff was reluctant to make any statements pending issuance of the SALP report.
It was Mr. Jaudon's subjective judgement that morale of the plant personnel had improved.
Recently some security problems have been unco.'ered that may require enforcement action, no specifics were revealed.
3.
Eerformance_Enhancem nt_Eragram Mr. Brey, PSC, described the performance enhancement program that is designed to improve the overall quality, management and operation at FSV in a controlled and timely manner. The program began with six segments but grew to seven.
Mr. Brey described the purpose and progress in each of these segments.
a.
Organisational Concerns - Mr. Walker, President of Public Service Company of Colorado brought the nuclear project (FSV) directly under his personal control.
A new division was created for the sole purpose of handling licensing and nuclear fuel related activities.
Seventy eight new positions were added to the nuclear staff.
Personnel retention problems are being analyned.
b.
Planning and Scheduling - A master plan and schedule were developed as continuing functions.
This provides senior management with a tool with which to set priorities and allocate resources as well as to monitor results.
This is not a new function but it has been strengthened.
c.
Preventive Maintenance - This too is a strengthening of an existing program.
A maintenance planning organization has been established and maintenance procedures are being revised.
3 77
Fort St. Vrain Meeting, April 2, 1986 Post-maintenance testing is being strengthened.
PSC reviewed the NRC report on tpe, Davis Besse experience'and has factored the 18 points brought odt in that report into the FSV program.
About nine technical professionals have been added to the maintenance staff, d.
Upgrade Nuclear Policies and Procedures - Public Service Company of Colorado recognized a need to upgrade procedures in a number of areas including system operating procedures, design related procedures, instrument and control calibration procedures, commitment control procedures, and emergency procedures.
e.
Training - Public Service Company of Colorado intends to
~
make a significant improvement in training.
They are seeking INPO accreditation for both their operator and non-operator training and plan improved training for the support divisions (QA, maintenance, etc.) as well.
f.
Conduct of Operations - This includes standardization of identification of components in the plant, defining responsibilities, upgrading of personnel facilities, and establishment of shelf lives for components and spare parts.
About 80% of the 475 people that support Fort St. Vrain are located at the plant site.
g.
Total Responsibility Management - This is an attempt to substantially increase employee commitment to Company goals at all levels within the PSC organization.
The program is underway; the results remain to be seen.
In an effort to measure the results of the performance enhancement program PSC contracted with the S.M. Stoller Company to evaluate the program.
Stoller concluded that if the program was executed as propc-ed it would improve the conduct of nuclear operations substantially.
However Stoller identified a number of potential problems, one of which was the high turnover rate in nuclear personnel.
Another was the slippage in the implementation schedule.
4.
S.tatus_cL_flant_.0 m ation Mr. Gahm, PSC, reviewed the major events since the last ACRS Subcommittee visit in May 1984.
At that time the plant was restarting after its third refueling, on June 12, 1984 the main generator was t
synchronized to the power system and power generation was resumed.
On June 23 a fault in a pressure relay caused a series of events that led to a reactor scram in which six out of 37 control rod pairs failed to fully insert.
The reactor was able to go to cold shutdown with 31 rod pairs inserted.
The six that failed to go all the way in automatic were driven in manually within 20 minutes.
Investigation into this failure led to the discovery of serious degradation in the control rod drives and in the Reserve Shutdown System.
In addition there was further water ingress into the PCRV and some damage to one or two of the gas circulators.
Later a decision was made to replace certain bolts in each of the four installed circulators.
B.ir June 1985 repairs and refurbishment of the circulators, control rod drives, and reserve shutdown system had been completed and the plant was authorized to operate at up to 15% of rated power pending resolution of the equipment qualification issue.
The reactor was taken critical on July 20, 1985 and on July 23 another moisture ingress event occurred.
After that the plant was limited to 8% power to dry out the core.
On November 7, 1985 the plant was shut down to start some equipment qualification work.
Public Service Company of Colorado had filed a 4
h2r
I Fort St. Vrain Meeting, April 2, 1986 3
request for an extension of the deadline for equipment qualification and one,wa/s' granted through May 31, 1986.
The NRC Staff limited the maximum p6 er level to 35% of rated power as that was the maximum from which shutdown could be achieved using nonqualified equipment.
As of 6:00am the morning of this meeting the reactor was at 12.3% of rated power with the expectation that the generator could be synchronized to the power system later in the day, weather permitting.
5.
Elant LLcensine Mr. M. Holmes, PSC, discussed the status of a number of regulatory issues.
Fort St. Vrain was licensed before 10 CFR 50 Appendix R was made part of the regulations; it met all of the existing requirements at the time of its licensing.
After the imposition of Appendix R, the licensee determined that the only section applicable were III-G and III-J.
Since the requirements of Section III-L were applicable only to water cooled reactors, some corresponding criteria were developed i
specifically for Fort St. Vrain.
For the congested cable areas the criteria were:
means shall be available to shut down and cool down the reactor such that the consequences of DPA-1 are not exceeded, achieve and maintain subcriticality, depressurize through the helium purification system, use the liner cooling system to maintain PCRV integrity and to remove decay heat, maintain certain monitoring and control functions, and provide necessary supporting functions.
In the non-congested. cable areas the criteria were:
shut down and cool down so as to not exceed a fuel particle temperature of 2900 F, no unmonitored releases of primary coolant from the PCRV, achieve and maintain subcriticality, maintain PCRV structural and pressure containing integrity, achieve and maintain forced circulation cooling, maintain certain monitoring and control functions, and provide necessary supporting functions.
Mr. Holmes again pointed out that it would be half a day into a total loss of forced circulation accident that substantial amounts of fuel particle coating failures would occur and fission products would be released into the graphite fuel blocks.
A very small fraction of the I
10 CFR guidelines would escape and reach the public.
It appeared that the allowable consequences of a fire in a non-vital area were more stringent than for a fire in a vital area.
Mr. Heitner (NRC/NRR) agreed to clarify this point at a later date.
Mr. Holmes (PSC) reviewed the history of the equipment qualification issue.
He pointed out that PSC had done extensive equipment qualification testing based on a 30 minute exposure to the harsh env-ironment temperature profile.
The worst case harsh environments were produced by a cold reheat steam line break in the reactor building and a hot reheat line break in the turbine building.
No equipment aging was considered since the required systems were in routine operation, were accessible for routine maintenance and would not be exposed to a high radiation field.
Current estimates, taking into account the recently added automatic isolation sgstem, are that the ambient temperature will return to about 100 F (normal power plant exposure) within about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
PSC is currently discussing with the NRC how I
aging should be considered in these circumstances.
5 W
~
Fort 'st. Vrain Meeting, April 2, 1986 o
9 It was noted that the steam line rupture detection and isolation w en ectuated, will produce a loss of forced circulation accident'p(BBA-1).
- system, The plant has 90 minutes to restore forced circulation using qualified equipment before it has to resort to the alternate (liner) cooling mode.
PSC objects to having to qualify the entire second suit of equipment, not needed for the first 90 minutes, to the harsh environment conditions.
Mr. Holmes described the Technical Specification upgrade program.
The objective is to simplify the Technical Specifications and make sure they are accurate, complete, and consistent with the existing design and the FSAR.
Because Fort S.
Vrain is different from all other U.S.
reactors, PSC must write the Technical Specifications in sufficient detail that a new NRC inspector can come in and understand what is required.
The NRC is trying to incorporate all of the innovations that are being developed at other sites.
Mr. Holmes also reported that the NRC has directed that Fort St. Vrain must comply with all NRC regulations and requirements unless PSC requests and is granted an exemption.
PSC submitted a multi page statement as to why they didn't have oil collection. pans under their reactor coolant pumps.
This has replaced the informal relationship where the project manager was able to say what was applicable and what was not, without documentation.
The lack of documentation has led to problems years after the agreements.
6.
Eublis_. Utility ~ Commission _ Issues Mr. Walker, PSC reviewed the history of Fort St. Vrain with regard to its costs and income to PSC.
He pointed out that the penalty clause in the purchase contract with GA resulted in a cost savings of about
$75 million to the ratepayers for seven years, from 1972 to 1979.
Fort St. Vrain was put into the PSC rate base in 1980 with a condition that it meet a certain capacity factor test.
PSC and the PUC concluded that it met the test but the intervenors did not and took the case to court, it currently resides in both the District and State Supreme Courts.
In November of 1984 the PUC instituted an " incentive"
?
plan for Fort St. Vrain that if Fort St. Vrain doesn't meet certain generating requirements then the amount of money in the rates should be reduced and refunded to the customers.
That has been appealed through the courts and is also in the State Supreme Court.
PSC is awaiting decisions on these cases.
7.
Continuing _ Technical _Suppori_from_GA Currently the Chevron Corporation owns GA Technologies and has put that entity up for sale.
A Denver organization called Blue Hill is negotiating to buy.
The effect on technical support is not clear at' this time.
8.
HTHE_ Development _ hnp_ ort Mr. Walker, PSC, described the formation of Gas Cooled Reactor Associates (GCRA) in 1977-1978.
GCRA together with the Department of Energy and GA worked on the design of a 855 MWe gas cooled plant.
In 1984 the approach was changed to the small (140 MWe) modular concept 6
L~
Fort St. Vrain Meeting, April 2, 1986 that would be inherently safe.
Hopefully the modules could be built i
and instal d on a short time scale and capacity could be increased by simply aWd ng more modules.
In the past, DOE didn't budget any money to support the HTGR and GCRA had to go to Congress to get money put into the DOE budget to support the HTGR.
In the recent past DOE started putting in money voluntarily now with the severe budget restraints there is only $5 million available in FY'87.
Mr. Walker indicated that he intended to continue to lobby Congress for support for HTGR development.
In response to Dr. Siess' question regarding utility interest in liquid metal cooled reactors, Mr. Walker replied that there were no utilities interested at this time but that there was a shell organization, the Advanced Reactor Corporation, formerly the Breeder Reactor Corporation, prepared to coordinate a program if anyone shows interest.
Mr. Walker is working with several utilities to perhaps form an electrical supply company.
With regard to Fort St. Vrain fuel, they have enough fuel on hand or on order to last through 1988.
They have contractual arrangements to assure fuel availability into the 1990s but no absolute fuel supply to the year 2008.
FSV is paying into the spent fuel disposal fund and is transferring ownership of its spent fuel to DOE.
9.
M21stura._li1gress Mr. Warembourg, PSC, reported on the work of the Moisture Ingress Committee that had been set up to attack the problem.
The Committee reviewed all of the events which had resulted in the introduction of moisture into the PCRV.
The Committee tried to develop solutions or modifications that might reasonably be done to eliminate or mitigate the moisture ingress.
This committee functioned until October 23, 1984 when it was replaced by the Fort St. Vrain Improvement Committee which had a broader scope.
Each gas circulator bearing water system is provided with an f
accumulator to provide water when the normal supply is interrupted.
It was discovered that there was no indication in the control room that an accumulator had been triggered.
It was not uncommon for water to be forced up the circulator shaft and into the PCRV when an accumulator was triggered and injected its water at high pressure.
Indicator lights have been installed in the control room.
In addition, the logic and set points for accumulator actuation have been reexamined and tuned to minimize the amount of water that goes up the shaft and into the PCRV.
l The backup bearing water system was reviewed and was interlocked to prevent both it and the normal supply from being valved in at the same time.
The size of the drain line from the high pressure helium / water separator has been increased and rerouted to the top of the bearing water surge tank.
It was also found that a great deal of transient information was not being logged on the control room data logger.
This has been corrected with a computerized data acquisition system that helps reconstruct l
7 p
_ _ ~
Fort'St. Vrain Meeting, April 2, 1986 transient e, vents.
A number of other! actions were also taken (see s
Attachment f'ep)lacement of the current circulators with electric motor A number of other actions are being considered, includink t driven circulators with magnetic bearings or converting the existing circulators to magnetic bearings.
The improvement committee continues to look for ways to reduce the frequency and quantity of moisture ingress.
~
Removal of moisture is accomplished with the helium purification system.
This system's capacity has been improved by the addition of chilled water units on the front end to condense out gross moisture and prevent premature freese-up of the purification train.
Portions of the circulator auxiliaries have been modeled on a simulator developed by PSC.
This simulator has been used for operator training and as a basis for rewriting system operating procedures.
10.
Circulater_ Bolting _Eailurea In early 1985, while reinstalling a refurbished gas circulator, a high strength primary closure bolt failed.
Subsequent investigation established that the failure resulted from stress corrosion cracking.
Further investigation revealed four areas of the circulator where high strength bolts-could be exposed th an environment that could support stress corrosion.
Bolt failure in any of these areas could result in circulator damage but no threat to the health and safety of the public.
Bolts from the four suspect areas of all five circulators were subjected to a number of nondestructive and destructive tests to determine the contaminants and their source.
Some of the cracks that were cbserved were old and probably occurred during original manufacture.
Some cracks were definitely identified with stress corrosion.
It is believed that the chloride constituent came from the reactor fuel as a residue from the fabrication process.
This is believed to have combined with the moisture from the moisture ingress events and oxygen from various sources to attack the bolts.
The corrective action was to replace the high strength steel bolts with Inconel bolts.
11.
Euture_ Gas _ Circulator _ Development Mr. Brey, PSC, described several concepts for alleviating the problems associated with the current gas circulators.
The first was the electric motor driven, hermetically sealed, magnetic bearing machine.
It would require a 5500 horsepower dual winding (Class I electrical) motor.
Such a device is feasible but would cost on the order of $40 million and would present a whole new series of licensing issues.
The second was to modify the existing circulators and replace the water lubricated bearings with magnetic bearings while keeping the steam turbine and the pelton wheel drives.
This does not address the problem with shaft seals.
The third proposal is to modify the existing circulators and install hydrostatic seals.
An initial proposal has been received from Westinghouse.
8 k,
Fort St. Vrain Meeting, April 2, 1986 l
Another idea is'to make major modifications to the current circulators to reduce of, eliminate moisture ingress.
GA Technologies has made a proposaltdloog these lines.
Lastly there is a proposal to make new circulators using some of the basic concepts of the current circulators but doing away with the need for high pressure, high volume water and using a combination of low pressure (15 psi above PCRV pressure) water and magnetic bearings.
PSC has authorized a study of this concept by a multinational group.
PSC is pursuing several studies to try and find a workable solution.
The costs will be many millions with implementation at least five years in the future.
In the meantime, Fort St. Vrain must live with the existing situation.
12.
Contrel_ Bod DriYe_Eailurea _QYerhaul. Modification. and Maintenance Mr. Novachek, PSC, described the events leading up to the June 23, 1984 reactor scram in which six control rod pairs failed to insert following an automatic scram.
They were eventually fully inserted by j
manual control.
The control rod drives for Fort St. Vrain are cable winches housed in the upper portion of the PCRV.
Each winch has two cables which control two articulated control rods.
The control rod drives are hou. sed in cavities in the PCRV and are exposed to primary helium.
Each cavity is cooled by the PCRV liner cooling system.
An electric motor (shim motor) controls the position of the rods through a three stage gear train and a deep grooved hub in which the cables reside.
Scram is~ accomplished by deenergizing the brake and motor and letting the rods drop in by gravity.
In the same cavity and as part of the control assembly there are the orifice valve control drive and the reserve shutdown hoppers.
PSC conducted an investigation to determine the root cause of the control rod malfunctions.
They began by examining the six drives that failed plus four that had not.
The investigation started out rather broadly but eventually focused on the shim motor bearings.
Moisture had at first been suspected but it was eventually concluded that what was observed was normal wear and tear.
A decision was made to refurbish all of the control rod drives and to install temperature monitors at several locations in each.
In addition to the mechanical problems identified, a number of continuing problems with instrumentation were addressed.
Stress corrosion cracking was found in the control rod cables (347 SS) these were replace with Inconel-625 cables.
The refurbishment program began in February 1985 and was completed in June.
Although these components were exposed to the primary reactor coolant, by careful management the total radiation exposure for this effo_t was only 29 man-rem.
The rest of Fort St. Vrain operations added another six man-rem making the total for 1985 35 man-rem compared to light water reactors which average about 473 man-rem per year.
9
Fort St. Vrain Meeting, April 2, 1986 a
The preventive maintenance program for the control rod drives has been 3.
modifiedso['thatonesixthofthedriveswillbechangedoutwith i
refurbish 4 drives at each refueling.
MR. Craun described the unique back EMF signature of each control rod drive in detail and explained how these signatures would be periodically examined for evidence of control rod drive degradation.
Mr. Heitner, NRC/NRR, pointed out that at the time of the June 23 scram certain helium flow lines to the control rod cavities were obstructed and no cooling flow was available.
Individual flow meters 4
have now been installed.
In gddition PSC is currently trying to qualify the rod drives to 300 F.
13.
BeaerYe_ShutdQHn System Mr. McBride, PSC, described the problems found in the reserve shutdown.
system.
In a test one hoper did not discharge all of its boron carbide balls.
About half were held back by a plug caused by agglomeration of some of the balls due to the formation of boric acid crystals.
The balls at either end of the hopper were free but those in the center were cemented together.
There is a significant temperature gradient over the length of the hopper.
It is believed that the 700 F. temperature at the bottom volatilizgd the boric acid out of that region and it condensed in the 500-300 F central region.
New balls were manufactured with much lower leachable boron and these
^
have been placed in all reserve shutdown hoppers.
Performance will be more closely monitored at each refueling.
14.
Enuipment_9ualification Mr. Niehoff, PSC, reviewed some of the history and licensing aspects of the equipment qualification issue.
PSC has developed a master equipment list of all equipment needed to: mitigate an accident, shut down the reactor, maintain shutdown, maintain primary coolant boundary i
integrity, or recover / restore forced circulation cooling.
Since the a
reactor was licensed prior to February 22, 1983 its equipment must be i
qualified at least to the requirements of " Guidelines for Evaluating Environmental Qualifications for Class 1E Electrical Equipment" (DOR Guidelines).
Replacement Equipment must be qualified to 10CFR50.49.
Because ofothe open and vented nature of the reactor and turbine buildings there is no long term pressure or humidity transient and the temperature transient is limited.
The nature of the transients and the reactor limit the radiation exposure to about 400 rads compared to a threshold of damage of about 10,00 rads.
It appears that the temperature transient is the only one of concern.
PSC is proposing a Steam Line Rupture Detection and Isolation System (SLRDIS) to limit the temperature transient, more on this later.
PSC is having trouble qualifying it electrical cable since some of it was purchased from a wholesaler and is not marked as to manufacturer.
PSC is attempting to demonstrate type qualification.
The issue regards cable temperature resistance and deleterious effects of aging.
10 p
i l
$7
Fort'St. Vrain Meeting, April 2, 1986 l
Dr. Siess s,uggested that the NRC Project Managers find out how this ndled for the nuclear units recently reviewed under the issue was/c@ revaluation Program.
Systemat1 Mr. Niehoff identified the SLRDIS as an important part of the qualification effort since it limits the maximum temperature to which the equipment is exposed.
The SLRIDS employs four 200 ft. long temperature detectors per panel.
The system interfaces with the plant protection system and can trip the circulators and introduce a scram.
It also causes valve closures to igolate a steam line break.
The system will be set to alarm at 135 F and to trip if the rate of temperature rise exceeds 55 F/ min.
Using this gystem PSC calculates a peak temperaturg in the turbine building of 360 F and in the reactor building of 377 F.
It was pointed out that the recent steam line break at the Mohave j
fossil fired plant resulted in severe damage in the immediate vicinity of the steam line but even though the control room filled with steam there were no failures in the instrument and control systems.
Detailed information regarding this accident has not yet been published due to ongoing legal actions.
Mr. Niehoff reported that 350 solenoid valves, 50 transmitters, 50 thermocouples,.and 12 electric motors had been replaced simply because they lacked adequate documentation, not because they were found defective.
About 4000 cable splices were replaced with Raychem splices-(bolted.or crimped and covered with heat shrink insulation).
There is a high likelihood of the fire protection system being actuated by a steam line break, therefore a lot of equipment is being protected from moisture intrusion 15.
EleacLGenerator_Inhe_lntecrity (NUBEG-0 Bill Mr. Holmes, PSC, pointed out that the Fort St. Vrain steam generators have two sections.
One that produces high pressure and high temperature steam and another that reheats steam between stages of the y
i turbine.
The' consequences of a rupture in either section are different.
In the high pressure section leakage will be water or steam into the PCRV and the primary helium coolant.
The primary helium is monitored for moisture content and a reactor trip will occur on high moisture level.
No radioactive material would be carried outside the FCRV.
The reheat section is at a lower pressure than the primary helium thus helium and fission products could leak in.
Radiation monitors are provided which initiate a shutdown and steam line isolation.
The amount of helium lost into the steam /feedwater system would not preclude cooldown.
The utility feels that the performance of the steam generators is one of the high points of Fort St. Vrain operations.
Only two leaks have occurred since plant startup.
Both were small and both have been isolated.
Fifteen percent excess capacity was provided in the steam generators to allow for tube failures and plugging.
The two leaks that have occurred are classed as random failures.
11 2
~
'""~
~ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ' _ _ _ _ _ _ _ _ _ _ _ _
Fort St. Vrain Meeting, April 2, 1986 e
Mr. Holmes pointed out the major differences between the Fort St.
Vrain stea9 generators and those for light water reactors.
These include:tihedwater inside the tubes rather than outside, fewer crevices, no horizontal areas of low flow for sludge buildup, thicker tube walls, all volatile chemistry with impurities down in the parts per billion range.
Because of the convoluted tube design and the location inside the PCRV the steam generators are not amenable to visual or NDT examination.
PSC provided a response to the NRC Staff on NUREG-0844 in about
~
mid-1985.
No questions or other feedback has been received from the NRC Staff.
While many of the NRC Staff recommendations are not practicable at Fort St. Vrain some can be adopted in a modified form.
Fort St. Vrain uses full flow polishing demineralizers and deaerators to remove impurities from the feedwater before it enters the steam generators.
Mr. Hinson, NRC, expressed interest in the consequences of boron balls from the reserve shutdown system getting into the steam generators.
Aside from the difficulty of the boron balls finding their way from their channels in the core to the steam generators, there is a plateout probe with metallurgical samples located at the inlets to the steam generators, these are periodically examined for deleterious effects such as carbonisation.
16.
Donclusion The Subcommittee planned to continue this meeting on the following day, April 3, 1986, but a severe snowstorm prevented that session.
No conclusions were reached nor any recommendations made.
NOTE:
A transcript of this meeting is available in the NRC Public Document Room, 1717 H Street N.W.,
Washington, D.C. or can be purchased from Ace-Federal Reporters, Inc. 444 North Capitol Street, Washington D.C. 20001 (202) 347-3700 12 p.
8580 Fedirtl Registir / Vol. 51, No. 48 / Wcdn sdzy, MIrch 12, 1986 / Notic;s s
pmposed NRC quantitative safety 8aats only during those portions of the the status of technical development and
)
which are to be discussed during the meeting when a transcript is being kept, support available to Fort St. Vrain.
mIeting with the NRC Commissioners.
and questions may be asked only by Oral statements may be presented by Procedures for the conduct of and members of the Subcommittee,its members of the public with the p:rticipation in ACRS meetings were consultants, and Staff. Persons desiring concurrence of the Subcommittee published in the Federal Register on to make oral statements should notify Chairman; written statements will be October 2,1985 (50 FR 191) and apply to the ACRS staff member named below as accepted and made available to the this portion of the meeting.
far in advance as is practicable so that Committee. Recordings will be permitted 10:00 A.Af.-f1:30 A.Af.:(Room 1130-appropriate arrangements can be made.
only,during those portions of the H)-Afeeting with NRC Commissioners During the initial portion of the meeting when a transcript is being kept,
.(Open)-Discuss items noted above.
meeting. the Subcommittee, along with and questions may be asked only by in view of the possibility that the any ofits consultants who may be
. members of the Subcommittee,its schedule for ACRS meetings may be present, may exchange preliminary consultants, and Staff. Persons desiring cdjusted as necessary to facilitate the views regarding matters to be to make oral statements should notify c:nduct of the meeting, persons considered during the balance of the the ACRS staff member named below : s planning to attend should check with' the meeting.
far in advance as is practicable so that ACRS Executive Director if such The Subcommittse will then hear appropriate arrangements can be made.
rescheduling would result in major presentations by ard hold discussions Duri"8 the initial portion of the i
inconvemence, with representatives of the NRC Staff, 8
1 8 th Further information regarding topics its consultants, and other interested nY s
ay h to be discussed, whether the meetm, g persons regarding this review.
present, may exchange preliminary 4
hrs been cancelled or rescheduled, the Further information regarding topics views regarding matters to be Chairman's ruling on requests for the to be discussed, whether the meeting considered dunng the bala:tce of the opportunity to present oral statements has been cancelled or rescheduled, the end the time allotted can be obtained by Chairman's ruling on requests for the meeting.
a prepaid telephone call to the ACRS opportunity to present oral statements
%e Subcommittee will then hear Executive Director, Mr. Raymond F.
and the time allotted therefor can be presentations by and hold discussions obtained by a prepaid telephone call to with representatives of the NRC Staff, Fraley (telephone 202/634-3265),
between 8:15 A.M. and 5.00 P.M.
the cognizant ACRS staff member, Mr.
Public Service Company of Colorado.
Paul Boehnert (telphone 202/634-3267) their consultants, and other interested -
5 Dated: March 7,1988 between 8:15 A.M. and 5:00 P.M. Persons persons regarding this review.
Ma C. floyle, planning to attend this meeting are Further information regarding topics Advisory Committee Management Officer.
urged to contact the above named to be discussed, whether the meeting (FR Doc. 86-53e1 nled S-11-46 8.45 aml individual one or two days before the has been cancelled or rescheduled, the
'ama come n"**
scheduled meeting to be advised of any Chairman's ruling on requests for the changes in schedule, etc., which may opportunity to present oral statements have occurred.
and the time allotted therefor can be Advisory Committee on Reactor obtained by a prepaid telephone call to Safeguards; Subcommittee on Dated. March 7,198a the cognizant ACRS staff member, Mr.
Emergency Core Cooling Systems; Morton W. Ubarkin.
John C. McKinley or to Mr. Richard K.
Meeting A ssistant Decutive Directorfor Pmject Major (telephone 202/634-1414) between I
g,yf,,
8:15 A.M. and 5:00 P.M. Persons planning
~ he ACRS Subcommittee'on (m Doc. 86-5392 Filed 3-11-as; E45 am) to attend this meeting are urged to Emergency Core Cooling Systems w.ll i
g same coce nose hold a meeting on March 26,1986. Room contact one of the above named individuals one or two days before the 1046,1717 II Street, NW, Washington, j
DC.
@dvisory Committee on Reactor scheduled meeting to be advised of any To the extent practical the meeting Safeguards; Subcommittee on Fort St.
changes in schedule, etc., which may have occurred.
will be open to public attendance.
Vrain; Meeting However, portions of the meeting may Datedmarch a tesa.
be closed to discuss proprietary
%e ACRS Subcommitee on Fort St.
Monoa W. ubarkin.
infomation related to Westinghouse Vrain will hold a meeting on April 2 and ECCS codes.
3.1986, at the Visitors Center of the Fort Assistant &ccutive DirectorforPro/*t A'"i'"-
ne agenda for the sub}ect meeting St.Vrain Power Plant at 16805 %tR 19%, Platteville, CO.
(FR Doc. 5393 R!ed 3-11-88. 8.45 am) shall be as follows:
Wednesday, Aforth 26,1986-d30
%e entire meeting will be open to
- coes neoe as A.Af. until the conclusion of business.
public attendance.
He Subcommittee will:(1) Continue He agenda for the subject meeting review of the Duke Power Company's shall be as follows:
Advisory Committee on Reactor request to delete use of the ECCS UHI Wednesday, April 2.1986-8.30 A.M.
Safeguards; Subcommittee on system at McGuire, and (2) discuss the until the conclusion of business.
Reliability Assurance; Meeting a
proposed NRR resolution position for
%ursday, April 3,1986-8.30 A.M.
De ACRS Subcommittee on GenericIssue 124 Auxiliary Feedwater until the conclusion of business.
Reliability Assurance will hold a De Subcommittee will explore System Reliability (tentative).
. Oral statements may be presented by technical problems addressed during the meeting on April 1,1986, Room 1046, 1717 H Street, NW, Washington, DC.
members of the public with the recent extended outage, discuss concurrence of the Subcommittee management changes made as a result ne entire meeting will be open to Chairman; written atatements will be of the licensee's independent public attendance, cceepted and made available to the assessment of management controls, ne agenda for tl.e subject meeting Committee. Recordings will be permitted discuss regulatory issues, and review shall be as follows:
WhY hW
ACRS SUBCOMMITTEE MEETING ON FORT ST. VRAIN l
LOCATION LONGMONT, COLORADO DATE APRIL 2-3, 1986 ATTENDANCE LIST NAME AFFILIATION Sirss.
(~,verrre
?
ACKS Succmur rre Carun Wnno. /24utp 8
Si'neounirirr MEMRFR M %r't rr v4x r' ACPS St a ff 171,unh /Lunw en oepr w neurn ith lIv>nc Peler niR c - AM p H eA m k < ~ <h L.
NRc - s u - ss ee 1
fM n. ann, MenA n o C
WRC - RW Ja u c o w, Joh ns P.
Nac-a / v fnrrr// E:fref L'At'- 917, Def
%,as$vc Pka N t2 O c L L 9
[WRKOLu', HER&& i~
l\\i$c-tven L MVGl l b.LL7~
NGC -A/Ri2-SS PB
/d /W500 Okada S.
/th? C
/0//2 - 5S PR L,s K_c>:<. M -
/yr v.-
we.
c N./.S SCfo4,n G+/ Tee 4,oJohi+ s
[.f $h(st~.
'O i.6 n? (l x
l A/. L. E v e PS;c.
v l
i Tha v/ne L c,eut c.
Pse_
oc',0 dew n;c To Rocer Psc w, o ii, s
/sc
',sa x
& 'o
~feery I's c
/
Grreemeur a
ACRS SUBCOMMITTEE MEETING ON FORT ST. VRAIN LOCATIbN LONGMONT, COLORADO DATE APRIL 2-3, 1986 ATTENDANCE LIST NAME AFFILI ATION F r e c t ~ G l e. o n pse jeu.{<
hem w ech, G A fac1b l os [es
/
n/
Mits Geoff Psc_
tancunet
- n. Ha v ~ es es c
$ 4,/./. d>au.J l'E &
h1.3 de*Ad_
Pse AL.+ J L 6K, c/ G-.s..~-(L,u/
LD w Loaauia Lc Puuc 6 c,r, n C.
Mnk &bc
- c. h p r,, 6. <. A 0 f f.> e Mmq Oes Y
- T) r>o s (0 c e e,<
fIC l
hbt hAu m oss
& ar
\\; ~,
sse Yn T, FERM s f5G R.&c,ce-s
('A-8, + A m /
in-a. Munr:
2.se.
H.G.C3twa WPsc Men >6er~
hn, if pn ?o c,r ca..-
sac
&M Laaa swEc k TTAcM P1FNT 0
ACRS SUBCOMMITTEE MEETING ON FORT ST. VRAIN LOCATION LONGMONT, COLORADO DATE APRIL 2-3, 1986 ATTENDANCE LIST PLEASE PRINT:NAME AFFILI ATION 1
f., A1,* I f a a
$ L SY e skL hbC-
$m ?4904est PSc d bUh(2cf f DCL% N l.
f;c i
l 9-
Rev.5 3/28/86 Tentative Schedule Fort St. Vrain Subcommittee Meeting April 2-3, 1986 April 2,1986 8:30am I.
Opening Statement (5 min.)
C.P. Siess, ACRS 8:35 II.
Introduction (10 min.)
R. Walker, PSC 8:45 III. Report by NRC/NRR (30 min.)
K. Heitner, NRR A.
Effects of Deregionalization l
B.
Status of Major Licensing Issues (10CFR50 Appendix R, 10CFR50.49, etc.)
9:15 IV.
Licensee Performance (45 min.)
J. Jaudon, et al, NRC Region IV A.
Inspection Results B.
Enforcement Actions 10:00 V.
PSC Administrative and Management Items R. Walker, PSC 1
A.
Performance Enhancement Program L. Brey, PSC B.
Status of Plant Operations J. Gahm, PSC C.
Status of Regulatory Issues M. Holmes, PSC 1.
Fire Protection (Appendix R) i 2.
Equipment Qualification 12:30 LUNCH 1:00 3.
Technical Specification Upgrade 4.
LWR vs HTGR characteristics D.
Public Utility Commission Issues R. Walker, PSC E.
Continuing Technical Support From GA R. Walker, PSC F.
HTGR Development Support R. Walker, PSC 1.
Gas Cooler Reactor Associates 2.
Department of Energy 2:30 VI.
Technical Issues D. Warembourg, PSC A.
Gas Circulator Issues 1.
Moisture Ingress Control D. Warembourg, PSC
/h77ACNNEN7~
D
2.
Bolting Failures D. Warembours, PSC 3.
Future Gas Circulator Development L. Brey, PSC B.
Control Rod Drive System 1.
Failures, Overhaul, Modifications F. Novachek, PSC and Maintenance 2.
Back EMF Technique to Evaluate R. Craun, PSC Control Rod Drive Performance 3.
Reserve Shutdown Material M. McBride, PSC Change-Out 5:00pm RECESS April 3,1986 8:30am C.
PCRV Tendon Corrosion Problems and R. Craun, PSC Corrective Actions D.
Equipment Qualification M. Niehoff, PSC E.
Steam Generator Tube Integrity M. Holmes, PSC (NUREG-0844)
F.
Masonry Block Walls M. Niehoff, PSC G.
Human Factors related to operations in M. Niehoff, PSC hostile environments (ice vests, etc.)
H.
Fire Protection Actions (Appendix R)
F. Tilson, PSC I.
Others (as may be identified by ACRS members at the time of the meeting) 12:30 VII. Summation 1:00 ADJOURN 2:00pm Plant Tour
/] Trac H HENY 0
DOCUMENTS PROVIDED TO ACRS SUBCOMMITTEE ON FORT ST. VRAIN I
APRIL 2, 1986 1.
Viewgraphs used by Mr. K. Heitner, U. S. Nuclear Regulatory Commission /NRR - 8 pages 2.
Visuals used by Public Service Company of Colorado - about 265 pages l
i ATTACHMENT D i
I
~. _. - _ _ - - _.. _ _. _ _ _ _ _ _ _... _ _. - _ _ _ - - _ _ _ - _. _. _ _ _. _... _. - _ _ _ - _ - _ - - - _ _.. _ _ _.. _. _ -. - _ _, _ - - _. _. -. _ _ _ _.
{
THE FORT ST. VRAIN IMPROVEMENT COMMITTEE WAS F0RMED
~
BY R. F. WALKER ON OCTOBER 23, 1984.
IMPROVEMENT COMMITTEE PURPOSE:
FORMULATE AND REVIEW PROPOSED TECHNICAL IMPROVEMENTS TO ENHANCE THE OPERATION OF FORT ST. VRAIN.
FINANCIAL OR REGULATORY ASPECTS OF POSSIBLE IMPROVEMENTS SHOULD NOT BE A PRIMARY CONSIDERATION.
OUTSIDE EXPERTISE WILL BE UTILIZED AS NECESSARY TO PROVIDE TECHNICAL ASSISTANCE, s
COMPITTEE MEMBERSHIP:
[
R. F. WALKER, CHAIRMAN H. L. BREY
[
J. W. GAHM L. W. SINGLETON D. W. WAREMBOURG l-MOISTURE INGRESS COMMITTEE DISSOLVED WITH ACTIONS ABSORBED BY THE IMPROVEMENT COMMITTEE.
'L t
L SrrecHMeer E lL r
- 4@
amm4A.J.-weh.5L4-4--
4--,4q,w.W-.2A2--
ELa-Sa4--.bhw W-rf
,.h44Mhh mCLAWA mma
_wE_M JA'-2M e h _4J.h.g.w WA4W,
.hrw. w 3A
,.E-.,
-^,mhQ__,_4_g_.u_,,
.5agA44;
,.m-e d
i a
ee e
t
(
)
l J
4 i
r 4
E k
1 5
I l
4 i
I i
I e
Ii I
f I.
h t
I i
p l
4j 3
l l
t I
+
+
t t
}
l t
MOISTURE INGRESS MITIGATING ACTIONS IMPLEMENTED BY MOISTURE INGRESS COMMITTEE 0
INDICATING LIGHTS HAVE BEEN INSTALLED IN THE CONTROL ROOM TO SHOW THE OPERATOR WHEN AN ACCUMULATOR HAS BEEN FIRED, 8
A SEAL-IN CIRCUIT WAS ADDED TO INTERLOCK THE BACK-UP BEARING WATER -2 VALVES WITH THE NORMAL BEARING WATER SUPPLY VALVE AND TO REQUIRE RESET ACTION TO OPEN THE SUPPLY VALVE.
^
8 EVALUATION OF THE ACCUMULATOR FIRING PROGRAM HAS BEEN
~
COMPLETED AND FOUND TO BE SATISFACTORY.
-(
0 SYSTEM 21 INSTRUMENT CALIBRATION FREQUENCY HAS BEEN EVALUATED AND NECESSARY MODIFICATIONS MADE TO THESE PROCEDURES.
8 THE SIZE OF THE DRAIN LINE FROM THE HIGH PRESSURE SEPARATOR HAS BEEN INCREASED TO HANDLE UP TO 20 GALLONS PER MINUTE FLOW RATE.
^
4 THE DRAIN LINE FROM THE HIGH PRESSURE SEPARATOR HAS i{
BEEN REROUTED INTO THE TOP 0F THE BEARING WATER SURGE TANK RATHER THAN INTO THE MAIN DRAIN LINE.
8 THE HELIUM WATER DRAIN LINE FROM THE CIRCULATOR TO THE HIGH PRESSURE SEPARATOR HAS BEEN MODIFIED TO lb ELIMINATE THE LOOP SEAL WHICH PREVIOUSLY EXISTED.
L<
L amenwr E L
- (
8 A TRANSIENT IMPROVEMENT COMMITTEE HAS BEEN ESTABLISHED TO INVESTIGATE ALL SERIOUS PLANT TRANSIENTS AND TO RECOMMEND PLANT MODIFICATIONS WHICH MIGHT ELIMINATE FUTURE TRANSIENTS FROM SIMILAR CONDITIONS.
4 A COMPUTERIZED SYSTEM 21 DATA ACQUISITION SYSTEM WAS DEVELOPED AND PLACED IN SERVICE TO PERMIT BETTER
~
ANALYSIS OF PLANT TRANSIENTS.
(THIS SYSTEM IS BEING EXPANDED AT THE PRESENT TIME).
8 AS AN INTERIM MEASURE, VALVE OPENING B0OSTERS WERE INSTALLED ON THE EXISTING MAIN DRAIN PNEUMATIC VALVES.
4 INSTALL NEW STRAINERS UPSTREAM 0F BUBW FILTERS.
[(
0 INSTALL NEW POSITIONERS ON HIGH PRESSURE SEPARATOR DRAIN VALVES.
0 REPLACE PRESSURE DIFFERENTIAL INSTRUMENT CABLES WITH SHIELDED CABLE.
4 INSTALL ELECTRONIC CONTROLS FOR MAIN DRAIN VALVES.
4 REPLACE BARTON LEVEL INDICATION SYSTEM ON BUFFER HELIUM RECIRCULATORS.
[
t COMPLETE AND ISSUE A MOISTURE INGRESS MANUAL.
L L t L
/?7T8cHMENT E
9 9
i i
n n
M N
r""
F F
F i
F r
^
c
,o
@ Relocate liigh Pressure l
Separator Drain to Top of Bearing Water Surge Tank.
-Vf.=:.:
- - id k
ty
@ High Pressure Separator to surre" Increase Drain Line Frosii 5 *Y l
l HE'ioM sot = p e n DRYER SOP Pt-Y
\\REcindutAToR[
C>:<0 llandle 20 GPM.
O D
@ Eliminate LT and FC "C
i l
ut fll C'RC 1
cinc Feedback to Main Drain.
i I
I '
n "SEPAR ATOR l
w. e.
i*
@ Control liigh Pressure P-5 EPARM08t n 1
I Separator l.eVel l
l l
t 3
l l
Independently.
l I
l l
l I
@ Control Main Drain on hl Cartridge AP Only. Use 1" I
V Q
I g
i h
g 3
Bypass Valve For Control.
i v
v Consider Electronic eso v i
I X - l I---g i
l o@
Controls For Fast Action.
oggio i
l I
l g-) (
i~ ~'i ~
@ Main Drain Must Have Fast
)(-% i
'l d
g_'
b b
@ g_ _ _ g _I_
t, _
i Opening Response.
ig. _gj i
bf
' @ liigh Pressure Separator Eliminate Loop Seal to Mb k
( aumos wane.
M g
l g
D ( SunE TAutC,g n e u n n5 MODIFICATIONS PROPOSED 1
N
'A To Hiif541E MolSTURE INGRESS D1h N
S
~.
ITEMS INVESTIGATED BY
(
MOISTURE INGRESS COMMITTEE AND REJECTED SUGGESTION ACTION PAIR A & C CIRCULAT' ORS AND WOULD REPRESENT SIGNIFICANT
~
B 8 D CIRCULATORS TO REDUCE CHANGES TO CONTROL SYSTEMS.
RISK OF LOSING TWO CIRCULATORS SEPARATION / SEGREGATION / FIRE IN A LOOP.
PROTECTION ISSUES SERIOUSLY
~
IMPACTED.
CHANGE NOT EFFECTIVE FOR OTHER MOISTURE INGRESS SITUATIONS,RECOMMEN-DATION REJECTED.
PROVIDE HIGH PRESSURE CHANGE NOT PHYSICALLY FEASIBLE SEPARATOR TO BUFFER SUPPLY DUE TO CHECK VALVES IN
~
DIFFERENTIAL PRESSURE CIRCULATOR CARTRIDGE.
EXISTING INDICATION TO THE CONTROL DIFFERENTIAL PRESSURE INDICATION ROOM.
BETWEEN PURIFIED HELIUM HEADER AND BEARING WATER SURGE TANK SHOULD PROVIDE ADEQUATE CONTROL ROOM INDICATION.
I INVESTIGATE UTILIZING THE COMPUTERIZED SIMULATION SMALL BY-PASS VALVE AROUND ANALYSIS COMPLETED.
USE OF THE MAIN DRAIN VALVE FOR SMALL VALVE DEGRADES DRAIN
{
CONTROL TO IMPROVE SYSTEM AND CONTROL SYSTEM.
NO RESPONSE.
FURTHER ACTION TAKEN.
"(
lL A?rrownwr E
p
(
ITEMS INVESTIGATED BY
~
MOISTURE INGRESS COMMITTEE AND REJECTED SUGGESTION ACTION REPLACE MAIN DRAIN VALVE HYDRAULIC VALVES HAVE PROVEN WITH A HYDRAULIC VALVE.
TO BE VERY TROUBLESOME.
A DIGITAL VALVE IS BEING INVESTIGATED.
SUGGESTION REJECTED.
MODIFY SYSTEM TO RUN FOR A THIS SUGGESTION WAS REJECTED.
PERIOD OF TIME WITHOUT ADDITIONAL CONTROL SYSTEM
[(
THE BUFFER HELIUM REQUIREMENTS AND POSSIBLE RECIRCULATOR.
BUFFER HELIUM UPSETS SERVE TO COMPLICATE RATHER THAN IMPROVE THE SYSTEM.
L l[
[t L
Jrrmewr E
P
(
ITEMS BEING CONSIDERED BY THE MOISTURE INGRESS COMMITTEE WHICH WERE TURNED OVER TO THE IMPROVEMENT COMMITTEE F~
ITEM STATUS REMOVE TRIP INHIBIT FOR PRESENTLY UNDER EVALUATION.
SECOND CIRCULATOR IN A LOOP.
+.
INSTALL A MOISTURE SLINGER INVESTIGATIONS WERE COMPLETED.
[(
ON THE SHAFT OF THE HELIUM WOULD REQUIRE MAJOR MODIFICATIONS CIRCULATORS TO CIRCUMVENT TO THE CIRCULATOR CARTRIDGES.
LARGE QUANTITIES OF WATER WOULD ONLY BE EFFECTIVE WHEN b
FROM G0ING UP THE SHAFT.
CIRCULATOR IS OPERATING AT RELATIVELY HIGH SPEED.
MAY BE WORTH CONSIDERING IF OTHER CARTRIDGE MODIFICATIONS WERE TO BE MADE.
l-INSTALL DIGITAL VALVES IN A DIGITAL VALVE WAS INSTALLED MAIN DRAIN LINE TO REPLACE ON ONE CIRCULATOR FOR TESTING
{
EXISTING VALVES ALONG WITH PURPOSES. THE VALVE B0UND-UP
- 7 ELECTRONIC CONTROLS FOR AND DID NOT FUNCTION PROPERLY.
L BETTER CONTROL RESPONSE.
THE VALVE WAS REMOVED AND RETURNED TO THE VENDOR FOR ib; FURTHER ENGINEERING EVALUATIONS.
l ELECTRONIC CONTROLS WERE INSTALLED AND ARE IN USE.
kr7'8CHMSk'7* E I
.---7
ITEMS BEING CONSIDERED
(
BY THE M0ISTURE INGRESS COMMITTEE WHICH WERE TURNED OVER TO THE IMPROVEMENT COMMITTEE ITEM STATUS
[
MODIFY CONTROL SYSTEM FOR THE ELE'CTRONIC CONTROL SYSTEM HIGH PRESSURE SEPARATOR THAT WAS INSTALLED PROVIDES AND MAIN DRAIN.
FOR CONTROL EITHER FROM CARTRIDGE DIFFERENTIAL PRESSURE OR WITH HIGH PRESSURE SEPARATOR
~
FEED BACK.
REPLACE BiJFFER HELIUM EVALUATIONS INDICATE THAT THIS RECIRCULATOR WITH AN MAY HAVE SOME ADVANTAGES, BUT EDUCTOR.
ONLY WHEN COMBINED WITH OTHER CHANGES.
~
REPLACE MAIN DRAIN WITH A EVALUATIONS INDICATE THAT
~
FIXED ORIFICE DRAIN SYSTEM WITHOUT OTHER CHANGES A FIXED WITH ASSISTANCE TO HIGH ORIFICE DRAIN SYSTEM WILL NOT PRESSURE SEPARATOR DRAINS FUNCTION ADEQUATELY FOR ALL
[-
BEING PROVIDED WITH A JET MODES OF CIRCULATOR OPERATION.
PUMP.
(I.E., START-UP, SELF-TURBINING, b
AND STEADY STATE).
[.
INSTALL FULL FLOW OR CURRENTLY BEING EVALUATED.
[(
BY-PASS FLOW FILTERS IN BEARING WATER SUPPLY LINES.
jQrpycHMCN7~ E
~
~
~
ITEMS BEING CONSIDERFD I
BY THE MOISTURE INGR$SS COMMITTEE
(
WHICH WERE TURNED' 0VER' TO THE IMPROVEMENT COMMITTEE I
JIEM STATUS
['
EVALUATE USE OF DIGITAL GIVEN EXPERIENCE TO DATE VALVES FOR BEARING WATER.
WITH DIGITAL VALVES, THIS BACK-UP BEARING WATER ITEM IS ON THE BACK BURNER SUPPLY.
UNTIL VENDOR EVALUATIONS OF DIGITAL VALVE DESIGN IS COMPLETED,
[
REPLACE LAMINAR FLOW A RESISTANCE DIFFERENTIAL
[
ELEMENTS IN BUFFER SUPPLY TEMPERATURE TYPE METER WAS LINES.
ORDERED AND INSTALLED ON A
((
TEST BASIS ("D" CIRCul.ATOR SUPPLY).
ADEQUATE CONTROL SYSTEM COORDINATION COULD NOT BE OBTAINED DURING TESTS.
THE METER HAS BEEN REMOVED.
FURTHER ENGINEERING ANALYSES ARE IN PROGRESS, fiPLACE THREE (3) HALF EVALUATIONS INDICATE THAT THE l[
CAPACITY BEARING WATER EMERGENCY DIESEL GENERATORS PUMPS WITH FULL CAPACITY ARE NOT ADEQUATE TO PICK UP rL PUMPS.
THE INCREASED LOAD.
NO FURTHER ACTION ANTICIPATED.
L t
/$y7pcpMEM7 E 1
o 7
(
ITEMS BEING CONSIDERED BY THE MOISTURE INGRESS COMMITTEE WHICH WERE TURNED OVER TO THE IMPROVEMENT COMMITTEE e
ITEM STATUS ELIMINATE CIRCULATOR TRIP EVALUATIONS IN PROGRESS.
ON POSITIVE BUFFER-MID-RESULTS IN PRIMARY COOLANT BUFFER (PRIMARY COOLANT BEING RELEASED T0 THE
[,
FLOWING DOWN THE SHAFT).
REACTOR BUILDING AND SUBSEQUENTLY TO THE
[
ENVIRONMENT.
F 6
I L
L
[
L L<
)?rrgenMENr E i
l
M0ISTURE INGRESS CURRENT ISSUES IMPROVEMENT COMMITTEE lIEM STATUS FLOAT BEARING WATER PRESSURE EVALUATIONS HAVE NOT YET WITH PCRV PRESSURE.
STARTED ON THIS ITEM, ADD AN UNINTERRUPTIBLE MODIFICATIONS WERE MADE WHERE POWER SUPPLY FOR CRITICAL POSSIBLE AND PORTABLE BATTERY SYSTEM 21 COMPONENTS, PACKS WERE PROVIDED Ill OTHER AREAS TO ENSURE AN UNINTER-RUPTIBLE POWER SUPPLY, INVESTIGATE / EVALUATE A PRELIMINARY ENGINEERING UTILIZING A HYDRO-EVALUATION HAS BEEN COMPLETED
[(
STATIS SEAL IN LIEU OF BY WESTINGHOUSE, PRESENTLY THE UPPER LABYRINTH STATIC BEING EVALUATED BY PSC HELIUM SEALS.
ENGINEERING, EVALUATE MODIFYING THE EVALUATION INDICATES THAT A HELIUM CIRCULATOR LOWER STRAIGHT FORWARD MODIFICATION WATER DRAINS.
CAN BE MADE WHICH WILL REDUCE THE AMOUNT OF WATER THAT llEEDS TO BE HANDLED IN THE LOWER DRAIN AREA, THIS MODIFICATION WILL BE CONSIDERED IN THE m
FUTURE AS CIRCULATORS ARE
[
REFURBISHED,
- L.[
arreawarE L
MOISTURE INGRESS CURRENT ISSUES IMPROVEMENT COMMITTEE ITEM STATUS REVISE THE I-02 CONTROL BOARD REVISIONS OF THE CONTROL IN THE CONTROL ROOM TO IMPROVE PANELS HAVE BEEN DESIGNED AS OPERATOR / CONTROL INTERFACE.
A PART OF THE CRDR PROJECT.
INVESTIGATE POSSIBILITY OF A PROPOSAL HAS.BEEN DEVELOPED INSTALLING MOTOR DRIVEN, AND PRESENTED TO THE COMMITTEE.
HERMETICALLY SEALED, THIS PROPOSAL WILL BE CONSIDERED
{(
MAGNETIC BEARING CIRCULATORS.
ALONG WITH THE VARIOUS OTHER ALTERNATIVES.
ECONOMIC
- EVALUATIONS BEING PREPARED.
EPRI INVOLVEMENT TO BE PURSUED.
kO INVESTIGATE THE POSSIBILITY NOT A LEADING CONCEPT.
NO WORK 0F INSTALLING MOTOR DRIVEN BEING DONE CURRENTLY.
OIL BEARING CIRCULATOR.
INVESTIGATE / EVALUATE THE INITIAL ENGINEERING WORK HAS POSSIBILITY UTILIZING HELIUM BEEN RELEASED TO PROTO POWER
[
CIRCULATORS WITH MAGNETIC CORPORATION UNDER A JOINT L
BEARINGS BUT RETAIN STEAM EPRI/PSC PROGRAM.
WATER DRIVE.
! [(
(
9 7 7/p u m o r E
.l
MOISTURE INGRESS CURRENT ISSUES IMPROVEMENT COMMITTEE F
a ITEM STATUS
[~
INVESTIGATE / EVALUATE SYSTEM GA HAS SUBMITTED A PROPOSAL MODIFICATIONS THAT PERMIT WHICH IS CURRENTLY UNDER MAXIMUM USE OF EXISTING EVALUATION, THIS PROPOSAL
~
CIRCULATORS.
INCORPORATES THE FIXED ORIFICE DRAIN, EDUCTORS, JET PUMPS, MODULARIZ5DAUXILIARYUNITS, COMPLETE CIRCULATOR INDEPENDENCE b
WITH THE OBJECTIVE OF ELIMINATINi BACK-UP BEARING WATER, AND ACCUMULATORS AND PROVIDING
[(
A MORE PASSIVE CIRCULATOR AUXILIARY SYSTEM.
EVALUATE SYSTEM 23 (HELIUM CHILLED WATER UNITS HAVE BEEN PURIFICATION SYSTEM) FOR INSTALLED ON THE FRONT-END POSSIBLE IMPROVEMENTS IN COOLER,.
OPERATIONAL EXPERIENCE CAPACITY.
PRESENTLY BEING EVALUATED.
DEVELOP BETTER OPERATOR PORTIONS OF THE HELIUM TRAINING WITH SIMULATOR CIRCULATOR AUXILIARIES HAVE
[
CAPABILITIES.
BEEN PUT INTO A SIMULATOR DEVELOPED BY PSC.
SYSTEM
[
OPERATING PROCEDURES HAVE BEEN REWRITTEN.
l [L i9ruannr E
,L l
L
.