ML19295B572
| ML19295B572 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/15/1968 |
| From: | Grier B, Thornburg H US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19295B571 | List: |
| References | |
| 50-010-68-02, 50-10-68-2, NUDOCS 8010080790 | |
| Download: ML19295B572 (24) | |
Text
.
U. S. ATOMIC ENERGY COMMISSION REGION III DIVISION OF COMPLIANCE Report of Inspection C0 Report No. 10/68-2 Licensee:
Commonwealth Edison Co. (Dresden 1)
License No. DPR-2 Category C Dates of Inspection:
December 27 and 29, 1967, and March 26-28, 1968 Date of Previous Ins etion:
March 4, 1968 I
v 4 Inspected By:
ornbur S
or Reactor Inspector April 15, 1968 fc.W$, n-< >f [irector, Region III Date of Report f
B. H. Grier D
April 11, 1968 Reviewed By:
Proprietary Information:
None SCOPE Routine announced visits were made to the Dresden Nuclear Power Station Unit 1 (630 Mwt Boiling Water Reactor) located near Morris, Illinois, as follows:
1.
On December 27, 1967, by C. E. Jones, Reactor Inspector, and C. D. Hamplecan, Radiation Specialist.
2.
On December 29, 1967, by C. D. Hampleman, Radiation Specialist.
3.
On March 26 and 28, 1968, by H. D. Thornburg, Senior Reactor Inspector, and on March 26 through 28 by C. D. Feierabend, Reactor Inspector.
SUMMARY
Safety Items - No safety items noted.
Noncompliance Items - None.
Soloo8.o77g
. Summary (continued)
Unusual Occurrences - Reactor scrammed from high flux signal (120% on 70 Mwt range) due to binding of the primary feed water control valve.
The resulting reactivity input was ~30c.
The matter was reported in the Dresden Annual Report.
See Section C.
Other Sienificant Items - The status of the primary system inspection was given in CO Report No. 10/68-1.
The control drive maintenance and inspection program, which was approxi-mately 507. complete, indicated that drive seal wear and/or the accumulation of foreign material in the shuttle piston-c ollet region of certain of the drives had contributed to the undesirable drive performance experienced during Cyc le V.
Though the core op increased at very near the predicted rate for Cycle V, considerable crud accumulation was noted on the irradiated fuel. The fuel to be reused in Cycle VI was cleaned and flow tested.
It appears that GE and CE have developed a reasonable method for compensating for progressive core op increases in the calculation of core heat flux variables.
It appears also that the core op increase rate is dependent upon the refuel-ing scheme for each cycle.
Three of the four fuel failures identified at the site were filled with powdered fuel.
Manacement Interview - The management interview was conducted with Messrs.
Hoyt and Redman representing Commonwealth Edison (CE). The following items were discussed:
1.
The writer emphasized the importance of an organized inspection of the core internals, based on experience at other reactor facilities in this country and abroad.
Mr. Hoyt stated that, in his opinion, their efforts to date would have identified any significant failure of core internals. He stated that consider-ation would be given to performing a more systematic inspection during a future outage.
2.
The writer encouraged CE personnel to continue their surveillance and maintenance of the control drives and associated components.
It was emphasized that whenever possible, systematic appraisale of past performance and test data should be used as a basis for surveillance and maintenance programs.
Mr. Hoyt stated that the control drives would receive extensive attention during the next refueling outage.
3.
The writer stated that from his observations of drive performance, it appeared that a massive maintenance program should have been
p.
. Summary (Cont inued) initiated earlier.
Mr. Hoyt stated that drive test results prior to the 1967 refueling outage didn't clearly support the writer's statement in his opinion. He stated that intensified surveillance of the control system was warranted.
4.
The writer; informed Mr. Hoyt that, in his opinion, every effort should be made to understand the nature of the circumferential cracks found in the primary piping. See CO Report 10/68-1.
DETAILS A.
Persons Contacted H. Hoyt, Station Superintendent G. Redman, Assis tant Station Superintendent R. Holyoak, Operations Engineer Mechanical W. Riedaisch, Supervising Engineer, Technical Group G. Diedrich, Engineer, Technical Group L. Butterfield, Jr., Nuclear Engineer, Technical Group H. Bullard, Operations Engineer Electrical (Fire Marshal)
O. Dodd, Maintenance Supervisor F. Morris, Shift Engineer G. M. Watson, Radiation Protection Engineer B.
Administration Mr. Richard Smith, Jr., Instruments Engineer, who was killed in an automobile accident on February 16 will be replaced by Mr. D. Riley, transferred from within Commonwealth Edison.
Mr. Riley is currently undergoing training on the data processing system to be installed in Units 2 and 3.
The functions will be covered, until Mr. Riley arrives on site, by Mr. R. Mef ford, an Engineering Assistant recently promoted from Instrument Mechanic "A".
Mr. R. Holyoak was appointed Operating Engineer, Mechanical, replacing Mr. B. Stephenson who has moved up to Assistant Superintendent of Quad Cities.
Mr. W. Kiedaisch was promoted to Supervising Technical Engineer, replacing Mr. Holyoak.
C.
Operations 1.
Reactor Operations a.
The reactor was shut down on October 11, 1967, at 5:25 p.m.
for modification of the Aa6 TV switchyard modifications and operator licensi The outage was extended for primary system piping ir.'
action. See CO Report No. 10/67-6.
. Details (continued) b.
Recovery from the above outage was initiated on November 4, 1967, but was interrupted by a high flux scram (120% on 70 Mwt range) (Section C.2.).
The recovery was continued later on November 4, 1967, but c.
was again interrupted by a scram from loss of condenser vacuum.
d.
The reactor scrammed again on November 6, 1967, due to loss of condenser vacuum The reactor was shut down on February 2, 1968, for refueling.
c.
2.
High Flux Scram During the recovery from the month long outage in October 1967, the reactor was scrammed from a high neutron flux signal. The following information is pertinent to the event:
a.
The scram occurred at 12:36 p.m. on November 4, 1967.
b.
The reactor was at rated temperature and pressure with 42 control rods withdrawn.
Preparations were being made to roll the turbine.
c.
d.
The steam drum level was low and the operator had opened the motor operated block valve in the primary feed water line.
The operator switched the primary feed water control valve c.
to the manual mode and increased the air load on the valve to open it.
He noted that the valve gave no response to the signal.
f.
While the operator was occupied at the primary feed water control station, the reactor scrammed from high flux on channels 1 to 6.
It appeared that they tripped at 120%
on the 70 Mwt range.
g.
The power level at the time of the scram was ~60 Mwt.
h.
Calculations following the event indicated that the total reactivity input to the system during the event was ~30c.
Personnel at the site stated that the transient occurred in a cime span of 10 to 15 seconds, considering the hydraulic time constant of the feed water, steam drum, and recirculat-ing systems. They consider both values to be conservative.
. Details (continued) 1.
The total primary system temperature transient as a result of the erratic operation of the primary feed water control valve was ~ 40 F in ~5 minutes.
J.
Personnel at the site stated that the primary feed water control valve stuck in the closed position and then released suddenly. They stated that the feed water valve seat was found to be scarred. The valve seat was lapped. Following repair, the valve operated properly.
Operating procedures have been revised to include cycling of the primary feed water valve prior to startup.
D.
Facility Procedures 1.
Site Emercency Plan The Dresden site emergency plan, entitled, " Emergency and Abnormal Procedures," as included in Volume 6, Section B of the Equipment Manual, was reviewed against the objectives and requirements specified in Section 0205, Emergency Plans, Chapter 0200, Facility Procedures of the Inspection Manual (Reactors), dated October 25, 1967, as a guide. The following observations were noted:
a.
Authority and responsibility is clearly established. The Station Superintendent has the responsibility for the administration of the plan. Alternatives are named.
Specific responsibilities for other operating personnel under varying emergencies are established.
b.
Medical assistance is available through company nurse, local doctors, local ambulances, and three hospitals located in Morris, Joliet, and Chicago.
Provisions are made for the immediate notification of the Grundy County Sheriff or State of Illinois Police, if necessary.
Ra d io-logical monitoring and other related assistance is available through the AEC Radiological Assistance Team located at ANL.
c.
The plan provides information for plaaned action regarding the type of incident. Three categorics of emergency are discussed. These are:
(1) Local emergency (2) Site emergency, and (3) Disaster e
e
. Details (continued)
A public address system may be used to inform personnel of conditions, and an evacuation siren is provided for site emergencies or disasters.
d.
Personnel listings are provided in order to contact individuals during off-shift hours.
Signals and communications are deter-mined utilizing the public address system, plan t telephone, and evacuation siren.
e.
Duties of individuals are outlined.
These include the Superintendent, Assistant Superintendent, Shift Engineer, Senior Control Operator, and various other operating groups including radiation protection and security personnel.
Per-sonnel decontamination procedures are outlined.
f.
Site evacuation plans are detailed for all operating groups.
Direction and assembly points are discussed with wind direction considered. Emergency equipment is located exterior to the
- building, g.
Directions are provided to enable personnel to drive to all 16 cnvirons monitoring in the event this survey is necessary.
An environs survey kit including necessary miscellaneous equipment and a ten foot ladder is located exterior to the
' site buildings.
h.
Instructions to the Radiation Protection Group include details pertaining to the gathering of radiation levels data,,
not onl, at the immediate site, but also exterior to the site buildings. The shift engineer is authorized to send survey personnel to downwind of f-site monitoring stations to gather dosimeter readings.
- i. Recovery action plans are outlined recognizing an orderly procedure towards recovery of the plant. These plans con-sist of five progressive steps; such as, 1) evaluation,
- 2) formulation and execution of plans, 3) decontamination,
- 4) caintenance of the plant, and 5) plant operational tests.
These recovery plans include considerations of radiation exposure to personnel.
J.
The " Emergency Plan for Dresden Nuc1 car Power Station" is reviewed periodically for accuracy by Co=nonwealth Edison Company personnel. The current revision was completed in December 1967.
. Details (continued) 2.
Maintenance Procedures g_.
Maintenance procedures were reviewed and discussed with the Maintenance Supervisor, Mr. O. W. Dodd.
The maintenance log and equipment record cards provide a history of all maintenance.
This is supplemented with " maintenance letters" to provide details of major tasks, discuss problems encountered, etc.
Maintenance records appear to be up to date.
3.
Fire Protection Procedures Fire protection procedures were reviewed and discussed with Mr. H. M. Bullard, the station Fire Marshal. The fire inspection records were complete and up to date. Procedures provide for weekly inspections and for more detailed bi-monthly inspections.
Mr. Bullard was aware of the recent fire at San Onofre. He indicated that he has recognized the cable trays as a potential fire hazard and is especially alert for indications of over-heating or of accumulation of trash in trays during his fire inspections.
E.
Primarv Sys tem 1.
Pain Steam Relief Valves This shutdown includes replacement of the five main steam drum relief valves. Valves will be inspected and operationally tested with nitrogen. Any valve that must be disassembled will be tested at temperatures on a boiler in the Joliet Station.
2.
Primary System Piping The status 7 the primary piping inspection was discussed in f
1 the report of the special visit to the facility on March 4, 1968. The primary piping inspection is te.r.porarily suspended until the next quarter as *he pipe insulators have nearly reached the limits of allawable radiation exposure (10 CFR 20).
The inspection will be completed during this outage.
3.
Pressure Vessel NDT The v: iter examined a document from GE which stated that the recommended hydrotest temperature was presently 230 F, based upon estimated current NDI temperature + 60 F.
--1/ CO Report No. 10/68-1
-... =
9
~
8 Details (continued)
F.
Reactivity Control and Core Physics Control Rod Drives Control rod performance Cycle V at Dresden (beginning June 1967) is described in the Dresden Annual Report for 1967 (Section III.A.4.) and in a CO Inquiry Memorandum dated October 2, 1967. A review of plant records indicated that the information contained in the above referenced documents accurately described Cycle V control rod performance at the subject facility. The control rod malfunctions are summarized in the attached Exhibit I.
An additional event which involved the malfunctiog of three drives occurred on October 9, 1967, when the Barksdale Vclve f stuck in the insert position. When drives D-7, E-7, and D-5 were individually selected for withdrawal and operated, each inserted until the selector switch was released. The built-in spare Barksdale valve was then placed into service.
The spare Barksdale valve operated properly. The original valve was repaired and returned to service on October 17, 1967.
Following the series of control drive malfunctions which occurred in August and September of 1967, Region III personnel met with CE per-sonnel to discuss performance criteria for the control drives. This action was deemed necessary on the basis of the performance of that system during Cycle V.
The CE personnel present were informed that, in the opinion of the inspector, control rod performance was becoming ma r gina l. The criteria which were evolved by CE in October 1967 are attached as E>.hibit II.
During October 1967, a program for daily flushing of all drives was instituted which appears to have significantly reduced the incidence of control drive malfunctions. This result suggests that the accumulation of crud in the region of the shuttle piston, which serves to cam the coilet fingers open, may have contributed to the malfunctions observed earlier.
As noted in C0 Report No. 10/67-1 and in a meeting with DRL, several of the pins which serve to attach the rollers to the index tube spud were badly worn. These were found during the inspection of drives which occurred in early 1967. During the course of control drive timed cycle testing which occurred during 1967, 33 drives were identified which dis-played short buf fer times and/or long insert and short withdrawal times.
Review of the test data indicated the following:
2/
- The Barksdale Valve is a 4-way 2-position pressure switching valve, the positioning of which determines whether the control rod will be driven in or out.
. Details (continued) 1.
Ten drives had short buffer times 59nging irom 150 to 250 msec versus a design guide valut of.00 msec. Average buffer time of all drives prior to shutdown was 420 msec.
2.
Drive F-6 had the shortest buffer time of 150 msec.
During Cycle V, a restriction was placed on the drive such that it could not be withdrawn from the reactor until reactor pressure was greater than 500 psig.
3.
Twenty-five drives exhibited long insert times and short with-drawal times.
a.
The long insert times ranged from 46 to 103 sec.
Drives B-3 and J-9 had insert times of 103 sec.
It should be noted that these two drives malfunctioned during Cycle V.
See attached Exhibit I.
b.
The design guide for insert times ranged from 12 to 18 sec.
c.
The short withdrawal ranged from 12.8 to 14 sec.
Drive J-9 had a withdrawal time oi 12.8 sec, I-1 had 15.3 sec, A-7 had 16.7 sec, and F-10 had 13.2 sec.
d.
The design guide for withdrawal times is 18 to 28 sec. The technical specification limit for reactivity insertion rate is 0.0029 ok/sec, which is interpreted by CE to be equivalent to a rod withdrawal time of 12.5 sec.
Personnel state the 12.5 see value is conservative if one assumes a maximum control rod (in pattern) worth of 0.4% Ak/k hot and 0.2%
- cold, c.
No long scram times were noted. The technical specification lirait is 2.5 see for 90% of travel.
Mr. Diedric a and the writer reviewed several of the graphs of perfor-mance data for individual drives. The performance data for two drives which had long insert times (J-9 and F-9) are shown in the attached Figures I and II, respectively. Performance data for two drives which performed properly (G-7 and C-4) are shown in the attached Figures III and IV, respectively.
1! The design guide value was reduced from 300 to 200 msec based on experimental scrams performed at GE.
In one case, 100 scrams were performed at 0 psig with no buffer seals installed with no damage noted to the drive, and 1000 scrams performed at 1000 psig with def ective buf fer seals also with no damage to the drive.
e
10 -
Details (con t inued)
It is interesting to note that the withdraw and inscrt times for drives J-9 and F-9 crossed several months (a year in the casc of T-9) prior to the time that the insert time for the driven increased signifi-cantly. Considering the design of bottom entry drives and location of the drive seals, one would expect the withdrawal time of the drive to decrease and the insert time to increase as the drive seals wear or begin to fail. When the scals wear badly or fail, the resulting driving forces on the index tube piston are appropriately reduced. The driving forces can be increased in the condition by adjusting the orifices in the Asco valves.
It can be concluded that the crossover point of the withdrawal and insert times for each drive can be used as a criterion to indicate the onset of significant drive seal wear. The writer encouraged licensee management to develop a seal wear criterion based on tcst data.
On the basis of the performance history of the control drives, test results as summarized above, the results of the 1967 drive inspection, and a routine maintenance schedule, 40 drives were selected for inspec-tion and repair. The drives selected were those which had a history of malfunctions or poor perf ormance test results. The control drive inspec-tion and maintenance procedure has been established since the major drive modification program which was performed in 1962. The procedure includes the following:
1.
Removal of the drive from the reactor.
2.
Disassembly and visual inspection of the drive.
3.
Dye check of major metal components.
4.
Inspection and/or replacement of all spud drive roller pins.
5.
Gaging of roller diameter and replacement of worn rollers.
6.
Replacement of vorn or cracked diive and buf fer seals.
7.
Reassembly of drive.
8.
Test operation in a test station outside the reactor (does not include scramming).
9.
Reinsta11ation of the drive in the reactor.
10.
Functional testing in the reactor.
9 wo---
m
-e e
. Details (continued)
As of March 26, 1968, the projected drive inspection was ~507.
cceplete. Essentially all of the drives which had malfunctioned er had exh!bited the poorest performance were noted to have been inspected or repaired. The significant results of the inspection as of March 26, 1968, are summarized in attached Exhibit I.
In the case of drives which had drifted out of the reactor, evidence of the presence of foreign material was f ound in the region of the shuttle piston in four out of five instances.
In several instances, flaking of the chrome plating of the guide plug was noted which could have been the source of some of the foreign material which had jammed the shuttic piston.
In all cases where long insert or short withdrawal times were noted, badly worn and/or cracked drive seals were noted.
In almost all cases where short buffer times were indicated, worn buffer seals were found.
Three badly worn roller pins have been found. Sixteen rollers have been found to be worn (3 mils reduction in diameter). All dye checks were negative. Test operation following inspection and maintenance of all drives, as of March 26, 1968, has met specifications.
The above data were obtained by review of licensee inspection and maintenance records and discussions with personnel at the site.
According to personnel at the site, all of the scratches found on the ID of the shuttle piston have been removed by light honing. The guide plugs where the flaking of chrome plating was observed have been repla c ed.
It was also reported that some light scratching of index tubes has been noted and removed by honing.
Mr. Feierabend observed drive inspection and maintenance operations.
The final results of the drive inspection and maintenance will be reviewed and reported upon prior to startup. The performance data ior the uninspected drives will be reviewed in detail prior to startup.
G.
Core and Internals Significant crud deposition was noted on the orifices and bottom tic plates of irradiated fuel elements which were inspected outsida the reactor during the 1967 refueling outage. On the basis of these observa-tions, all of the fuel elements which were to continue in service were mechanically cleaned and flow tested. See C0 Report No. 10/67-1. The matter was reported to DRL by CE.
On the basis of last year's experience and the Cycle V core op trends, it was decided that the fuel elements which were to be returned to service for Cycle VI would be flow tested and mechanically cleaned.
,p eww e sem-m-**
4 4
. Details (continued) 1.
Core tp Trends When the irradiation of Cycle V was initiated in June 1967, the core op was 3 psi with four loops in service.
Core op increased to 6.3 psi in October 1967, when the reactor was shut down for approximately one month. Reactor operation was resumed in Novem-ber 1967 with three loops in service and a core op of 3 psi.
During the next two month periods of three loop operation, the core op increased to 4.6 psi. When four loop operation was resumed on December 10, 1967, the core op was 8 psi. The core op had increased to 9.2 psi when the reactor was shut down early in February 1968.
The average rate of core op increase during Cycle V was 0.7 psi per month, which was slightly greater than the rate predicted prior to Cycle V startup. Review of core op trend records at the site indicated that the core op increase rate was greater by a factor of ~2 for a cycle in which 1/5 of the total number of fuel elements had been replaced by unirradiated fuel.
Characteristically, the Dresden 1 reactor is refuelled in multiples of 1/5 or 2/5 of the total number of fuel elements in the core.
A summary of previous core op trends is shown in the following tabic:
Core op Increase Rates Bv Cvele Approxima t e Fraction of Core Average Core op Cycle Irradiation Replaced By Increase Rate No.
Da t es New Fuel psi / month V
1967 1/5 0.7 IV 1965-1966 2/5 0.4 III 1964-1965 1/5 0.8 II 1963-1964 2/5 0.4 I
1961-1962 0.1 2.
Irradiated Fuel Element Flow Testinn and Cleanine During the current shutdown, the reactor was completely defueled, partially to permit draining of the reactor to facilitate the removal of a large nu-ber of drives for inspection and maintenance.
The fuel elements whic.
<ere to be returned to service in Cycle VI vere flow tested prior to and following mechanical cicaning of the e
N
13 -
Details (continued) fuel element orifices with a long handled brush. The flow data prior to and following cleaning is shown in the table below.
Personnel at the site stated that only five elements had to be recleaned on the basis of the second flow test.
They also stated that visible crud accumulations were noted on the fuel element flow orifices and bottom end fittings. The orifices associated with several fuel elements were noted to be almost completely closed.
Fuel Element Flow At 7 psi Number of Flow Prior to Elements Type of Orifice Cleaning Clean Tested Fuel Designation Min.
Avr.
Max.
Flow
- 74 V
E 146 156 163 166 60 III-F E
135 148 166 166 129 III-B E
136 150 166 166 31 V
D 61 78 96 113 40 III-F D
27 66 85 113 1
I 0
81 81 81 8
I B
44 51 60 57 I
C 26 38 48 1
SA-1 178 178 178 PF-10 64 64 64
- The target for fuel element cleaning was a flow rate within 957. of the specified clean flow.
3.
Cvele V Thermal Hydraulic Limit Adjustments GE personnel used the individual " dirty" fuel element flow mersured during the 1967 shutdown to select the lowest fuel element flows for each type of fuel element in the two flow regions of the core. The highest radial flux weighting factor (1.7 in the inner flow region and 0.99 in the outer flow region) combined with an assumed conservative axial flux shape and the lowest individual fuel element flow for each type of fuel ele-cent in each of the two flow regions were used to calculate the limiting MCFIR at overpower.
The values were recalculated periodically, assuming an average Ap increase of ~0.7 op/ month by CE and GE technical personnel
' ~
and transmitted to the operating personnel.
CE calculations reficcted the most recently measured flux profiles.
.%aw es,ee
+
e e
e.
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. Details (continued)
A cursory review of Cycle V operating data indicated that the observed MCHTR evaluated at 1257. power during Cycle V did not fall below 1.8.
The MCIER's which existed during Cycle V wil' be reviewed in more detail during a subsequent visit to the site.
It appeared that the above described method was adequate tc, cope with cere op increases during the irradiation of Cycle V.
4.
Fuel Sippine Da ta Seventy-four of the 106 Type V fuel elements contained in the core were sipped. No failed elements were found. All of the Type III-B fuel elements contained in the core (190) were sipped.
No failed elements were found. All of the Type III-F fuel elements contained in the core (100) were sipped.
Four leaking Type III-F fuel elements were found. Three of the four contained powdered fuel. The failed elements have not been inspected in the fuel storage as of March 26, 1968. The special elemcats, SA-1 and PF-10, were sipped and were found not to be failed.
5.
Tvpe I Fuel Element Cap Screws It was noted during the 1967 shutdown that cap screws were missing from Type I fuel element lif ting bails. This matter was reported to DRL.
On the basis of the problem noted during the previous outage,
all of the Type I fuel element bails were inspected using a TV ca=cra prior to removal from the core. Only one element was found to have a significant number of cap screws missing from the lif ting bail according to perronnel at the site.
In this case, all of the cap screws (four) on one side of the lifting bail were missing. A special tool was fabricated which engaged the lip inside the fuel element top end fitting. The special tool was attached to the refueling grapple and was used to unload the element in question, according to personnel at the site. They also reported that to their knowledge, no cap screws wcre found to be cissing in addition to those previcusly noted.
It should be noted that the newer fuel elements do not include the use of cap screws for joining the lifting bail to the fuel assembly.
6.
Leadine Pattern for cvele VI The final Cycle VI loading configuration will be performed in the usual " scatter" pattern. The four element basic fuel cell will include enc unirradiated element and three elements selected in descending order of exposure.
Emphasis will be placed on symetrical arrangement of equivalent fuel cells.
.=_w
~~~.-
. Details (continued)
Cycle VI is now planned to be composed of the following fuel:
TypeiII-F
- 96 elements a.
106 elements b.
Type V 163 elements c.
Type III-B 96 elements d.
Type VI 1 element c.
Type I 1 element f.
Type SA-1 1 element g.
Type PF-10 7.
Core Internals Examination Mr. Hoyt stated that a quartz light and periscope was used to examine the bottom of the reactor vessel through the bottom cere support place holes. He stated that no foreign objects were noted. The inspector asked if this was a systematic examination of the individual components of the core internals.
He stated that such an inspection had not been performed since 1962.
P.
Radiation Protection 1.
Personnel Protection Personnel exposure records showed that the highest absorbed dose received by any individual as of December 28, 1967, was 3,520 mrem gamma. The highest quarterly exposure as of Decem-ber 28, 1967, was noted to be 960 mrem. Forms AEC-4 and 5 are maintained and all visitor and contractor exposures are recorded.
R. S. Landauer, Jr. Company processes the film badges.
Administrative controls limit exposures to 50 mr/ day or 300 mr/ week.
Plant superintendent approval is required for expo-sures above this rate.
Supervisory approval is also required when personnel are exposed in radiation fields of 3 r/hr or higher. Work permits are required for all work performed in areas of radiological hazard.
Dosimeters are worn by all personnel and read daily.
-ew -e a
+ * =+
e
. Details (continued)
Whole body counting services are provided by Helgeson Nuclear Services, Inc., of Pleasanton, California, on a quarterly basis with all plant personnel scheduled for counting either quarterly, semi-annually, or annually. These records for the year 1967 were reviewed and showed that no Dresden I personnel had received any internal exposures for that period.
2.
Radiation Protection Personnel Mike Watson is the Radiation Chemical Protection Engineer for Dresden I.
He is assisted by Jim Marshall. The monitoring crew consists of eight persons. Messrs Watson and Marshall report to Mr. Warren E. Kiedaisch, Technical Supervisor, for Dresden I, who formerly was responsible for radiation protection.
Mr. Robert Pavlick, Engineering Assistant, also formerly with the Radiation Protection Group, reports to Mr. Kiedaisch, and both are available to provide technical guidance to the Radiation Chemical Protection Group when necessary.
Mr. Kiedaisch reports to Mr. H. Hoyt, Plant Superintendent.
Q.
Radioactive Waste Systems 1.
Liquid waste is dumped by the batch system.
During the year 1967, a total of 4.319 curies was dumped via the 166,000 gallon per minute condenser cooling water. During 1966, a total of 11.312 curies had been dumped, representing 35.4% of the maximum allowable.
Average concentration during dumping is 30 pc/1. The State of Illinois limits the maximum allowabic concentration during dumping to 100 pc/1 including background which is approximately 12 pc/1.
2.
Stack gaseous activity is monitored daily. These records showed that the present release rate is approximately 15,000 c/sec.
Prior to the January 13, 1967 outage, the release rate was 30,000 pc/sec. Following the tby 29 startup, the release rate was generally around 13,000 to 14,000 c/sec.
A stack particulate and a stack I-131 sample is also collected daily.
Results of the stack 1-131 sample ranged in the order of 0.00027 c/sec. The stack particulate samples ranged in the order of 0.0028 c/sec. During 1967, a total of 256,000 Ci of mixed gasses were discharged via the stack. This amounts to approximately 1.367. of the annual limit. Additionally, 0.0357 Ci of mixed particulates and 0.005 Ci 1-131 were discharged to the stack during 1967, each less than 1% of annual limits.
. Details (continued) 3.
Solid wastes are disposed by commercial waste collection contractors. During 1967, a total of 112 concrete casks cen-taining approximately 30 curies per cask of spent resin were released to California Nucicar at Sheffield, Illinois, for burial. During January 1967, 717 me of waste was released to Nucicar Fuel Services at West Valley, New York, and California Nuclear. During November, 178 me of solid waste was released to California Nuclear.
The licensee had records of all shipments showing complete data pertaining to all shipments.
R.
Environmental Monitoring 1.
Dresden I has the services of Isotopes, Inc., of Westwood, New Jersey, for environmental monitoring which provides data on air particulate, gamma background (including dosimeters and film badges), soil, rain, food crops, surface water, bottom sediments, slime, well water, and milk and grass.
2.
The program does not require sampling of small animals and fish.
3.
Data is presented by gross beta activity and by graphs which are designed to show possible trends.
4.
Comparison of the Isotopes, Inc. data with available PHS data shows no areas of increase. Gaman scans are required and specific isotopes, suspected to be present, are monitored.
I-131 analysis is required of milk and I-131 is also monitored at eight of the seventeen air monitoring stations.
Attachments:
1.
Exhibits I and II 2.
Figures I thru IV 4
IIBIT.I 1
COtiTROL DRIVE ItiSPECTIOi! & PERFORMA! ICE DATA Drive Date of Ocscription of Suspected Cause Results of Coordinates li11func t ion (s )
IL11 func t ion Action Taken of Malfunction Drive Inspection B-3 5/22/67 Drifted Out* - tiotch 6 Disabled drive in f tdly
- 1. Foreign niterial in 1. Flaking chrome &
to tiotch 7.
Later inserted position.
collet & shuttle chrome flakes on.
drifted all of the way Reset Asco valves, piston.
guide plug (Inter-out of core.
flushed & scrammed
- 2. Worn or broken drive ference with drive. Returned to
- scals, shuttle piston service.
indicated).
- 2. Worn drive &
buffe, seals.
- 3. Worn roller pin (probably hadn't contributed to drive perform 1nce)
A-6 9/19/67 Drifted Out Flushed & cxercised
- 1. Foreign material
- 1. Scra tches on ID Returned to service in collet &
of shuttic piston.
9/19/67. Flushed shuttle piston.
Indicates presence again 9/20/67.
of foreign atteria 11/5/67 Fail 500 psig.
- Rods driftcel out individually when selected. Appears impossibic for more than one rod to drift out simultaneously.
4
October 10, 1967 EXHIBIT II CRITERIA TOR CONTINUED OPERATION AS RE1ATED TO DRIFTING - OUT CONTROL ROD DRIVES The procedures given below are applicable to Dresden Unit 1 operation, reflect the requirements of License DPR-2 and consider possible imminent failures of control rod drives.
It is recognized that in some respects these criteria may be more restrictive than a liberal interpre-tation of the license. However, these criteria do not replace the need to evaluate other factors such as shutdown margin requirements or other drive malfunctions.
Definitions
" Drifting-Out" - Continued movement of a drive out of the core beyond latching positions when the drive is not simultaneously being actuated.
Drifting out may occur spontaneously or following drive actuation in the
" insert" or " withdraw" mode.
" Deactivate" - Prevention of further rod movement by disconnecting Asco solenoid lead wires (disarmed) or by valving the control rod out of service.
Drive Condition and Action Reauired 1.
One drive exhibits drif ting - out behavior af ter actuation.
under all operating modes (including insert, withdraw and scram).
If license shutdown requirements can be met, operation may be continued with the drive valved out so as to deactivate the drive.
2.
A second drive exhibits drif ting - o at behavior af ter actuation under all operating modes (including insert, withdraw and scram).
Shutdown the reactor and evaluate the cause and significance. Take safe course of action.
3.
An accumulation of less than six drives exhibit drifting - out behavior af ter actur tion under normal insert or withdraw modes, but remains fully inserted at Position 0 following each scram.
Continue reactor operation with the drive Asco 1 cads pulled (disarmed) so as to deactivate the drive.
EXFTPIT II
. 2 4.
An accumulation of six drives which are deactivated for either or both reasons listed above (drifting - out behavior in which drives may or may not be retained in the inserted position after scram).
Shutdown the reactor and evaluate the cause and significance. Take safe course of action.
5.
Less than six drives in one test series (such as weekly instrument response tests or control rnd worth tests, etc.)
exhibit drifting - out behavior af ter actuation by normal insert or withdraw modes, but which are corrective by exer-cising or flushing.
Continue operation.
6.
Six drives in one test series exhibit drifting - out behavior af ter actuation by normal insert or withdraw modes, but which are corrective by exercising or flushing.
Shutdown the reactor and evaluate cause and significance. Take safe course of action.
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