ML19295B575
| ML19295B575 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/23/1967 |
| From: | Vorderbrueggen US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19295B574 | List: |
| References | |
| 50-010-67-05, 50-10-67-5, NUDOCS 8010080792 | |
| Download: ML19295B575 (11) | |
Text
{{#Wiki_filter:s / U. S. ATOMIC ENERGY COMMISSION REGION III ( DIVISION OF COMPLIANCE August 23, 1967 CO REPORT NO. 10/67-5
Title:
COMMONWEALTH EDISON COMPANY (DRESDEN 1) LICENSE NO. DPR-2 q w C & d i _ ~, 1967 Date of Visit: July 18 Reactor Inspector By: L. E. Vorderbrueg >.,
SUMMARY
The scheduled plant outage for partial core refueling and system inspection was completed and power operation of the Dresden 1 unit was resumed on May 29, 1967. Full power was attained in mid-June af ter the calibration requirements had been satisfied for the newly installed in-core neutron detectors. Stack effluent activity has not exceeded 15,000 pc/sec at rated full power condi-tions. One control rod drive mechanism was deactivated in the fully inserted position af ter remedial measures failed to correct an outward drif ting condition during the subcritical testing phase of reactor restart. W.th this one exception, rod drive perf ormance has been satisfactory. A Core pressure drop with full reqqiculation flow was noted to be substantially lower than the expected value of 7.7 psig when plant operation was resumed. Crud buildup in the primary syste= is being closely monitored by analyzing, continuously recorded core pressure drop data. Littic progress has been made in Commonwealth Edison Company's program for performing a technical evaluation of the suspected contributory causes of the notorious cracks in the small diameter primary system piping. To date, no re-currence of Icakage from piping cracks has been detected. An electrochemical corrosion exposure facility has been installed in one of the four recirculation loops for the purpose of obtaining 1cng-term rate / time corrosion data on various austenitic materials. The facility is sponsored by the General Electric Company and provides for studying the corrosion behavior cf etainless steels in an operating BWR coolant stream which contiins radio-lytic decomposition products and miscellaneous crud and corrosion products. No safety items or items of noncompliance were noted. DETAILS I. Scope of Visit A visit was made to the Dresden Nuc1 car Power Station on July 18, 1967 for the purpose of performing a routine announced inspcction of the Dresden 1 c ontinued -
. Scope of Visit (continued) reacter facility. Mr. C. E. Jones, Reactor Inspector, Region III accompanied the writer in order to become more familiar with the facility so as to assume t.c inspection respontibility. Reviews were made of various operating logs, records and reports, and discussions were held with the following persons: Mr. H. Hoyt, Station Superintendent Mr. G. Redman, Assistant Station Superintendent Mr. H. Bullard, Operating Engineer, Electrical Mr. R. Holyoak, Supervisor, Technical Group Mr. L. Butterficid, Junior Nucicar Engineer, Technical Group Mr. M. Watson, Radiation Protection Engineer, Technical Group II. Results of Visit A. Administration Two changes of significance have been made in the administrative structure of Dresden 1. Effective June 26, 1967, Mr. G. Rednin was appointed to the position of Assistant Station Superintendent replacing Mr. C. Zitek, who was r,amed Director of Advanced Reactor Research. Mr. Red man a t one time was Operating Engineer, Mechanical at Dresden 1, and most recently was attached to the staff of the Superintendent of Generating Stations in Commonwealth Edison Cempany's main office in downtown Chicago. Mr. H. H. ILibermeyer was promoted from Senior Control Operator to the position of Shift Engineer on July 3, 1967. Mr. Hoyt stated that Mr. Haberacycr would not be delegated the duties and responsibilities of the Shift Engineer until he satisfactorily com-pleted the examination for Senior Reactor Operator, which was scheduled for early August, 1967. B. Operations Summary Sug to the successful performance of the previcusly reported hydro tests yequent en the reactor and primary system, final preparations were made ano the reactor was made critical on May 21, 1967. Following a training per-iod for all operating crews, the plant was returned to service. Turbine oper-ation, however, was delayed an additional five days due to the need for an in-spec tion of the generator windings. This was necessary to verify that no dam-; age had resulted from the " bump" which the unit experienced when the generator was closed in on the system grid while slightly out of synchronism due to the undetected failure of a potential transf ormer. Full power was finally at-tained in mid-June af ter the new in-core neutron detectors had been fully cal'- ibrated. Plant operating highlights are as follows: - continued - 1/ - C0 Report No. 10/67-4
. Results of Visit (continued) Reactor critical at 8:23 p.m. following refueling in-May 21, 1967 spection and maintenance outage; 150-second period with 29 rods withdrawn. Temperature coefficient checks at 120, 250 and 370 F. May 23, 1967 Turbine placed on line at 7:00 p.m. but disconnected at 8:31 p.m. to correct faulty generator instrumenta-t ion on pha s e "A". Reactor suberitical at 9:45 a.m. and shut down at May 24, 1967 3:00 p.m. May 25-27, 1967 - Generator windings inspected and found to be undam-aged. Reactor critical at 7:10 a.m. Turbine on system at May 28, 1967 9:06 p.m. at 25% load. Turbine off system at 1:13 a.m. for overspeed test. Iby 29, 1967 Normal overspeed trip at 1900 rpm. Backup trip set at 1910 rpm. Emergency overspeed trip set at 1920 rpm. Turbine back on system at 1:51 a.m. with power limited to 110 Mwe (507.) until in-core nuclear chan-nels are calibrated. June 16, 1967 - Maximum capability check performed: 212.5 Mwe at 695 Mwt. Turbine off system at 5:18 p.m. to repair leak in "C" July 5, 1967 primary feed water heater. Reactor subcritical at 5:33 p.m. Reactor critical at 5:04 a.m. and turbine back on July 9,1967 line at 3:29 p.m. Load increased to 210 Mwe. C. Cvele No. 5 Startup During and subsequent to the core loading for cycle 5 operation, many individual checks were nide to locate the strengest control rod and determine the total shutdown margin. According to Mr. Butterfield, the strongest red was found to be D-10 located on the periphery of the redded region of the The shutdown margin with this rod assumed to be noninsertable was cal-core. culated to be 1.7 + 0.27.. Moderator temperature coefficient determinations were made during re-actor restart at averagc coolant temperatures of ~115,120, 250 and 370 F. - continued - ewee
_4 Results of Visit (continued) i Reactor power was used to increase coolant temperature and reactivity measure-cents were made at the various temperatures using a positive period frcm four decades below a power 1cvel of ~25 Mwt. The data show that the moderator co-ef ficient changes from positive to negative below 115 F with the extrapolation 0 indicating the transition to occur slightly below 100 F. This is consistent with the values measured at the beginning of previous cycles, and, according to Mr. Butterficid, indicates that the license requirement (that the coeffi-cient be negative at operating temperature at any time during the cycle) is amply met. Mr. Butterfield was asked what the predicted critical rod configura-tien was f or cycle 5 startup. This prediction is a requirement of the Techni-cal Specifications. He stated that the predicted critical rod position for the 5-B withdrawal pattern being used was 32 + 4 rods withdrawn at 70 F, and 29 rods were required for attaining the criticality. He po'.nted out during the discussion that the dif ference between the actual and predicted critical rod positions was primarily due to the uncertainties associated with the un-evenly distributed " burnable" gadolinia in the Type V fuel used for refueling the core. Mr. Butterfield also stated that, as required by the Technical Speci-fications, adequate sensitivity of the in-core flux monitoring syrtem was de-constrated during the cycle 5 startup activities. This system had been re-equipped with new detectors during the recent refueling outage. At the time of the visit, three of the 16 channels were bypassed due to " jittery" response; otherwise the system had been working well, according to site per-sonnel. D. Control Rod Drives On May 22, 1967, during che performance of criticality checks with the reactor suberitical, control rod B-3 was withdrawn to position 6 in singu-lar increment steps. It then continued to move upward to position 7 without further manipulation of the rod control switch. The operator reinserted the rod to position 6 whereupon it drifted out to position 7. At this point all rods were completely inserted. B-3 continued to drift out from position 0 to the fully withdrawn position (12). Several unsuccessful attempts were made to lock the drive in place in the full-in position. Flushing procedyres using elevated operating pressures were also employed to no avail. The crive was finally scrammed, at which time it locked and was deactivatcd both hydraulic-ally and electrically. The malfunction was attributed to an accumulation of foreign material between the collet and shuttic piston, according to Mr. Redman. Rod timing tests, performed on Fby 27 and July 6,1967, indicated no insert or withdrawal times significantly out of the ordinary. No particular - continued - e-A*so--ow w+ eM
. Results of Visit (continued) problems were evident, although 3 drives displayed insert time in execss of 70 seconds. This is indicative of worn seals and suggests a need for orifice ad-justment, according to Mr. Redran. He stated that, generally speaking,' drive performance to date has been good. E. Fo110 wen on Pipinn Cracks Section II.F of CO Report No. 10/67-4 listed five items related to the investigation of the small diameter piping cracks at Dresden 1 which had not been fully examined as of the time of that report (May 31, 1967). A dis-cussion was held with Messrs. Hoyt and Redman regarding these followup activi-ties which were intended to help ascertain the cause(s) of the cracking that was experienced. The following comments are pertinent to this continuing in-vestigation of piping cracks as of the time of this inspection visit. 1. Cracked specimens scheduled for microprobe analysis by Eattelle Memorial Institute-Columbus have not yet been shipped frem the Dresden site. 2. No report has yet been received from General Electric Company discussing the findings of their metallurgraphical studies of cracked pipe specimens. 3. Site perronnel appeared to be uninformed about the certified chemistry and physical properties of representative specimens of the small diameter seamless tubing which should have been deter-mined at the tLac the plant was constructed. G-E and B&'w' were supposedly attempting to locate these records. The site person-nel stated that to their knowledge no results of this search have been forthcoming. 4. Fk. Hoyt was unaware of any effort that had been devoted to the reexamination of all available radiographs of primary system velds. In summary, CE Company has little progress to report in their er. dea-vor to determine to a reasonable degree of certainty the underlying causes of the small diameter pipe cracking phenomena, F. Specimen Exposure Facilitics 1. General Discussion As' a result of the small diameter piping failures at Dresden 1, the long-term corrosion behavior of type 304 stainless steel in a nuclear environment has received considerable attention. The - continued -
6-Results of Visit (c on t inued ) General Electric Company has suggested using the Dresden 1 reac-tor environment for studies of basic corrosion behavior for the following reasons : a. The failure mechanism or mechanisms are definitely not fully understood at the present time; b. Attempts in comprehending failure specimen characteristics by post-event examination and analysis have been largely incon-clusive; and, c. Simulation testing experience has also been inconclusive, if not disappointing. Additionally, since little detailed knowledge has been gained in relation to crud formation and deposition on reactor components, such studies also are considered appropriate at the Dresden 1 fa-cility. To these ends, exposure loop facilities have been pro-posed as follows: a. An electrochemical corrosion facility for obtaining rate / time corrosion data over a period of two years on type 304 and al-ternate austenitic materials. Exposure to the influence of accumulative corrosion products, radiolytic decomposition products, and miscellaneous crud (all of which are normally encountered in a reactor coolant stream) is expected to pro-vide the corrosive media, b. A water chemistry test loop for studying the nature and pos-sible control of crud accumulation and deposition on various devices representative of reactor components. According to Mr. Iloyt, the water chemistry exposure facility is still undergoing review by the Safety Review Committee. 2. Electrochemical Corrosion Exposure Facility The electrochemical corrosion exposure facility was approved by the Conmittee and installed in recirculation loop "D" curing the recent refueling outage. It is made up of cylindrical test elec-trodes of various austenitic materials mounted on a special flange and fitted into the 4" decontamination stub on the 6" by-pass line around the 18" isolation valve. The test electrodes are fitted through the flange using special insulators so as to maintain electrical isolation from the main piping structure. - continued -
. Results of Visit (continued) Mr. Hoyt stated that the decontamination stub in one of the re-circulation 1 cops was specifically chosen because of its pr cx im-ity to regions where failures had occurrec, even though it is possible that in this position the coolant environment may be stagnant,or nearly so. Mr. Hoyt was asked what preliminary ac-ceptance testing had been performed on the equipment before in-sta11ation at the Dresden site. He rcplied that General Electric Company had informed them that the equipment was considered ac-ceptabic for installation on the basis of the following: a. All materials used in the assembly and which will be expoced to the reactor coolant were laboratory tested and found to be completely inert to high temperature water over extended per-iods of time, b. Autoclave tests up to 1800 psig have proven the mechanical stability of the high pressure Conax fittings more than ade-quate. c. The complete assembly was tested in flowing water loops at the Sar Jose laboratory, d. A precautionary feature of the design provides for a perfor-ared screen surrounding all the specimen electrode to pre-vent loss of an electrode to the main flow stream in the un-likely event of a specimen falling off. According to Mr. Hoyt, General Electric Company believes that the use of electrochemical techniques for obtaining corrosion data both in and out of a reactor environment is a new concept in this fie:d of study. They claim that this technique offers the unique opportunity to obtain thermodynamic and kneti'esdata on corroding materials by remote instrumentation in~aTeas where safety hazards would preclude the use of any other technique. They also con-clude that the electrochemical technique presents no harmful ef-f ects on the system recirculation piping materials. The tech-nique proposes to determine the oxidizing or corresion potential of the Dresden coolant in an area adjacent to where cracks have occurred. A simple high impedence potentriometric measuring technique is employed to obtain the data and the ef fcct of stress sensitization heat treata:nt and alloy composition will be stud-ied. Static anodic polarization will also be carried out on cer-tain of the exposed specimens. General Electric Company expects the work to result in the following conclusions: a. Whether or not the reactor pipe failures are due to a basic - continued -
. Results of Visit (continued) incompatibility between the materials and the environment or due to some other cause such as fatigue, systen design, poor machining or improper c1 caning procedures or other fabrica-tion factors. b. Whether or not such failures are due to corrosion (stress as-sisted or otherwise) and if they can be inhibited by electro-chemical or alloy development techniques. c. Whether or not a correlation between in-pile corrosion behav-ior and that obtained in out-of-pile simulated environments can be found. Such data would then provide the basis for further alley develop-ment for the production of higher corrosien resistence. / ~" 3. Water Chemistrv Test Loop The water chemistry test loop proposes two parallel runs of piping cf type 304 or 316 stainless steel. The piping is interconnected and arranged so that either reactor recirculating water or feed water may be routed through the loop since the deposits can orig-inate from either source. The main piping run is 2" schedule 60 pipe and includes a coupon section, a flow monitoring scctien, and a test chamber in series. Stacks of coupons or long pieces may be exposed in the coupon section. The test chamber is a piece of 6" schedule 80 pipe with approximately 20" of usabic, test length for testing scaled models of fuel bundle entry pieces or rod bundles with spacers. The smaller branch piping run is schedule 80 pipe and is provided for testing a heat transfer sur-face. Automatic provisions are included to control the flow of water to an electrically heated rod. A piece 5/8" in diameter and a heated length of 24" can be accommodated. An open tank with a small positive displacement feed pump is included for the addition of dilute chemicals when desired. The assembly is pro-posed to be located at the 517'6" elevation with four main con-necting lines with appropriate double valving connectir.g the pri-mary feed water line, a secondary recirculation loop, and the in-let to the cicanup systen.. A drain line to the equipment drain is also provided. As previously mentioned, the Safety Committee has not yet approved installation of the water chemistry test loop. G. Containment Intecrity Verification A brief discussion was held with Mr. licyt in regard to centainnent - continued -
9 Results of Visit (continued) Icak rate tcsting. One page 10 of the Technical Specific 1 ions it is stated that following a study of the results of the first leak rate test and discus-sions of these results with the AEC, a future program of testing would be agreed to. Mr. Hoyt van asked if such a testing program had ever been devel-oped and, if not, what was its current status. He replied that censiderable effort had been devoted to the development of a realirtic and acceptable leak rate testing program and that considerabic discussion of this subject had been undcrtaken with DRL. This situation has not yet been resolved and at the pre-sent time the Dresden 1 operating license contains no specific containment Icak rate testing requirement. H. Health Physics The plant personnel exposure records and the area smear survey re- 'ords were examined and it was evident that no significant exposures or con-tamination prob 1 cms have been experienced. It was also evident that exposures were s omewhat elevated for the operations and maintenance personnel who were involved in the extended piping inspectien program. The maximum exposure re-ceived by a single individual through July 1,1967 was 2.73 rem. Following reactor restart for cycle 5, the condenser off-gas activity appeared to increase with power level and has stabilized at 15,000 c/sec at reactor full rated powcr. Mr. Hoyt stated that this value was somewhat higher than they had expected; however, similarly high activity rates have been seen following restart from previous refueling outages. He stated furthe. that the fact that the activity rate had stabilized icnt a note of optimism to the plant operation. A discussion was conducted with Mr. Watson regarding containment building airborne activity since reactor restart. He stated that the exper-icnce to date had been quite satisfactory with no occasions of activity high enough to require the wearing of filter masks. He showed the inspector the records of measured airborne contamination within the sphere and it was noted that the values were generally in the neighborhood of 5 to 7x10-12 c/cc total beta-gamma activity. At the time of the visit, the activity was running ap-proximately 1x10~11 c/cc. This is a factor of 30 below that which rcquires the wearing of half-masks. I. Miscellaneous 1. Core Pressure Drop and Crud Buildun According to Mr. Redman, the pressure drop across the core was considerably below the calculated 7.7 psig value when the plant was restarted subsequent to the refueling outacc. He stated that he could not recall the exact measured value, but that it was ap- - continued -
. Results of Visit (continued) proximately 6.5 psig and at the present time it was still below 7.0 psig. He stated further that the crud buildup on the core components does not appear to be proceeding; at a high rate as in-dicated by the core pressure drop rate of increase. Mr. F,cdnan stated that they were exercising close surveillance over the core pressure drop due to its relatienship to critical heat flux val-ucs. The data is recorded on a continuous basis and is plotted to show time related trends. According to Mr. Eutterfield, a wide safety margin exists in the calculated values of burnout ratio at reactor full power conditions. He stated that the cal-culations are derived from General Electric Company's computer-1:cd values of total flow through each fuel element and the radial flux peaking factors for the core. 2. Plant Operation Durine Tornado Alerts A discussion was conducted with Mr. Iloyt relative to Dresden op-erating philosophy at times when revere weather threatens and a tornado alert status is established by the local weather b'ureau. He pointed out that at such times considerable reliance is placed on the local electric utility by all segments of the populace and that the governing philosophy for all operating stations as es-tablished by CE management is that maximum effort must be concen-trated to keep power service uninterrupted. As far as Dresden 1 is concerned, Mr. Iloyt stated that there was no specific operating procedure for use in times of impending severe weather, but in order to maximize plant reliability the diesel gen;rator is started and placed in service on the emergency bus when a tornado alert is established. 3. Primarv System Intecrity Mr. Redman indicated that the established Icak surveillance n:cas-ures that were adopted to assure that no primary system leakage in the containment building would go undetected were being closely followed. 'lha pointed out that the low airborne activity in the containment sphere, which has prevailed since cycle 5 restart, indicates that there has been insignificant leakage from the pri-mary system. J. Exit Interview A brief exit interview was held with Messrs, licyt and Redman to ap- __.__s _ praise them of the results of the inspection, and during which the following - subjects were discussed: - continued -
. Results of Visit (continued) 1. Follow-Up Activities on Primary Systen Pipinc Cracks Messrs. Hoyt and Redman were made aware that, in the opinion of the inspector, the follow-up activities to determine the cause cf the piping cracks in the primary system were progressing at a very slow rate. They indicated that increased emphasis would be given this particular endeavor. 2. Leak Rate Testine and Updating of Technical Specifications Since the Technical Specification exhibit a longstanding defi-ciency relative to specific requirements for containment leak rate testing, it was the inspector's opinion that this situation should be corrected. Mr. Hoyt indicated that work was in pro-gress on this matter. 3. Core Pressure Drop and BOR Surveillance The participants were reminded of the importance of this matter and that during future Compliance inspections this subject would be examined in greater detail. wes .mw~~~ =}}