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{{#Wiki_filter:-s. e -J "g O ) U. S. ATOMIC ENERGY COMMISSION REGION I DIVISION OF COMPLIANCE Report of Inspection I CO Report No. 219/70-1 ~ k
- l Licensee
JERSEY CENIRAL POWER & LIGHT COMPANY j_ Oyster Creek 1 p License No. DPR-16 Category B Date of Inspection: January 6-8, 1970 Date of Previous Inspection: December 3-5, 1969 I "10 Inspected by: R. T. Dodds, Reactor Inspector (In Charge) Date l Oh D. L. Caphton, Reactor Inspector Date ' N L 0 R. T. Carlson, Senior Reactor Inspector Date Proprietary Information: None SCOPE Type of Facility: Boiling Water Reactor Power Level: 1600 Mwt Location: Lacey Township, Ocean County, N.J. Accompanying Personnel: D. L. Caphton, Reactor Inspector (Prepared Sections B (in part) D.E.3,E.4,F & K) L. Higginbotham, Radiation Specialist (Prepared Sections P.2, P.3 & Q) Scope of Inspection: Review of records of power ascension tests, secondary containment leakage rate tests, chemistry records, facility records,and tour of the facility. 9508070166 950227 PDR FOIA DEKOK95-36 PDR
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SUMMARY
? Safety Items - No items of safety significance were identified during the visit. ~ Noncompliance Items - Two items of noncompliance and one item involving the licensee's f ailure to notify the Commission in accordance with DRL's memordndum to JC dated November 6, 1969 were identified during the inspection. The items were as follows: 1. The secondary containment railroad air lock inner and outer doors were not sealed and the reactor was operating. The licensee was in noncompliance with technical specification 3'.5.B.1. (See paragraph K.2). 2. Testing of the standby gas treatment particulate filter is specified by technical specification 4.5 L.1.d. to be conducted "at intervals not to exceed six months between refueling outages". This testing frequency was not met. (See paragraph D.3.). 3. Two control rod drive scram times were outside of guide linee established by a DRL letter to the licensee and were not reported as called for by the letter. The licensee was in nonconformance with the DRL requirement. (See paragraph F.). Status _of Previously Reported Problems 1. Action taken by the licensee to correct previously reported items of noncompliance was as follows: The General Office Review Board has performed a recent quarterly a. audit of plant operations as required by the technical specifications. (Section B.2.) b. Essential'ly, the interlocks for the reactor building personnel doors have been repaired and appear to be operating properly. However, in the inspectors judgement, there will probably still be problems with these interlocks until a better system has been devised that is i not dependent upon maticulous adherence to administrative procedures. (Section K.2.) c. The licensee has instituted a surveillance program to correct the j items of noncompliance noted during the September,1969 inspection. 1 However, at least one item was overlooked as evidenced by the item of noncompliance noted above. (Section D.3.) ) Other Significant Items 1. The licensee has hired a permanent chemical supervisor. The post-warranty run GE support organization has been defined. (Section B) i 2. All of the power ascension tests have been completed. The 100 hour
m r s e 4 s . electrical output and heat rate warranty test was completed at 11:00 p.m. on December 22, 1969. Commercial operation commenced at_00:01 a.m. on i December 23, 1969. Total gross electrical generation thru December 31, ~ 1969 was 351,437 Mwhe, total thermal generation was 49,806 Mwdt,(Section C). [ 3. The reactor ' power level was reduced to hot standby during the visit to tighten the seal packing gland on one of the primary system recirculation i pump isolation valves. The combined identified and unidentifie.d leakage I rates had reached 22 GPM. (Section C.) 4 A steam line initially reported to be 2. inches but later confirmed to be 9 3/_4-inch line broke on January 7, 1970 when the reactor was at hot standby. A second occurrence of the event is reported to have occurred on January 10, subsequent to the inspection. (Section C.) 5. Section C. contains a summary of reactor scrams that have occurred since the previous visit. Some, as expected, were the result of the power ascension test program. 6. TheinspectionconfirmedtheresultsofthedatapreviouslysuppliedCO l regarding the possibility of steam leakage through one of the steam i line isolation valves. (Section D.2.) 7. Section S. and Attachment 4 contain the results (including many charts and" graphs) of the power ascension tests that were completed after achieving full power operation. All required startup tests have been completed. 8. Section E.2 contains information obtained from GE on oxygen concentrations in reactor steam, reactor water and condensate. Management Interview - The results of the inspection were discussed with Messrs. Mc Cluskey, Hetrick and Ross. y 1. Mr. McCluskey acknowledged the two items of noncompliance involving the surveillance testing of the particulate filters and the failure to maintain reactor building containment integrity during reactor operation. Heistated that the railroad doors were now properly secured. He believed that GE had probably broken containment while removing some of their spare control rod drive modules and/or when taking inventory subsequent to the 100 hour warranty run. He estimated that the inner door lateb had been broken around December 31, 1969. Mr. McCluskey stated that th'ey would attempt to identify the responsible party (ies). 2. JC was still not satisfied with the present interlocks for the reactor I_ building personnel doors and will be considering an appropriate f modification. The present system tends to be dependent upon rigid { personnel procedures. ( 3. Mr. McCluskey stated that he had asked his people about the reportability l
w 4 . ;s e*- '{ -} t -4 of the control rod scram times and had been informed that this was not - required since this occurred after full. power operation had been achieved.. 6 They overlooked the requirement that'any times that did not meet the spec-ified limits were reqsired to be reported until completion oS the " plant-startup program". Mr. Ross.said that, we didn't sea the forest for the 1 I trees." In the future, any specific reporting requirements will be [ adhered to, so' stated Mr. McCluskey. -6 '4. Messrs ~ McCluskey and Ross stated that JC will continue to correlate data i and make secondary building leakage rate tests as wind conditions permit. However, they do not plan'. to take any more measurements below 30 mph wind conditions since it appears that the building pressure becomes positive ; substantially above this wind velocity. t
- 5.. The inspectors emphasized the need for strict adherence to license requirements. Mr. McCluskey stated that they were expending,every effort in this regard but felt that some items may get overlooked during the initial operating periods.
6. McCluskey agreed to notify Inspector Caphton of the tnain steam isolation valve test so that the test may be observed by CO. The test is to be conducted during the maintenance outage scheduled for late January 1970. 6 DETAILS A. Personnel Contacted: Personnel contacted during the visit included the following: Jersey Central Power Light Company (JC) T. McCluskey Station Superientendent D. Hetrick Operations Supervisor Technical Supervisor D. Ross I. Finfrock Maintenance Supervisor l J. Sullivan Associate Engineer - D. Kaulback Radiation Protection Supervisor General Electric Company (CE) W. Hess Site Operations Manager J. Nickols Shift Supervisor R. Woods Test Engineer M. Meek Radiochemist .1 B. Administration and Ornanization 1. General I Mr. McCluskey discussed the status of organization at 5 ster Creek now
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~ .s', y e s that GE had completed the 100-hour power demonstration.run. Most of j the GE startup crew will have left by January 31, 1970,.however,._ i F several members of th'a group will stay on at Dyster. Creek.f'R. Hurd, I Nuclear Test Engineer, will stay until May, ~ 1970. In February, 1970, 1 JC will be. sending Mr. Minue, newly-hired. Nuclear Engineer, to San ' Jose, California to attend a six-week GE training session for power plant' l n K e _ Nuclear Engineers. 1 -iS Two GE Shift Supervisors (R. Elems and J. Nickols) _ will continue ,Ei on shift for approximately another six months until JC hos qualified- .l several more operators and supervisors.: 'This will strengthen the shifts - i 1 lnd enable the.JC Shift Foreman to spend more time out in the plant. l JC has now hired a Chemical Supervisor,'Mr. J. R. Pelrine. Mr. Peirine. was employed as E Radiochemist and Supervisor, Radiochemistry Section from. 1959 to 1969 at Industrial Reactor Laboratories. Prior to this he had' ten years experience at KAPL and in the U. S. Army as a,Laboratoryz Assistant (spectroscopy) and as a dental and X-ray technician-(Army). Mr. Ross stated that Mr. Pelrine would begin. work at OC-1 on or about January 19, 1970. ,1 -Mr. Svend Daalgard, Chemical Supervisor'(consultant) has been replaced by Mr. Tom McGraph, JChemical Supervisor (consultant _ from Pickert and-Lowe Associates). Mr. McGraph, per Mr. Ross,'will-be at 'the plant' on a
- five day per week schedule. Mr. Ross further stated that Mr. McGraph had OC-1 familiarity thru design experience at Burns and Rowe,-the OC-1 AE.
Mr. Ross stated that Mr. Robert Stoudenour, JC Chemistry Supervisor,'will-continue at the plant-for an indefinite period of time. GE is also con-tinuing to provide a chemist at the facility until Mr. Pelrine is ready - to take over the full responsibility 'of Chemistry-Supervisor. 4 I 2. General Office Review Board (GORB) It was observed during the November 1969 inspection that GORB had not conducted a routine' quarterly audit at Oyster Creek since July 2,.1969.. This item was included in a Form AEC-592 that was sent _to.the licensee. The required quarterly audit was finally made by Mr. D. R. Rees,. member ~ i GORB/PORC, on November 24, 1969. His report to W. _ H. Hirst, Chairman GORB, stated that he had reviewed the results of several.startup tests'as requested by CORB. The test results were compared with predicted per- .. formance_and acceptance criteria. No'significant deficiencies 'were identified. C. Reactor Operations All of the power ascension tests have been completed. The 100 hour electrical output and heat rate warranty test was completed at 11:00 p.m. on December 22, 1969. Commercial operation commenced at 00:01 a.m..on
-.~.-.-: - - -. - Q ) . December 23, 1969. Total gross electrical generation through December 31, 1969 was 351,437 Mwhe; total thermal generation was 49,806 Mwdt. The reactor power level was reduced to hot standby the evening of January 6,1970 because of high primary system leakage in the drfwell (22 gpm c t otal)'. The identified leakage had increased from about 5 gpm to 19 gpm [ and the unidentified from about 1 gym to 3 gpm. The increase of unidentified leakage was attributed to overflow of the reactor drain tank because of s maloperation of a drain pump. The leakage was traced to a high heal packing leak-age of the discharge isolation valve for recirculation pump "A". The manu-facturer's instructions stated that the valve packing shouE be tightened after substantial usage which had not yet been done for this valve. The packing for all valves was adjusted during the outage. The reactor was shut down at 9:05 a.m. on January 7,1970 when the reactor prinary system was isolated because of a reported 2-inch steam line break. The 9:05 a.m. reactor log book entry stated, " Steam leak reported in turbine building. Two-inch pipe break at line to turbine stop valves. Roof vents closed. Lowering reactor pressure to less than 600 psig to close MSIV.'.' The reactor pressure had been reduced to 600 psig by 10:00 a.m. The line that broke was determined to be a vent line that had been attached to one of the four 18-inch main steam lines entering the steam chest. Mr. McCluskey informed the inspectors that the line had been repaired by re-moving the valves and capping the stub. During later conversations with the licensee,, it was learned that the line that broke was not a 2-inch line at all but rather, it was only a 3/4-inch line. It was further stated that the hole size in the steam line where the pipe connects was only 1/4-inch diameter. The line, which had two closed valves and a cap on it, was about 18-24 inches long and was unsupported. The break appeared to be vibration induced and occurred close to the attachment weld. Each of the other three main steam lines also have a similar line attached to them. ~ Mr. McCluskey did not know the reasons for the lines but assumed they had been used for venting and pressure testing of the main steam lines. He stated that for now, the lines will be temporarily supported. They will be namoved entirely during a scheduled maintenance outage in late January:,1970 An examination of the other three lines did not reveal any anomalies. During a subsequent telecon with Mr. Hetrick, it was learned that the line that broke originally had not been removed and capped as originally communicated, but instead had been welded back in place. Purther, that the line had failad again on January 10. He stated that at the time all four lines were removed and capped. The following scrams have occurred since the time of the last visit (December 5) to January 5,1970.
I; ,R, . ; *h ; a__,_. A_ .a (a f ' e i ";, ;. . p.- L'. p + p y,- t 1. Generator Trip Test ? u t k As. expected, the reactor scramed..at 9:30 p.m. on December 8,1969 as a result of the generator trip test'at 1580 Hwt,3526 Mwe. The @?% l reactor scramed on high APRM neutron flux caused by a reactor' press'ure{ - L increase of.50 psi /sec. The scram occurred 0.9 seconds after the generator trip and the APRM's peaked at'1507., 1.1 seconds after trip' and were above 120% for 0.25 seconds. The pressure. peaked'at 1055' I psig 2.9 seconds after trip and no pressure scram trip signa 1' occurred.- e A low level scram trip signal c'ame'1.7 seconds after the trip but the i level did not drop below'3 feet (scram point - 4.3 feet) on the level indicators. No significant safety problems were identified as. a result. of this test. 2. Main Steam Line Isolation valve Test As expected, the reactor scramed at 12:20 p.m. on DecemberL9,1969 -i from 10% closure of the main steam line isolation' valves as the result of.a test at 1580 Mwt. The resultant pressure transient was not sufficient to actuate the isolation condensers ( T1060 psig). - No : significant safety problems were identified as a result of this test.. l -3. Turbine Trip Test (Reduced Flow) j As. expected', the reactor scramed at 5:40 a.m. on December-12, 1969 'l from high flux because of a turbine trip test at 1135 Mwt, 370 Mwe, i on flow control (reduced flow) from full power rod pattern. The i APRM signal increased from 70% to a peak of 118% (scram set at 103%). The scram point was determined. (biased) by the core flow which was 69% of J! rated. The pressure peak was.only 36 psig. above the normal operating. l pressure of 1000 peig. No'significant problems were~ identified as a-4 result of this test. i 4. Turbine Trip Test (1600 Mwt) i The reactor scramed at 6:21 a.m. on December 13, 1969 from high -l neutron flux as the result of a turbine trip; test" at 1591' Mwt, l ~ 541 Mwe. Due to the power-steam flow mismatch, reactor pressure .l 4 increased at an initial rate of 300 psi /sec for.0.1 seconds when an 1-APRM high flux scram occurred. A pressure scram-signal occurred O.8 seconds later at 1060 psig. The pressure peak was 1060 psig. and the heat flux increased less than 1% although the APRM's; peaked i at 105% 0.2 seconds after trip. The APRM's went to 0.to 0.4 seconds following the trip. No significant safety problems were identified-j as a result of the test. I 5. Main Steam Line Break (False Indication) The reactor scramed at 10:47 on December 16, 1969 because of a false
. - e h ') 1 main steam line break indication. The reactor power level at the time was about 400 Mwt. The false indication occurred from low sensing line pressure as an instrument technician valved in main steam line break and flow instruments after calibrating the flow transmitter. r The technician had already successfully performed this manipulation for the other three sensing lines. No safety problems were identified as a result of this scram. 6. Low Reactor Water Level The reactor scrammed at 2:45 p.m. on December 16, 1969, during a reactor startup (*s 90 Mwt), from low reactor water level. The bypass valves were being used for level control and were opened a little too wide. This caused the reactor pressure to drop which caused swelling in the reactor giving a 1cvel increase. In opening the bypass valve wider to drop the icvel, the swell increased giving a higher icvel. Closing the valves increased the pressure collapsing the voids and causing the level to drop to the scram trip point. This situation becomes a cascading problem; since, if level is brought up by a rapid feedwater flow increase, a flux scram would probably occur. 7. Condenser Low Vacuum ~ A reactor scram occurred at 12:26 a.m. on December 18, 1969 from low condenser vacuum at a reactor power level of 740 Mwt. Loss of con-denser vacuum was caused by failure to isolate waste' sample tank "A" when the level had dropped to the pump trip set point and air was pulled into condensate pump suction header through the vaste sample tank by condenser vacuum. The pump trip set point had been set down; but even with a pump trip, the problem would still exist until the isolation valve was closed by the operator. According to Mr. Hetrick, in order to prevent this problem in the future, the isolation valves V-22-17 and -18 have been wired to close on low level when the pump is tripped with an emergency backup trip set at 5 percent. The water remaining in the tank can be used for backwashing filters or discharged into the canal. 8. Reactor Low Water Level A reactor icw water level scram occurred at 2:05 p,m. on December 31, 1969 at a reactor power level of 1595 Mwt, 540 Mwe. Just prior to the scram, the operator notices the water level in the reactor decreasing and that the feedwater flow had dropped from 5.9 x 106 to 3.8 x 106 lb/hr. Prior to the low level alarm, the operator switched the master feedwater control to manual and attempted to increase the level.
!? g ; ;,3 ' grJ .s.; -v K. -y ex m IJ ~ , However, before the '1evel started to recover, the reactor scramensd on . low level; and the turbine tripped, causing the voids to collapse, e further depressing ~ the water level to the low-low level trip which - .I' isolated the reactor and started the core' spray pumps. The operator y returned the master feedwater control-to auto, and at this tims, the j~ reactor level. increased. This increase may have been due to the time l' i required ' for ' water to go from feed pumps to the reactor vessel after the ( 11-operator changed to manual and. increased the. set point. When level re- ,i turned, the feedwater ptsups were tripped and core spray system reset 'with the pressure still at approximately 1000 psig. The trip and isolation all functioned properly with all equipment operating. as it: should. While: shutdown,- the feedwater system was thoroughly checked and no trouble.could be located ~or found in the pumps or. control. system, and no leakage or other evidence of loss of coolant could be found. After a thorough investigation.of the plant. and equipment, it was-decided to go back to het' standby to check out the operation of'the feedwater controls. This check out turned up no evidence of any nal-functioning of equipment or controls so power was increased gradually. The only unusual condition during the startup was that one of the' steam flow instrument's excess flow checks' stuck and had to be. tapped in order j to get it open. Further investigation during operation-has turned up no evidence of what might have happened to the feedwater system or controls.. 1 9. condenser Low Vacuum 1 A condenser low vacuum scram occurred at 11:06 a.m. on January 3,- 1970 at a reactor power-level of 1425 Mwt.. -l Prior to this scram, the operator noted that condenser vacuum was decreasing and checked the air ejectors which were operating properly,. then started to reduce the load to attempt,to hold vacuum. During this time approximately 45 minutes, other operators'tried to locate the, leak but were unsuccessful, and the reactor scramned on low vacutsu. Further investigation efforts disclosed that 'a flange on the auxiliary flash-tank pump line was leaking, This flange was then tightened. However,- while starting up, there were still indications. of excessive in-leakage to the condenser as indicated by an increate of approximately_30 percent f. in the off-gas flow rate, This leakage could not be located and as power was increased, the vacuum improved.so the plant was slowly-re-turned to 500 Hwe and investigation for condenser in-leakage was' continued-but to no avail. The vacuum leak was finally located in an expansion joint between the turbine and "C" condenser on January 6,1970.. Vacuum returned to normal of)30 inches mercury..once the leak'was plugged. p ~ f l l
, -----~~ _L- . ;d e O g . D. Facility Procedures 1. Minimum Check, Calibration and Test Frequency for Protective Instru-mentation f i f A followup audit was made of three instrument channel surveillance items listed in technical specifications Table 4.1.1. These items were -previously audited, in September 1969*, and the licensee found to have been in noncompliance with surveillance requirements. The surveillance I checks audited by the inspector subsequent to the September inspection s were determined to comply with the requirements of the technical speci. l fications. The items audited were: Tech Spec Item ** Instrument Channel 6 Low-Low-Low Water Level 14 High Radiation in Reactor Building 15 High Radiation on Air Ejector Off-Gas The inspector determined that a discrepancy existed'in the air ejector off gas calibration of November 18, 1969. A 1.12 R/hr (at 3 feet) source was used for the calibration. The indicated chamber out withthesourceatifootforthetwochannelswasonly7.5x10 gut The inspector calculated this to be low, i.'e., it should have read 3 ae9 x 10. Mr. Ross stated that the cause for the error would be determined and corrected. Mr. Ross stated that a new portable calibration cart had been placed in service since the initial CO audit of this item supposedly to enable JC to perform this calibration accurately. The " Surveillance Schedule Book" that was being maintained by Mr. Sullivan was reviewed. Mr. Ross stated that three copies of the book was being maintained, one for the instrument foreman, one for the shift foreman and one for Mr. Sullivan, who amintains the surveillance records. The " Surveillance Schedule Book" was initiated *** to aid in coordinating the overall surveillance of technical specification items that occur on a frequency of three months or less. ' 2. Surveillance of Core Spray System A sampling audit was made to determine if surveillance requirements listed in technical specifications 4.4 paragraph A. were being followed. Items 1, and 3., pump operability and auto actuat. ion.taat, were audited. An in-
- C0 Report 219/69-16, paragraph D
- See Technical Specifications Table 4.1.1.
- The
- See Technical Specifications Table 4.1.1.
" Surveillance Schedule Book" was first mentioned by Mr. G. Kelcec, ' JC, I in a letter to Mr. R. W. Kirkman, CO:I Director, dated November 12, 1969 as an improvement in the JC surveillance program.
h. M ; y'. q.;_a.L - a, -a '1 pfp' c ..m 1 s i l spection of the records.for these items determined the test.to have [ been conducted in accordance with-the technical specifications. 2 3 -3. Testina of Particulate Filter i i: A sampling audit was made-to determine-if the technical specification: -j I: 4.5, paragraph L.1.d. regarding testing of partiev. late filters-had.. T been performed at intervals not to exceed six months between refueling. 't ' outages. It was determined that the 'last test was made on June '12,_.1969. y The frequency of testing failed to meet the specified testing frequency J; required by the technical specifica$ ions. 'The licensee is in noncomp1,1-j ance with technical specification item 4.5 L.1.d. l An audit of the June 6, 1969 test determined that the-licensee was in- -i compliance with paragraph L.1 at that-time. The filter efficiences. ~ determined during the June test.were: l Train A (EF 1-8) I Upstream afficiency, greater than 99.957. Downstream efficiency, greater than 99.93% l Train B (EF-1-9). Upstream efficiency, greater than 99.95% Downstream efficiency, greater than 99.957. The June test were made by the Cambridge Filter Corporation. -l 1 E. Primary System Il 1. Reactor Power Level (Heat Balance) i Frequent heat balances have demonstrated that theLsteady state reactor. '{ power level has been maintained at or below 1600 Mwt..It was observed.- that heat balance checks were made by JC operations personnel = anytime; l there was eignificant change in power ~ 1evel' and, 'as a minimum, at least, once per shift..The technical. specifications only require a heat-balance surveillance verification:of nuclear;instrisment power levels once^' every third day. The operations ~ log usually indicated that the-nuclear instrimments had been> adjusted to agree with each new heat - r i '3 balance if any substantia 1'varianbej axisted. j 2. Steam Line Isolation-Valves ) No further' testing has been done o'n the leakage rate of the-main steam line isolation valves beyond that reported in CO Inquiry - Memorandum 219/69-I, dated December 18, 1969. The raw data"that was .I obtained during the testing of the valves on December-13, 1969 was reviewed and determined to be consistent with the information reported by the licensee and contained in CO's Inquiry Memorandtsn. j ) ~ 5
. r- .-~ ,[] ;; 3. Water Chemistry y I Mr. Ross reviewed the current chemistry program with the inspector. [ Mr. Ross stated that it was his aim, now that the reactor was considered [ to be in comercial operations, to have the sampling frequencies l~ brought in phase with their planned operating schedule. Attachment No. 1 } is dataifrom a sampling a~udit of reactor water chemistry. Attachment No.2 gives metals analysis data for condensate, demineralizer effluent and t-feedwater. 4. Oxygen The following information was obtained from Mr. Myran Meek, GE Radio-chemist on January 7, 1970: Oxygen in reactor steam = 24.5 ppm Oxygen in reactor water = 200 ppb (Sample from A recirculation loop) Oxygen in condensate = 40 ppb
J . ~ s ~ O 3 pj,
- F.
Reactivity Control and Core Physics Control Rods n Mr. McCluskey stated that GE was preparing a position type report on the control h rod drive system which would be used as a basis to reply to a request in Dr. Morris' L letter to Mr. Ritter dated November 8, 1969. Mr. McCluskey stated that GE wanted { to remove their 30 point scram time recorder, however, he stated that the recorder i would remain in service until a decision was reached on JC's overall position on the ( subject.* i Mr. McCluskey stated that GE has proposed to change the initial N2 accumulator charging pressure for all control rod modules.** He stated that this work would probably be accomplished during the scheduled January, 1970 outage. JC continued surveillance on control rod drive scram times in compliance with the guide lines established by Dr. Morris' letter to Mr. Ritter, dated November 6, I 1969. The table on the following page gives data from five different scrams for the 26 monitored control rods.*** 1 The inspector noted that scram times for two coordinates (06-19 and 46-11) for the December 16, 1969 scram were below the 1.9 second guide line in Dr. Morris' letter. (The average for the 26 monitored rods was also below the 2.4 second guide line.****) The inspector asked Mr. Hetrick if any additional scram tests were made on the two fast drives or any other drives to determine if the condition was real. He stated, "No". He stated that the recorder was suspected to have failed to start on time, that possibly a relay was slow in picking up. He stated that a j 60-cycle /second input signal was used to check the recorder, however, it responded i properly during the check. i j Mr. Ross was asked why CO had not been notified regarding the two fast scram times in conformance with the letter from Dr. Morris. He stated that it was assumed i that the recorder had malfunctioned and the condition was not real. He stated that they had not experienced any problems with short scram times before or after the December 12, 1969 scram. 1 The inspector asked if any accumulator overpressure alarms had been noted at the time of the December 12, 1969 scram. Mr. Hetrick stated that none had been observed. The inspector stated that JC appeat ed to be in nonconformance with the 1.9 second requirement stated in Dr. Morris' letter concerning the two individual drives and JC's failure to notify C0:1. The inspector expressed concern that the reactor wat started up without verification scram tests on any drives. i i
- For a description of this recorder...se'e CO Report No. 219/69-10, Paragraph B and footno$e.
- CO Report No. 219/69-13, Paragraph F.
- For previous data see C0 Report No. 219/69-13, Paragraph F.
- Note, the 2.4 to 3.1 second range stated in Dr. Morris' letter is for the average times for all 137 control rods.
- For previous data see C0 Report No. 219/69-13, Paragraph F.
- CO Report No. 219/69-13, Paragraph F.
-. ~ s ( 3 14 - Scram Insertion Time (Seconds) Date of Scram 12/8/69 12/8/69 12/16/69 12/18/69 1/ 3 /70 8:00 pm 9:30 pm Reason for Scram Manual Gen "Irip Low H O Low Vac Low Condenser Vac 2 Coordinate 90% 90% 90% 90% 10% 50% 90% l 06-43 2.32 2.39 1.93 2.37 0.37 1.32 2.36 06-27 2.59 2.60 1.96 2.53 0.38 1.48 2.64 j 06-19 a 2.35 2.37 1.88 2.33 0.37 1.32 2.33 10-39 a 2.38 2.42 1.90 2.40 0.37 1.36 2.48 14-19 2.50 2.62
- !.03 2.50 0.37 1.37 2.59 l
14-27 2.61 2.64 1.94 2.49 0.38 1.40 2.51 14-35 2.48 2.45 1.90 2.43 0.37 1.40 2.46 14-43 a 2.61 2.70 2.08 2.66 0.38 1.45 2.62 18-07 a 2.47 2.51 1.96 2.39 0.37 1.35 2.36 18-23*a 2.71 2.67 2.00 2.35 0.37 1.34 2.40 18-47 a 2.61 2.71 2.09 2.61 0.38 1.45 2.62 22-27 2.44 2.43 1.99 2.30 0.38 1.37 2.39 22-35 2.48 2.50 2.03 2.50 0.39 1.42 2.50 22-43 2.42 2.41 1.97 2.43 0.38 1.39 2.43 30-19 2.47 2.45 2.03 2.58 0.38 1.40 2.46 ) 30-27 2.48 2.53 2.08 2.50 0.42 1.52 2.58 30-35 2.36 2.40 1.96 2.38 0.38 1.39 2.43 30-43 2.46 2.36 1.91 2.32 0.38 1.33 2.33 30-51 a 2.68 2.75 1.99 2.57 0.37 1.40 2.62 j 38-11 2.36 2.44 1.91 2.48 0.37 1.38 2.46 38-27 2.59 2.67 1.98 2.43 0.37 1.44 2.59 38-35 2.50 2.57 1.97 2.48 0.38 1.41 2.58 38-43 2.50 2.42 1.90 2.47 0.36 1.35 2.45 46-11 2.37 2.46 1.86 2.39 0.35 1.33 2.44 j 46-19 2.51 2.59 1.97 2.54 0.38 1.42 2.62 46-35 2.38 2.38 2.03 2.42 0.38 1.37-2.39 907. Averages 2.45 2.51 1.97 2.46 2.49 l a. Drives with filters l Erratic scram times i l l
-- - ?.. i..~~.. :. t. .te. .c y 15 - Control rod drive 18-23 has performed erratically relative to its scram time . pe rformance.* The drive is being followed closely by JC. The following tabulation gives insertion times for this drive, t.r Scram Insertion Time (Seconds) Control Rod Drive 18-23 t l'. ] Date Recorded 107. 507. 907. l'2/8/69 0.38 1.36 2.71 12/8/69 0.36 1.41 2.67 12/9/69 0.36 1.56 3.52 12/12/69 0.35 1.35 2.50 12/16/69 0.32 1.30 2.32 12/16/69 0.33 1.10 2.00 12/18/69 0.36 1.30 2.35 12/31/69 0.38 1.35 2.37 1/3/70 0.37 .1.34 2.40 4
- C0 Report No. 219/69-10 Paragraph D.1 and CO Report No. 219/69-13, Paragraph F.
i ) A } e e O ~ ' ) K. Containment i 1. N2 Inerting i The drywell was initially inerted on December 23, 1969. The technical f specifications
- require that the oxygen concentration be checked at least weekly. JC is routinely making this check hourly. A sampling audit of surveillance records was made and data is tabulated below:
% 02 Concentration ** Date N9 Meter Reading Pressure (Psig) Drywell Torus 1/2/70 8526 0.4 4.50 3.50 1/3/70 9150 0.4 4.55 3.50 1/4/70 9418 0.4 4.80 3.50 It was determined that N2 inerting surveillance'and 02 concentrations meet requirements of the technical specifications. JC has not yet determined accurately the amount of N2 make up to the dry well. Several variables are being appraised to determine the actual make up, for example: N2 make up through the TIP system N2 removed via sampling N2 supplied via the normal make up. It was necessary to deinert the containment for the January 7,1970 shut-down in order to permit personnel to 6nter the drywell. Mr. McCluskey stated that a special truck was required for N2 inerting and required scheduling which he stated could present a problem; the truck at that time was in Pennsylvania. Mr. Ross also stated that N2 inerting cost were approximately'$7,000. per charge. 2. Secondary Containment Doors { During the tour of the facility a inner personnel interlock door located at the south east corner of the reactor building was observed to be cracked open several inches. Inspection of the door determined that the electrical interlock solenoid plunger was holding the door open. The companion out-side interlock door was closed, however, its solenoid interlock plunger was in the retracted position thus potentially enabling the door to be opened from the outside. This condition would have permitted anyone entering.from the outside to have compromised the secondary containment. The inspectors
- Technical Specifications 4.5 paragraph M.
- Technical Specification 3.5 A.6. specifies less than 5.0% O -
2
s ) 4 17 - did not observe anyone to enter the outside interlock door during the r time of their inspection, however, the potential existed for compromising i secondary containment. This condition was observed by inspectors Dodds and Caphton and Mr. Sullivan of JC. The reactor was operating at the l- { time. t L The inner railroad air lock doors were observed to be sprung inwardly at [ the center bottom. The screw Jack was missing from the center bottom of the inner doors. The upper center cam was also observed to be broken - + i completely off of the door.* The cam was observed to be' laying on a crate approximately 15 feet from the doots. The inspector.placed his hand at the center of the inner doors and could' feel air (wind) entering the secondary containment. The inspector observed that the doors could be physically moved,therefore further confirming that the doors were not against their sealing gaskets. The inspector requested Mr. Sullivan to open the inner railroad air lock doors so that an inspection could be made of the outer railroad interlock doors. Approximately 40 minutes lapsed before the electric control box was unlocked and the inner doors opened. The inspector observed that all Jack screws were missing from the outer railroad air lock doors.* The inspector observed tha't the upper center cam was completely un-cammed. The lower center cam was in the cam position and was the only mechaniam being used to hold the doors together. The { inspector could visually see light through a crack at the bottom center of j the doors. The doors could be physically moved,therefore further confirming I that the outer doors were not against their sealing gaskets. The inspector j observed jack screws of the type useo to seal the outer doors. laying on the. I floor of the railroad air lock in the vicinity of the outer doors. The reactor was operating at'the time. Reviewing the above observed facts along with prior observations from previous inspections regarding the requirements to effect the sealing of the railroad airlock outer and inner doors in order to achieve compliance with technical specifications paragraph 4.5 J.2, (leak rate testing), it appears that JC was probably in non-compliance with this requirement. It appears that JC was also in non-compliance with technical specification 3.5 B.1. (containment integrity). 3. Secondary Containment Leak Rate Test The inspector reviewed leak rate test results with Mr. Ross. Six tests have been conducted from September 3, 1969 to January 6, 1970, the date of the inspection. The results of the tests were:
- The screw jack and cam are necessary to effect a seal.
Reference to CO Report 219/69-8, paragraph K.
- C0 Report 219/69-9, paragraph K.1. describes the jack screws necessary to achieve sealing integrety of the outer railroad air lock double doors.
Q. i' a a .._ a _ ____ i I 'J t Wind Vacuum Direction Speed mph in. H O 2 i (NNW) 0 0.274 1 SE 8 0.253 - I S 10 0.312 b SE 22 0.252 . [ SE 22 0.266 h' W 25 0.210 4 Mr. Ross is maintaining a graph of the above' data in order, he stated, to develop a curve from which it may be determined for any wind condition, whether secondary containment'is in compliance with the 0.25 "H O under l 2 calm wind condition. requirement.* l i The inspector asked Mr. Ross when he estimated completion of the data acquisition' program. He estimated two years ami added that the program - had low priority. He stated that he was only testing if the wind velocity-was above 30 mph. The inspector received confirmation from Mr. Ross that the existing data and graph was essentially unintelligible at this time. The inspector stated that it appeared that to meet the requirements of the technical specifications, omni-directional curves would be required. The inspector stated his concern to Mr. Ross, regarding the slow progress being made by JC in obtaining the needed data. Mr. McCluskey subsequently stated that they were only trying to determine the velocity of the wind that causes the building pressure to reach zero, not the different directional curves. P. Radiation Protection 1. Tour During a tour of the facility, it was observed that radiation areas were appropriately posted. However, contaminated areas were not always clearly identified to preclude one from walking into the area without first going past " shoe covers required" sign and step off pad. During the exit int e r-view Mr. McCluskey stated that radiation warning rope would be placed in the deficient areas to herd personnel through the step-off pad area.
- Technical Specifications 4.5.J.1. and 4.5.J.2.
^ ( t V 1r o 2. Radiation Surveys Mr. Kaulback said that the results of test 5, Radiation Measurements. C performed at 1600 MWt show no results which differ significantly from ? expected radiation levels. A review of the survey ~results and comparison with dose rates predicted by GE verify Mr. Kaulback's statement. Radiation s levels in the turbine building operating floor vary from 220 mr/hr, at [ contact with the High Pressure turbine to 160 mr/hr at 1 foot, and 30 to 60 mr/hr in the immediate vicinity of the Low Pressure turbines. General radiation levels at 20-25 feet from the turbines average about 20 mr/hr. Radiation levels in the control room were measured as less than 0.1 mr/hr. Radiation levels in the bay housing the air ejectors varied from 310 mr/hr at 1 foot from the air ejectors to general area levels of 25 mr/hr in the bay. Levels in the labyrinth entrance to the bay varied from 4 mr/hr to 12 mr/hr. Mr. Kaulback said that the problem with radiation levels in the counting room still exists.* Radiation levels are caused by the TIP monitors when they are retracted into the TIP room, which is adjacent to the counting He said that when time and other workloads permit, corrective action room. will be taken with regard to this matter. 3. Personnel Monitorine Review of records revealed only minimal accumulated exposures, which is to be expected considering the short operating history of the facility. Highest exposures were about 50 mrem / month. Q. Waste Disposal Systems 1. Liould l Records showed the following amounts of liquid wastes discharged: 1 i ToYel Total Average Concentra-(1969) No of Waste Dilution B -Y tion After Dilution Trit-- l Month Releases Gallons Gallons Curies uC1/cc Cu rie I 5 9 August 68 9.7 x 10 2.8 x 10 0.003 3 x 10-10 0.0 9 September 115 1.8 x 106 9,4 x 10 0.115 3 x 10'9 0.1 October 99 1.4 x 106 9 5.7 x 10 0.07 3 x 10' 1.2 1 November 121 1.6 x 106 9 7.5 x 10 0.087 3 x 10*9 1.5 6 9 December 123 1.3 x 10 5.4 x 10 0.2 1 x 10-8 2.2
- C0 Report 219/69-11, Section P.3.
s - 1 o 3 e - A sampling review of individual releases showed them all to be within applicable requirements. The licensee uses the MFC value of 1 x 10-7 uCi/cc (unidentified) from 10 CFR 20, Appendix B, Note 3.c. 2. Gaseous Records showed the following gaseous activity released through the stack: Ev ~ Month Average Particula tes Halogens (1969) Total Curies Release
- Rate (Curiet)
(Curies) September 28.8 11 uCi/sec 0.000153 0.000065 October 57.6 22 uCi/sec 0.000019 0.0 November 939.7 360 uCi/sec 0.000202 0.000259 3. Evaluation of Gaseous Releases While discussing the stack releases and calibrations of off-gas and stack monitors with Mr. Ross, he supplied the following information: s. Prior to November 25, 1969 the stack monitor recorder had been indicating about 6 counts per second (cps) (2 cps net over normal background of about 4 cps) at nominal power levels of 1200 MWt. Based on the calibrations performed by Mr. Kaulback with liquid Cs-137 solutions, 2 cps was about the minimum sensitivity of the monitor. b. On November 25, with the plant at a nominal 1200 MWt, the stack monitor recorder indicated an increase in activity as shown on. As indicated by Mr. Ross's marginal notes in Attach-ment 3, samples taken of the off-gas system gave results which in-dicated a stack release rate of about 3250 uCi/second of noble gases. Mr. Ross said that since November 25, the normal indicated stack release rates have been about 3000 to 5000 uCi/second at nominal reactor power levels of 1600 MWt. Based on analysis of grab samples of off-gas the calibration assigned c. to the stack monitor was (from results of 6 samples at stabilized power levels from 1200 to 1600 MWt) about 100 uCi/second per 1 cps. Mr. Ross said that using this calibration the stack monitor "hi-alarm" was set at 1/10 of the annual average continuous release rate for noble gases.
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i Racetor Power Nwt 1417 1600 1590 1600 1593 1600 ' Racetor Power Mwe 470 510 548 538 550 540 r W = Feedwater C = Condensate DE = Demineraliser Effluent
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Tj =. ' ' ,) O s y q JERSEY CENTRAL POWER & LIGHT COMPANY ) C0_ Report No. 219/70-1 ) ~ ATTACHMENT 4 ) I EXPERIEMENTS AND TESTS I l f 3 1 f 1. Matti Stream Line Isolation Valves l l ,l [- Date Performed: December 9, 1969 1 9 t Reactor Conditions: 1581 Mwt, 526 Mwe t 61 x 106 M lb/hr Core Flow
Purpose:
a. Determine reactor transient performance, b. Determine reactor fixed heat losses, c. Determine isolation valve closure time and leak tightness. Criteria: a. Reactor pressure must remain below the setpoint of the first j safety valve (1212 psig). b. Isolation valve closure time should be 10 seconds or less. Results: The isolation valves were closed resulting in a scram when the valves reached their 10% closed position. The indicated closure times of the valves were: NS03A = 8.2 sec. NS03B ="7.7 sec. NSO4A = 9.4 sec. NSO4B = 7.9 sec. The relatively long time between scram and valve closure caused a reactor t depressurization to about 800 psig. Level dropped about 4.7 feet due to g void collapse from the scram r ing in an increase in feedwater flow I to 133% of rated. M' (3 l-Heat flux dropped rapidly after the scram and never exceeded its initial I value. Level increased again after the isolation to about 7.4 feet with a resulting decrease in feedwater flow. Pressure increased again due to E decay and recirculation pump heat input. Two of the Sandborn recorder makes are attached. (Figures 1 and 2). Valve NSO3B was found to leak during the leak check. Valve NSO4A appeared to have a small leak. The other valves were tight. Leakage as it relates to the technical specification limits will be tested during a maintenance outage in late January,1970. t 9 w ___m. .4
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Date Performed: De'cember 8, 1969 3 m j' Reactor Conditions: 1579 Mwt, 519 Mwe~ r 6 ,y t; 61 x 10 lb/hr Core Flow g- ] h.
Purpose:
Evaluate flow-power transient resulting from tripping the recir-culation pumps and the steady state natural circulation operating conditions. MCHFR (minimum critical' heat flux ratio) h 1.5 during flow-power Criteria: a. transient. b. MCHFR calculated from operating data must be greater than or equal to the pre-calculated values shown in APED topical report 22A2206. Results: The trip of all five recirculation pumps at 1600 Mwt resulted in a flow-power' coastdown rate which was slightly slower than predicted.. The actual ~ flow - power, data is plotted on the predicted response curves, Figures 3 and 4. The MCHFR during the transient increased from the initial.value: of 3.24 and at all times was greater than predicted. l The MCHFR did not decrease because the peak heat flux location was' low in the core and the quality remained. less that 10% at the peak heat flux position. The MCHFR based on operating. s data is plotted on the predicted MCHFR response curve, Figure 5. At equilibrium natural circulation, the reactor power was 623 Mwt and 205 Mse j or 40% of rated power with 22% of rated flow.. The actual natural ~ circulation i characteristic for the plant is shown on the attached Flow Control: Map, 1-Figure 6. I e i 7 ATTACHMENT 4 (page 5 of 46 pages)
~W .4% p k, - N... s tr. 'h 4. 6 L The APEH neutron flux decreased faster than flow while heat flux legged
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due to the fuel time constant. Pressure drops due to decreased steam flow ~ -r (the EPR holds turbine throttle pressure constant) which in combination with the decreased flow causes increased void content and consequent level I swell. The pressure dropped a total of 44 psi to 956 psig and level in-u creased one foot to 6.5 feet. 5 y l 'e} At the natural circulation condition, the following tests were performed: SUT #19, EPR setpoint decreased than increased 5 pai, SUT #20 Bypass Valve
- 2 tripped open, SUT #21 reactor level setpoint decreased then increased 5 inches and SUT #22 flux response from insertion of rod 26-27 from 30-10-30.
These tests demonstrated the stability of the plant at natural circulation conditions. 4 i i 1 i b ATTACHMENT 4 (page 6 of 46 pages)
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j; 1 I i i te i; n u i :r: le - "n ..i! pt:dii.h.Ju ll_ l..ii.' iil ld.;..tjr[-. .h_h. FICURF'.6 i l' li lt ~! ! e. l'l. 1 l 1- " n e-d 3.aD z UD' @". " yd w.Y i. .t ATTACHMENT 4 s t ' dv. %page 10 of 46 pages)- e i.
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3. Flow Control Warranty Demonstration- -Date Performed:. December 12, 1969 - l Reactor Conditions: 80%. 430 Mwe (gross) 412 Mwe (net)- 90ir. - 486 Mwe (gross) 464 Mwe (net). 48 100% - 535 Mwe (gross) 713 9twe Xnet) 4 i
Purpose:
Lemonstrate compliance with warranted plant load.following. ,y I capability. Criteria: The plant.shall respond to-ramp load changes at a rate of i 8 per cent of full load rating per minute for a magnitude of 2 to 1 within a plant load range of 40%.to 100% rated. Results: We test was performed as described in the test procedure. The plant-gross generator output records were used to determine load change rate. The gross electrical output corresponding to 80%, 90%, and 100% of rated power is 432 Mwe, 485 Mwe, and 535 Mwe respectively based:on a value of 20 Mwe plant auxiliary load. The_ low end of the flow control test correspond to 40%, 45% and.50% with gross output of 225 Mwe, 250 Mwe, 272 Mwe based on a plant auxiliary load.of 18 Mwe. The plant auxiliary load decreased.due to-the reduction in recirculation pumpipower and'feedwater pump power. ne power decreases were initiated.by stepping the master flow controller demand free 80% at rated flow to 10% and flow decreased at.the control sys- [ tem limitor rate of 15 - 20%/ min. The power increases'were intiiated by stepping the master controller from 10% to 40% then 5% steps-every 20 see-onds to allow the operator to terminate'the increase before rated ~ficw was reached, if necessary. As in the flow ~ decreases, the' flow rate limit'de- { termined power increase rate, however, the last 20-30 Mwe increase was due 3 ~ to recovery'of feedwater temperature which took 1 to 2 minutes. j g Table 2 and Figure 7 tabulate the results and show a copy of the plant .[ generator output record. 4. 3 i fi. ATTACHMENT 4 (page 11 of 46 pages)
y, ,; s -o 3 p './ ' 't 3 t c ';5[ ' ~ ' i( 1 1 L(: TABLE 2 i p t -(js POWER }-
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Low -Decrease Increase f Gross Net Gross Net Time Rate Net-Time. Rate Mwe Mwe "3 Mwe Mwe }
- 7. Rated Min. % Min.
- 7. Rated Min. 7. Min.
T 432 412 80 224 226 40 40 2.5 16 40 5.0 8 l 485 464 90 250 232 45 45 3.4 13.3 45 5.0 9 535 5 15 100 275 257 50 50 3 16.5 50 5.6 8.9 $r 1 ,1 t i i I t i t 9 ? l L ATTACHMENT 4 (page 12 of 46 pages) .i
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,Mn'- i ,y .g 1-1 14 -- I -i-i v i b - 4. Flow Control 'Date Performed: December 8, 1969 h Reactor Conditions: 1571 Mwt, 530 Mwe .$I 61 x 10 lb/hr core flow lv l -i
Purpose:
a. To determine plant response to recirculation flow control. i b. To demonstrate compliance with warranted plant load following capability. Criteria: Thedecayrat'ionmustbe(1.0. Results: The following flow transients were performed: From rated power - a. A 167. step decrease in flow followed by 3 b. L 147. step increase in flow c. A 20% step decrease in flow followed by d. A 157. ramp decrease in flow to about 72% power. ] From 71% power - a. A 147. step decrease in flow followed by b. A~147. atty increase in flow followed by 4
- c. ' A 397, rag decrease in flow to about 44% power.
From 487. power - a. A 127. step increase in flow followed by j b. A 13% step decrease in flow followed by ~ c. A 65% ramp increase in flow to about 947. power. The readings of system parameters from the Sanborn traces for these transients are listed in Table 2 All variables were better than quarter damped for all transients. A graph of thermal. and electric, power versus flow is given in Figure. 8 7.,' A ' ummary of core thermal power versus flow is seen on the flow-s I control map in Figure 9. ATTACHMENT 4 1 (page 14 of 46 Pages) -
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.O 3 i/ '( [ i e-1 r-i f-In the final ramp increase from.746 Mwt to 1492 Mwt,.the electric power 1 increased from 240 Mwe to 495 itwe in 480 seconds, an increase of 8.25*/. of. .i-rated load per minute. This demonstrates. compliance with the' warranty -{' requirement that "the plant shall respond to ' ramp load changes of a mag-nitude of 2 to 1 at a rate of'i8 per cent of full load rating per minute". jf. ~ 4 t . i 'I - t F 1 i i i t v 3- .I ATTACHMENT 4 (page 15 of 46 pages) 2 ) 1 .+ w w = w
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- 5. 15 290 989 5.65 470 '16% step'decreas -
152,000 98 98-1557 5.80 '; 287 999 5.45-525 '14% step ine reas. 1155,000 10 0 99 1571 5.87 292 .999 - '5. 5 530 Initial j 123,000 79.4 84 1333 4.90 282 986
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j 101,000 65 71 1129 4.00 271 975 5.65 374< Initial 9 80,000 55 60 953-3.35 265 967,5 5.78 319 l 13. 6% step dec. ,'101,000 65 71 1129 4.00-270 975 5.63 373: 13. 6% step inc. 1 50,000 32 44 698 2.38.. 260 959 4.90 237' 39. ' % ramp dec. l 4 J i. . 57,000 ' ' 36. 8 48 762 2.59 248 960 5.83 '245 Initial 76,000 49 '59 936 3, 15 253 '972 5.73 295
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I 157,000 10 1 94 1492 5.45 275 1004 5.44 495 65% rainp inc. a ~ l 1
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............ ~.. t ATTACHKENT 4 (page 16 of 46 pages) i
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c - ~. O 3 7-7, 1 5. Turbine Trip {- i Date Performed: December 13, 1969 k, Reactor Conditions: 1591 Mwt, 541 Mwe 4 fl-61 ) hlb /hr Core Flow .,g
Purpose:
Determine reactor response and pressure peak during tubine trip. criteria: The pressure peak Ilmited to 1175 psig and thermal limits not exceeded. Results: The turbine was tripped by the master trip solenoid causing the stop valves to close in 0.1 seconds with consequent bypass valves ope *ing. Due to the power-steam flow mismatch reactor pressure increased'at an initial rate of 300 psi /sec for 0.1 seconds when an APRH high flux scram occurred. A pressure scram signal occurred 0.8 seconds later at 1060 psig. The pressure peak was 1060 psig and the heat flux increased less than 1% although the APRM's peaked at 150% 0.2 seconds after trip. The APRM's peaked at 150% 0.2 seconds after trip. The APRM's went to 0 in 0.4 seconds following the trip. Figures 10 and 11 show two of the Sanborn recorder traces. t ).l-i I ATTACHMENT 4 (page 19 of 46 pages)
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O d s ,' /f .i 4 , ll, [ 6. Generator Trip t M Date Performed: December 8, 1969 .e .i t Reactor Conditions: 1585 Mut, 526 Hue 6 g) 61 x 10 lb/hr core flow with 4 w6 recirculation ptanps running. .1 f
Purpose:
' Determine reactor response and pressure peak during generator j trip. o Criteria: The pressure peak limited to 1175 psig and fuel thermal limits not exceeded. Results: l .The generator was tripped off.the line by opening the main breakers. The turbine speed governor closed the control valves and the pressure' regulator opened the bypass valves within 0.5 seconds. The turbine speed' reached 1940 RPM before the stop valves were closed by a no-load turbine trip. 1 The reactor scranuned on high APRM neutron flux caused by the initial pressure ration of 50 psi /sec. The scram occurred 0.90 seconds.after the generator l trip and the APRM's peaked at 150%,'1.1 seconds after trip and were above 120% for 0.25 seconds. The pressure' peaked at 1055 psig 2.9 seconds afte'r ~ trip and no pressure scram occurred. A low level scram'came.l.7 seconds after the trip but level did not drop below 3 feet on the GEMAC level in-i, 'dicators. The LPRM heat flux ' indication increased less':than 2% before de - j creasing. The core flow increased from 160,000 gym to 182,000 gpm-due.to-the frequency increase on the auxiliary transformer. caused by generator overspeed and consequent increase in recirculation MG set speed. Since.the '4 j heat flux increased less than 2% and core flow increased, thermal. limits.' ] were not approached. 4 h Following the generator trip, a loss of auxiliary power occurred due to ~1 operator failure to transfer the auxiliary load to the startup transformer; the transfer is not automatic unless the generator fault relay is tripped. I' ATTACHMENT 4 (page 22 of 46 pages)
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culation pumps and feedwater pumps to trip.. Reactor water inventory was 'j maintained due to reactor isolation and the auxiliaries were transferred 8 manually to the startup transformer and restarted innediately. I 6 Two Sanborn traces are attached. (Figures 12 and 13). .] ..y 1 8 l \\: j l l l l I l i + t tt L v
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.s i O 3 '/ } 1 J 7. Pressure Regulator l1( Date Performed: December.8, 1969 6 .3 Reactor Conditions: a. 1518 Mwt, 530 Mwe, 61 x 10 lb/hr recire flow ir 6 [f b. 1135 Mwt, 385 Mwe, 38 x 10 lb/hr recire flow 6 '^ j c. 768 Mwt, 237 Mwe, 21 x 10 lb/hr i 6 e d. 623 Mwt, 205 Mwe, 14 x 10 lb/hr recirculation i flow with full power control rod pattern.
Purpose:
a. Determine reactor and turbine control system response to pressure regulator setpoint charges. b. Demonstrate stability of the power-void reactivity feedback loop. Criteria: a. The decay ratio of reactor process variables must b.e $ 1.0. i b. For acceptable performance the decay ratioiof reactor process variables is expected to be $ 0.25. l c. During the backup regu'.ator takeover, the MPR _should control thetransient such that the pressure is limited to 1045 psig. Results: a. 1518 Mut, 61 x 10 Id/hrRecirculationFlow The test was performed in three parts: first with EPR (electrical pressure { regulator) in service, then with the MPR (mechanical pressure regulator) i in service, and finally, the MPR takeover was; demonstrated with thva EPR { setpoint to 10 psi greater than the setpoint of the MPR. I With the EPR controlling pressure, the setpoint was first decreased-i } 5 psi then increased 5 psi. Following this demonstration, during which l all variables had a decay ratio less than 0.25 and the flux peak was { 1027., the pressure setpoint was decreased then increased 10 psi..This. transient was also better than a quarter damped with a APRM peak of 104% 1 from a stable condition of 987.. All setpoint changes were made at a rate i of 1 psi /sec. ATTACHMENT 4 (page 26 of 16 Pages) i ~ en
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6 v: 4 .x , d( a W I l - With the MPR controlling pressure the. setpoint was decreased then in-; creased 10 psi.. W e response was better than quarter danped with an - g; APRM flux peak of 101%. H e peak was lower than with the EPR because Q of' the slower setpod,nt change of the MPR. The EPR 10 psi setpoint charge ; J took seven seconds and the MPR 10 seconds. .g The takeover of the MPRvas demonstrated by increasing the EPR setpointi at 1: psi /sec. from 962 psig (6 psi below MPR setpoint). to 972 psig. s'T The response to transfer was better than quarter damped with'an APEM - j flux peak of 102% and pressure peak of 10M psig. i 0 b. 1135 Mwt, 38 x 10 lb/hr Core Flow. 976 psi. EPR in Control The power was reduced grom 1518 to 1135 Mwt by reducing the' recirculation flow to. 62.5% (38 x 10 lb/hr). At this time, the pressure perturoation was induced by changing the EPR setpoint down 10 psi and back up.10 psi. ~ Reactor pressure changes 10 psi in 7 seconds, with.a 3 psi overshoot. . Neutron flux experienced a 7% change and returned to its original value, steam flow and feedwater flow exhibited similar 6-8% fluctuations.- All-plants variation demonstrated better than a 0.25 decay ratio., c. 768 Mwt.-237 Mwe. 21 x 10 lb/hr Core Flow. 970 psin EPR in Control The power was reduced from 15,18 to 768 Mwt on flow control' and a pressure regulator setpoint change test was performed. The pressure - setpoint was decreased 7 psi then increased 7 psi. Reactor pressure changed 7 psi ) 4 with a 3 psi overshoot.. Pressure exhibited 0.3 decay ratio.on. pressure. setpoint increase and less than 0.1 decay ratio on the setpoint decrease.- Neutron flux varied 2% but restored itself to its original value within. one cycle. Feedwater flow varied 4% and was well damped. Steam' flow' varied 2%. Reactor water level, responding to the change in pressure fluctuated two inches. With the possible exception of. reactor. pressure, i all plant variables had a decay ratio less than 0.25 and'the plant. I l[ showed good stability characteristics at-these operating conditions. g I d. 123 Mwt. 205 Mwe. 14 x 106 lb/hr Core Flow. 959 asia EPR in Control { Natural Circulation b At the natural circulation condition following the five recirculation h. pusy trip from rated power, the EPR was tested by decreasing and in- '(, creasing the setpoint by 6 psi. Decreasing the setpoint resulted in 6.5 psi vessel pressure change with a 1. psi overshoot. He ratio of ATTACHMENT 4 (page 27 of 46 pages)
+emh i, *-~ 3 l Q O %, y-f ' 1 second to first overshoot was less than 0.25 but a.1 psi peak to pestk. - l 0.18 cps pressure oscillation was evident.. Neutron flux initially W ' dipped about 37,and returned to its initial'value with a decay ratio L i of less than 0.25. Reactor level increased by about an inch before l f/ . j returning to its initial value with no. apparent overshoot. Steam flow 1 initially increased by ~ about 2.97. of rated and settled out at a = slightly H L L - [- higher value with no apparent overshoot.- Increasing the setpoint by 6. l N psi produced essentially the same effect in the opposite direction, l
- t.
The test demonstrated good system stability at this low power, low flow I i { condition. j 1 1 8. Bypass Valves ^ Date Performed: December 8, 1969 Reactor Conditions: a. 1550 Mut, 520 W e,.61 x 10 lb/hr Core Flow 0
- b.. 1135 Mwt, 385 Mwe, 38 x 10 lb/hr c.
_768 We, 237 Mwe, 21 x 106 lb/hr' s d. 623 Mwt 205 Mwe, Natural Circulation 6 14 x 10 lb/hr.
Purpose:
a. Demonstrate the ability of the pressure regulator to minimize the reactor pressure disturbance during a large change in reactor steam flow. i-b. Demonstrate that a bypass valve can be tested for proper funo-tioning at rated power without causing s'high flux scram. Thedecayratiojofeachtransientvariablemustbe(1.0. Criteria: a. b. The decay ratio is expected f 0.25. g Results: Bypass valve No. 2 was tripped open by the valve test switch. The test opening valve was adjusted to open the valve in three seconds rather than 20 t seconds. The test was performed first with the MPR in service and then i> the EPR in service. The generator output decreased-to 500 Mwe from 520 Mwe when the bypass valve was opened. ATTACHMENT 4 (page 28 of 46 pages) i__.____
g e ;) %~.w !.w ',' w ~ w -- i t u 4 p n. W -- 29 f I 1 The-test with the MPR in service caused _a pressure peak of 2 psi and had' 7 negligible affect'on the APRM' flux signal. All process variables which g responded were better than 0.1 damped.: i s. 1135 Mwt. 38 x'106 lb/hr Recirc Plow. EPR'in Service t With power reduced from 1518 Mwt to 1135 Mwt on flow control, the by ; 4 pass valve No. 2 was tripped open as.above.. - The resulting perturbations j were very mild.. Power fluctuated only 1%, and was slightly more notice-I !1 able on closing the bypass valve than on opening. Pressure peaked.3 psi [ initially and had a_small affect on level, steam flow and feedwater flow. I All reactor variables had a decay ratio less than 0.25, and indicated-j good reactor stability. e b'. 768 Mwt. 21-x 10 lb/hr Core Plow. EPR in Control The power was reduced from'1518 Mwt to 768 Mwt on flow control. The . bypass valve test was done by tripping open bypass valve No. 2.. Reactor l pressure had a 2 psi peak. The APRM's' changed 27. but quickly stabilized-j at their initial. values.. The plant showed good stability characteristics as-all plant variables had a decay ration less than 0.25. ] ? 6 c. 623 Mwt 205 Mwe, 14 x 10 lb/hr Core Plow. 959 pain Reactor' Pressure EPR in Control e ) At the natural circulation condition, following the.five recirculation-1 pump trip from rated, bypass valve number 2 was tripped open resulting 'j in a mild transient.- Reactor pressure had a peak of about 1.7 psi and j a decay ratio of approximately 0.25. Neutron flux initially dipped by i about 1% and returned to its original.value with a decay ratio of less than 0.25. Reactor water level changed by less than an inchLand re-turned with no apparent overshoot. Steam flow was also better than 0.25 damped. Closing the bypass valve produced an opposite effect. _ The [. test demonstrated that the system is well damped and pressure is well 5 controlled by the EPR. ~ $4 } t b V u ATTACHMENT 4 (page 29 of46 pages) I 1
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...J - ) ~ t > 9. Feedwater Pumps ( Date Performed:- December 8, 1969 6 Reactor Conditions: a. 1550 Mwt, 520 Mwe, 61 x 10 lb/hr core flow 6 b. 1135 Mwt, 385 Hwe, 38 x 10 lb/hr h c. 768 Mwt, 237 Mwe, 21 x 10 lb/hr j d. 623 Mwt 205 Mwe, Natural Circulation Flow g ( 14 x 10 lb/hr.
Purpose:
Determine the effect of changes in subcooling on reactor power and pressure, also show that the reactor is stable for sudden changes of inlet subcooling initiated by the feedwater system. Show that that the reactor is stable for rapid changes in feedwater flow. Thedecayratioofeachtransientvariablemustbef1.0. Criteria: a. ~ The' decay ratio is expected to be f 0.25. b. Results: The reactor water level. was raised to 80 inches on the Yarways from 72 inches to allow the level setpoint changes to be made without causing a low level scram. 6 a. 1550 Mwt, 520 Mwe, 61 x 10 lb/hr Core Flow I The level setpoint change consisted of a 12 inch decrease, then a six inch decrease, followed by a six inch increase then decrease and then a nine inch increase. This program demonstrated the reactor i stability over a range of level control of 18 inches. I i i The six inch level decrease caused a. peak decrease in feedwater flow of i [ 0.6 M1b/hr from 5.8 M1b/hr, and decrease of APRM flux signal of 57. from l l 98%, 13 seconds after the feedwater flow change. k The six inch level increase caused a peak increase in feedwater flow 0.6 M1b/hr from 5.8 M1b/hr and an APRM increase of 4% from 987.,13 seconds after the feedwater flow change. b ATTACHMENT 4 (page 30 of 46 pages)
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- The 12 inch decrease caused a peak decrease in feedwater flow of 1.5 i
M1b/hr and an APRM peak of 6% from 98%, 13 seconds after feedwater j 1 g , flow change. The Sanborn traces, two of which are attached (Figures 14 and 15, show t that in all cases, the primary response has a decay ratio less than 0.1. .) 1 .a t F 6 +.! b. 1135 Mwt, 385 Mwe, 38 x 10 lb/hr Core Flow 976 psi, EPR in Control i The reactor power was reduced from 1518 to 1135 Mwt on flow control. A ) level setpoint change test was performed at that time. The actual level drop was eight inches which occurred in 30 seconds, and had a one inch overshoot. The feedwater flow responded by depreasing 1 M1b/hr from 4 M1b/hr in er seconds, then increased to the o'riginal valve in 35 sec-onds. The flow increased 1.2 M1b/hr on level increase, the APRM's dipped to 65% from 71% on level decrease and increased to 79% on level increase 18 seconds after the feedwater flow changed due to changes . in subcooling. The primary response of all process variables had a decay ratio of less than 0.1. 6 d. 623 Mut, 205 Mwe, 14 x 10 lb/hr Core Flow l At the natural circulation condition, following the 5 pump trip from rated power, a level setpoint change test was performed. The level setpoint was decreased then increased five inches. Both increase and decrease caused a feedwater flow change of 0.6 Mib/hr from an initial valu'e of 1.9 M1b/hr. The APRMs responded less than 2% of rated from an initial value of 40% for both decrease and increase. The maximum deviation occurred 65 seconds after the feedwater flow change. The primary response of all variables had a decay ratio of less than 0.1. The response of the feedwater control system in terms of time to j change level and the reactor power increased from 13 seconds at rated to 65 seconds at 623 Mwt. This change is due to the lower feedwater flow f' t and consequent greater feedwater loop response time. The controller l gain was not changed during this series of tests and was set to 80% proportional band and 0.4 repeats / minute. f ATTACHMElfr 4 (page 31 of 46 pages)
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m' y d?ge _,4 fQgff ~ l l .. + _ ^ L q. T, f ( 34 - 4-I p i ' 10. Flux Response To Rods q ~ Date Performed: December 8, 1969 Reactor Conditions: a. 1550 Mwt, 520 Mwe, 61 x 10 lb/hr core flow a-6 b. 1135 Mwt, 385 Mwe, 38'x 10 lb/hr core flow .f ~, 6 ~;
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768 Mwt, 237 Mwe, 21 x 10 lb/hr core-flow-I 6 d. 623 Mwt, 205 Mwe, 14 x 10 lb/hr core flow, natural circulation. i furpose: Demonstrate the relative stability of the power-void reactivity _ l feedback loop to rod movement. ~ t Thedecayratiomust;be$1.0. Criteria: Results: i 6 a. 1550 Mwt. 520 Mwe 61 x 10 lb/hr Core Flow Control rod 34-19'was continuously withdrawn from position 18 to 30 and inserted from position 30 to 28 to 18. Withdrawing the rod produced no obvious change in APRM flux or reactor pressure. The recorded LPRM-signal which was one fuel cell away--from the rod initially, increased' by about 47.. There was a corresponding increase-in the recorded heat flux which is the LPRM signal modified by a time constant of 7.8 seconds. i Steam flow increased by about 1% with no secondary oscillations. Level changed by less than 0.5 inches.- Inserting the rod caused no overshoot' in the LPRM signal. The test demonstrated that the system was'well ) -}- damped. ^ 6 [ j b. 1135 Mwt. 385 Mwe. 38 x'10 lb/hr Core Flow f The power was rr.duced from 1550 Mwt to 1135 Mwt by flow control and the
- E flux response to rod motion test was performed by inserting and then-
] withdrawing rod 34-19 from position 32 to 22 to 32. The rod motion pro. 'J( duced no apparent changes or oscillation in reactor steam flow or pressure. The recorded LPRM signal showed no-oscillatory behavior indicating that F the power m id reactivity loop is well damped at this reactor condition. ATTACHMENT 4 l (page 34 of 46 pages) 1 I l
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.T 0 J. ? t ; c. At the end point of the flow-power curve from 1600 Mwt, the test was pr-E formed by inserting rod ~26-27 continuously from position 30 to position i 10 and withdrawing it again. Inserting the rod produced a very mild. j transient. Neutron flux decreased by about 17.. The recorded signal .i from the LPRM which was two fuel. cells away from.the rod, decreased by (, approximately 17, with a corresponding decrease in simulated heat. flux. i J di No secondary oscillations were apparent,. indicating that the system is well damped with decay ratios of less than Q.1. Withdrawing the rod .g t caused a slightly greater swing in the above variables but with no secondary oscillations. I f d. The test was performed at the natural circulation condition following the five recirculation pump trip from rated power. Rod 26-27 was with - drawn from position 30 with no appreciable effect on the important plant variables, indicating a well damped system. 11. LPRM Calibration Date Performed: December 19-20, 1969 Reactor Conditions: 1585 Mwt, 550 Mwe, We - 160,000 gpm j
Purpose:
Calibrate local power range monitor system Criteria: The calibration will be such that the meters will read directly 2 in watts /cm, p Results: The calibration was performed in accordance with Startup Test Procedure j 'l
- 23. An octant symmetric rod pattern was established:. prior to the calibra-tion. Axial power shapes were obtained with the TIP system.
} NOTE: The calibration data and calculation sheets have not been included because the inspector does not believe they will serve any useful { purpose. t t k t ATTACHMENT 4 (page 35 of 46 pages) _m -_m___
a. e <. M ---.__me ~ s l ! h l t 12. Core Performance Evaluation ? ~ Date Performed: December 20, 1969 j f Reactor Conditions: 1585 Mwt, 550 Mwe, We = 160,000 gpm j Inlet temperature 519 F, 1001 psig Rx Pressure k
Purpose:
Evaluate core theftpal and hydraulic performance. Criteria:.a. Maximum {uelrodsurfaceheatfluxshallnotexceed104.3 watts /cm for all steady state conditions. b. MCHFR shall not be less than 1.50 when evaluated at 1207, of the operating power level. The MCHFR was evaluated for the above reactor conditions and found to be 2.94 (2.33 at 1207. overpower) in fuel assembly 27-18. Peak heat flux was found to be 2.94 (2.33 at 1207. overpower) in fuel assembly 27-18. Total peaking factor was 1.32 x 1.427 x 1.31 - 2.47 as compared with the techni-cal specification limit of 3.08 for scram setpoint of 120 per cene at 1600 Mwt. NOTE: This is just one of many MCHFR calculations. No anomalies were found during a review of the other checks. 13. Power Calibration of Rods Date Performed: December 9, 1969 Reactor Conditions: a. At start of test 1215 Mwt, 369 Mwe, We = 160,000 gpm. b. At completion of test 1550 Mwt, 525 Mwe, We = 160,000 gPm
Purpose:
Obtain for typical rod movements the effect on reactor power of ( control rod motion. I criteria: None ATTACHMENT 4 (page 36 of 46 pages) A
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y j n C .b i + -l - 37'- t -~+ Results:- ) e l During the power increase form 75'/. to 1007 power, data was taken as the S. rods were pulled in Sequence A., The sequence had been previously modified. .l such that rod groups 17 and 18 were not pulled beyosd position 24. The re- ' t. suits are shown on the attached Figure 16. i t ,e 1 I I i i i i .i r i k 1 ,I l. . ',h f~ e ATTACHMENT 4 (page 37 of 46 pages). 4 4 g
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J. 1 1: f - 9 l 14 Axial Power Distribution' i f l g Date Performed: December 19,1969;(one of many) l l Reactor Conditione: a. At start of data'taking l 1585 Mwt, 550 Mwe, We - 160,000 gym ' " 'q q; ' b. At completion of data taking l 1572 Mwt, 546 Mwe, We =.160,000 .-i L
Purpose:
Obtain axial power distribution for various reactor conditions. Criteria: None Results: This axial power distribution test was performed in conjunction with the l LPRM calibration test at 1600 Mwt. During the 2-5 hours required to take J the datp, power decreased 13 Mwt. The TIP's were'intercalibrated by placing [ each TIP probe adjacent' to LPRH chamber 28-25B and adjusting the TIP flux amplifier output to read the same as the LPRM. Several of.the.TIP traces- ] (Figures 17, 18, and 19) and a table (Table 3) of the axial power factors-are attached. The maximum axial peaking was 1.83 at core location 44-09, ~ 1 l j t L l t 6 } I i f l, s t i i I I ATTACHMENT 4 (page 39 of 46 pages)
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f - s-O ') i l. 1 15. Control Rod Pattern Exchange Date Performed: December 12, 1969 i
Purpose:
Perform a representative change in basic rod patter at a high re- ) actor power. N Criteria: Complete exchange without violating core thermal limits. 1 Results: Starting from 100*/. power rod pattern, the reactor power was reduced to 1236 Mwt. From this point, control rods were inserted and withdrawn in the order listed i on a rod movement schedule until the target rod pattern in sequence "B" was achieved. A low power density in the region of the rod exchange was main-tained, and excessive axial. peaking was not observed. The exchange proceeded quickly and was finished in two hours. The predicted rated power rod pattern for sequence "B" had considerably less rod notches inserted than the actual sequence "A" rated power rod pattern. The rod exchange sequence in the Oyster Creek Startup Test Procedure (CE22A 'i 2109) was modified to reflect this difference such that the sequence "B" rod pattern would have more rod notches inserted than the sequence "A" rated j power rod pattern. After the exchange was completed and flow returned to rated, the reactor power was 1350 Mwt. Sequence "B" rod groups 20 and 21 were then withdrawn in sequence from 12 to 20 and 0 to 12 respectively to increase power to rated. The sequence "B" rod 30-03 is inoperable and valved out of service due to a failure to couple. Therefore, this rod is out of sequence and at position 48 rather than 36. The MCHFR for sequence "B" full power rod pattern was 2.36 at 120*/. over-power in Fuel Assemble 05-24, i i I 4 4 i i ATTACHMENT 4 (page 44 of 46 pages) c
s. e O -d, P v / k E ' + l E. 16. Steam Separator - Dryer Performance il$ Dates Performed: December 26, 27, 28, 1969 l E hh Plant Conditions: a. Full power, full flow, various levels. I l N l ,jy b. 75% p6wer, reduced flow, various levels. l. Sp Water carryover in the steam to the turbine jl 0.2 weight per i. Criteria: a. ~ cent. I b. The steam carryunder, X, with the recirculation coolant flow 0.1$X$1.0weightpercent. I Results: The results of the carry under and garryover tests for 1600 Mwt and 1200 Mwt i at. reduced flow are tabulated below. All resbits meet the criteria. TMu 4 i Steam Separator - Dryer Performance Rx Power Recire Flow Level
- Carryunder Carryover o
. Mwe) (ib/hr) (ft) % W % W ( i 6 i 1220 42x10 6.60 0.16% 6 1220 42x10 7.25 0.16% 5x10 3 6 1600 61.4x106 7.1 0.145% 4x10 1600 61.1x10 6.7 0.09% 1600 63x106 6.1 0.12% 5x10-3 6 1600 61x10 5.3' O.10% 6 i-1600 61.2x10 4.72 0.04% I 1600 60.9x106 4.12* 0.08% 1.0x10-4 i
- Note: GEMAC indication reads lower than the Yarway level indicators.
The corresponding reading on the Yarways for the 4.12 ft. reading [ was 57 inches. The reactor level scram setting of 51 inches is r based on Yarway readings. b ATTACHMENT 4 (page 45 of 46 Pages) m v m w* ww ______m_
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1 '17. LPRM Response,
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Date Performed: December 20, 1969 and December 26', 1969 Reactor conditions: a. 1600 Mwt, 160,000 GPM b. 1520 Mwt, 160,000 GPM
Purpose:
Determine the response
- characteristics of the LPRM chambers at the typical flux levels encountered in operation and to-demonstrate t
the LPRM response to the neutron flux near the maximum steady-state y I heat flux limit. ga3 { Criteria: The core operating thermal limits will not be exceeded. E 2 Results: t The first set of data was taken at 1600 Mwt. The core power decreased 0.37. during the time required to take the TIP trace. The second set of data was taken at 957. power 1520 Mwt after power was decreased by inserting 20 notches in Sequence "A". To assure that thermal limits were not exceeded, the TIP was recalibrated in each channel to correspond to the "B" LPRM reading for that nannel..The maximum allowable h peak TIP reading was calculated for 07., 257. and 507. control density based on the LPP' to peak ratio h/m. For the 0, 25 and 50 percent control density cases f the peak. allowable LPRM/TIP remaing is 106, 100 and 100 respectively. For most TIP traces, the rods were withdrawn to within 107. of the peak. Then, the t current at'105 w/cm2 was determined for each TIP trace. l k I i 18. Electrical Output and Heat Rate i' -\\ I Date Performed: December 22, 1969 ii. I Reactor Conditions: 1590 Mwt, 533 Mwe, 160,000 GPM Core Flow f i
Purpose:
The purpose of this test is to demonstrate that the plant net t electrical output and net heat rate requirements are met. . J. t criteria: 1) Net plant heat rate must be S 10,600 Btu /kwh g{
- 2) Net electrical output must beh 515 Mwe Results:
Rff The warranty run was first started on December 17, at 1900 but was terminated 3 [ early on December 18. A valving error on a sample tank in radwaste had caused p. the condenser vacuum to be lost and resulted in a-reactor scram. The warranty run was started again on December 18, at 1900 and ran continuously for 100 hours 3 .t until 2300 on December 22, 1969. The net plant heat rate averaged about 10,203 -l Btu /kwh and the net electrical output was 531 Mw. According to procedure, heat j rate data was taken every 10 minutes during a two-hour period. The net plant heat during this period was 10,203 Btu /kwh, net electrical output of 531 Mwe j and core power of 1588 Mwt. + iTTACHMENT 4 (page 46 of 46 pages) ^ _1 I J i}}