ML20087B067

Co Rept 50-219/69-13 on 691203-05.No Items of Noncompliance Noted.Major Areas Inspected:Records of Power Ascension Test, Rod Drive Performance & Water Chemistry

ML20087B067
Person / Time
Site:	Oyster Creek
Issue date:	01/20/1970
From:	Caphton D, Robert Carlson, Dodds R US ATOMIC ENERGY COMMISSION (AEC)
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References
FOIA-95-36 50-219-69-13, NUDOCS 9508070291
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{{#Wiki_filter:. -. ,A; s U. S. ATOMIC. ENERGY COMMISSION REGION I DIVISION OF COMPLIANCE Report of Inspection CO Report No. 219/69-13 Licensee: JERSEY CENTRAL POWER & LIGHT COMPANY Oyster Creek 1 License No. DPR-16 Category B Dates of Inspection: ' December 3-5, 1969 Dates of Previous Inspection: November 4-10, 1969 Inspected by: C= I O Re3ctorN nspector (In Charge) Date R. T. Dodds, I 9t4WA no D. L. Da[hton, Reactor Inspector ~ 'te Da Reviewed by: /a b '" I '8# R. T. Carlson, Senior Reactor Inspector Date Pr5prietary 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 - December 4-5, 1969 (Prepared Sections D, E and F.) scope of Inspection: Review of records of power ascension a tests, rod drive performance and water chemistry, and tour of the facility.

SUMMARY

Safety Items - No items of safety significance were identified during the visit. 9500070291 950227 PDR FOIA DEKDK95-36 PDR

v; 8 i ' i t l Noncompliance Items - No items of noncompliance were noted during the visit. Status of Previously Reported Problems - 1. Action taken by the licensee to correct items of non-f y compliance. observed during previous visits was.not l specifically reviewed during this visit but.will be during the next visit, j 2. JC.has taken positive measures to assure that shift supervision of operators is being provided by JC rather [ than'GE. (Section B.l. ) i 3. Additional technical assistance has been obtained by JC for the power ascension' test program. (Section B.2.) .} .i 4. A nuclear engineer has been hired to replace Mr. Sullivan, j (Section B.4. ) ? 5. The control rod drive and water chemistry surveillance 7 programs have revealed no significant problems. (Section l E.1 and F.1.) j ~ ! Other Significant Items - t 1. The 1200 Mwt power ascension tests were completed on November 30, 1969. Several scrams have occurred because of problems experienced with. the level controls for the turbine reheaters and moisture separator drain and flash. tank systems. Full power operation at 1600 Mwt was finally achieved at 0555 hours on December 7,.1969. The full power warranty run was expected to be completed by December 20, 1969. (Section C.) l 2. Section C contains a summary of the reactor scrams that have occurred since the previous visit. Many, as expected, were the result of the power ascension test program. 3. The emergency procedures for the. facility instrument air. j system were reviewed and found to be adequate. (Section D.) 4. The results of a test to determine the assist effect the { control rod drive hydraulic pumps have on scram times is discussed in the report. (Section F.1.) .i i i r

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^ -3. 5. A change in nitrogen charging pressure (reduction) in the. control rod drive accumulators, previously incorporated T at Nine Mile Point, is'under consideration for this facility. (Section F.1.) l 6. One of the steam isolation' valves may have steam leakage at a high pressure differential across the valve.. l (Section S.2. ) ) 7. Section S contains' the results of power ascension. tests a completed since the previous visit. Management Interview - The results of the inspection were discussed ' t with-Messrs. McCluskey, Hetrick, Heward and Hess. Mr. Heward,-Vice-I Chairman of GORB, did. not enter into any of the discussions during the exit interview. Since no items of noncompliance or significant safety' items were identified, no commitments were proffered byLthe licensee. However, Mr. McCluskey was contacted on December 9, 1969,. to obtain the results of the steam leakage tests. The results:of this.and subsequent contacts are contained in CO Inquiry Memorandum No. 219/69-I, dated December 18, 1969. T DETAILS-A. Personnel Contacted: Personnel contacted during the visit included the following: I Jersey Central Power & Light Company (JC) a W. Hirst, Chairman, General Office Review Board R. Heward, Vice Chairman, General Office Review Board T. McCluskey, Station Superintendent j D. Hetrick,-Operations Supervisor D. Ross, Technical Supervisor. I. Finfrock, Maintenance Supervisor D. Kaulback, Radiation Protection Supervisor J. Maloney, Shift Supervisor S. Daalgard,- Chemistry Supervisor (Consultant) R. Stoudenour, Chemistry Supervisor - (Relief) I l l i + -m---w -m <w--s

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q ? i i -General Electric Company (GE) W. Hess, Site Operations Manager ' I W. Bibb, Operations Supervisor D. Diefenderfer, Principal Test Design and Analysis Engineer-F. Brutschy, Senior Chemist C. Hill,, Chemist l B. Administration and Organization i 1. Operating Organization Mr. McCluskey discussed the operating organization at Oyster Creek. JC has. requested GE to live up to the terms of:the contract regarding the Shift Foreman-Shift Supervisor-arrangement. All operator direction is now being provided by the JC Shift Foreman, even if it takes a little longer to complete a given task. This appeared to be the case,- based upon the inspector's observations while watching control room operations. Mr. M. Daniels, formerly at SEFOR and other GE. installations, has been in training to become the fifth Shift Foreman at Oyster Creek. He obtained his senior operator's license for Oyster Creek last September. Mr. McCluskey has not yet made a decision on whether to resume a five-shift operation or continue with the four shifts he now has. This is not imminent since JC has only eight licensed operators. Regardless, the foremen will; probably go on a - five-shift { schedule. 2. Technical Assistance L According to Messrs. Hirst and Heward, a number of additional engineers and technical personnel have recently been added to support the JC staff through completion of the power ascension test program. I 1 Either Mr. Hirst, Project Manager, General Public Utilities (GPU), or Mr. Heward, Nuclear Safety GPU, ChairmanL and Vice Chairman of GORB, will be at the plant week days until the commencement of commercial operations. Their function will e--n

.. _ n....-- s , be to assist the Station Superintendent as needed and to keep JC upper management informed in a timely manner of plant needs. They represent upper management. In addition, Mr. McCluskey has been given permission to purchase any i equipment necessary to get the plant operational as soon as possible. Assisting the Operations Supervisor as needed will be J. Barton, Startup Engineer for Burns and Roe. One of the following engineers from GPU has been assigned to each shift to assist the JC Shift Foreman as needed: Nuclear Engineer (fluid systems design), Project Nuclear Engineer (Senior Shift Test Engineer - nuclear navy), Senior Mechanical Engineer (Metropolitan Edison Company), Maintenance Supervisor (Metropolitan Edison Company). In addition, on each shift to assist the Shif t Foreman will be a consulting engineer with considerable nuclear background from United Engineers and Constructors. l In discussing the additions to the plant staff at Oyster creek, Mr. Hirst stated that JC upper management felt that they could best assist and support the operation by supplying minds since they could not supply hands. The big push is to have the plant in commercial operation by December 31, 1969. 3. Maintenance i Mr. McCluskey said that they now have extra maintenance coverage around the clock to keep up with the odds and ends ("on-the-spct") maintenance requirements. The three maintenance foremen were also on full-time shift work for the purpose of speeding up maintenance jobs. 4. Technical Engineers Mr. Robert Minue, Nuclear Engineer, has been hired as an associate engineer to replace Mr. Roger Sullivan, who left on October 31, 1969. Mr. Minue has a B.S. degree in Nuclear Engineering from the University of Tennessee and has completed about half of the requirements for a M.S. degree. He also had additional training at Oak Ridge National Laboratory in molten salt reactor technology,

~ i y , IBM computer programming school, and Purdue University Nuclear Fuels Management Course. 5 Mr. Minue worked as a Nuclear Engineer in the Y-12 Reactor Division Reactor Operations Gr oup. He then worked for Westinghouse APD as an Assistant Nuclear Engineer in the Nuclear Operations Department. His experience included the preparation of start-up test procedures plus zero power physics testing and analysis. Also, he worked on CVTR fuel management analysis, assisted in design of Saxton's Core III, and performed experimental nuclear tests (power distribution and power density) at Saxton employing various nuclear codes. C. Reactor Operations The 800 Mwt power ascension tests were completed on November 16, 1969. Several scrams have occurred because of problems experienced with the level controls for the turbine reheater and moisture separator drain tanks and flash tanks.* Once these controls had been modified, the power ascension program was continued at 1200 Mwt. The 1200 Mwt tests were c?mpleted on November 30, 1969. During the planned increase of reactor power to 1600 Mwt, the reactor scrammed at 1540 Mwt, again because of high level in the moisture separator drain tank. Investigation disclosed that the impeller for the large drain pump for the auxiliary flash tank was loose and was sitting in the bottom of the bowl. The next attempt to make 1600 Mwt on December 5, 1969, was foiled because of a high flux scram that was caused from cold water in the core, introduced by automatic actuation of one of the isolation condensers at 1030 psig rather than 1060 psig. The real problem involved the load limiter for the turbine controls that also was set too low for full power operation. Mr. McCluskey informed CO:I by telephone that 1600 Mwt was finally l achieved at 0555 hours on December 7, 1969. The full power warranty run was expected to be completed by December 20, 1969. The following scrams have occurred since the time of the last visit November 9, 1969 to 1200 hours on December 5, 1969. 1. Isolation Valve closure As expected, at 2050 hours on November 9, 1969, a high flux scram occurred at a neutron flux equivalent to 1080 Mwt following a steam line isolation valve closure during the simulated steam line break test (start-up test 12A) at

• Attachment 1 is a print of the reheater and moisture separator drain system.

Drain tanks 1-3 and 1-6 have a high level turbine trip function.

~ s ( ) 1 800 Mwt. The maximum reactor pressure during the test was 1060 psig, the peak neutron flux was equivalent to a reactor power level of 1170 Mwt, and the simulated heat flux only increased about 4%. [ 2. Loss of Auxiliary Power Test As expected, the reactor scrammed at 0610 hours on November 14, 1969, as a result of the loss of auxiliary power test. All systems functioned as expected. No significant problems were encountered as a result of this test. 3. Manual Scram The reactor was manually scrammed at 1217 hours on November 16, 1969, following the carryover tests at 800 Mwt because it was necessary to break condenser vacuum to put the turbine on the turning gear. 4. Turbine Trip During operations at 1200 Mwt, the reactor scrammed at 1042 hours on November 20, 1969, from a high flux spike (129% - 1630 Mwt) because of a turbine trip that was caused by high level in the moisture separator drain tank. The reactor pressure peaked at 1030 psig. No significant safety problems were identified as a result of this scram. 5. Generator Trip Test As expected, the reactor scrammed at 1845 hours on November 25, 1969, from high flux as a result of the generator trip j test at 1200 Mwt (380 Mwe). No significant safety problems were identified as a result of this test. 1 6. Main Steam Line Isolation at 1200 Mwt The reactor scrammed at 1629 hours on November 30, 1969,. i from a high flux trip as a result of the 1200 Mwt main steam { line isolation valve closure during a simulated steam line break test. The peak pressure achieved was 1110 psig ] (increased at a rate of 27 psig/sec) before being reduced by I the scram and isolation condenser operation (isolation condensers actuated at 1060 psig).

1 v.,.. .4--- 1 ( ^ e , 7. Turbine Trip at 1157 Mwt The reactor scrammed a't 0558 hours on December 1, 1969, from .I a high flux trip as athe result of a turbine trip at 1157 Mwt. The turbine trip occurred because of a high drain tank { level while attempting to put the second stage preheaters in service. I 8. Turbine Trip at 1370 Mwt The reactor scrammed at 2016 hours on December 1, 1969, from a high flux trip as the result of a turbine trip at 1370 Mwt. The turbine trip' occurred because of a high drain tank level because of pressure fluctuations in the main flash tank. 9. High Flux Scram at 1530 Mwt The reactor scrammed at 0106 hours on December 5, 1969, from a high flux scram as the result of cold water being introduced into the core by automatic actuation of one of the isolation condensers at 1032 psig rather than 1060 psig. Investigation disclosed that two of four pressure sensors were set low. The neutron flux peaked at 118% or an equivalent power level of 1850 Mwt. The pressure transient (1032 psig rather than 1000 psig) occurred because the steam flow limit on the turbine control had been reached (set conservatively) ; the pressure increased as reactor power was being reduced. No significant safety problems were identified as a result of the scram. D. Facility Procedures The instrument air system procedures were reviewed and discussed with Mr. Hetrick.* The facility and emergency procedures appear to be adequate. JC procedure No. 503 describes the loading and un-loading pressure set points for the two compressors. The automatic equipment and operator actions are covered by the procedure in the event of a complete loss of instrument air. The procedure specifies that the reactor will be shut down when the " air pressure reduces to 50 psig and the situation cannot be quickly corrected." i

• Memorandum, O'Reilly to Carlson, dated Sepember 23, 1969.

ei . ~, ,-. -_.:.~- ~ i 1 , E. Primary System 1. Water Chemistry The inspector questioned GE Chemists, Dr. Brutchey and C. Hill, and JC Chemists, S. Daalgard (Consultant) and R. Stoudenour, regarding the - reactor water chemistry program. Continued emphasis is being placed on the surveillance of reactor and feedwater chemistry. is data from a sampling audit of JC records of reactor water chemistry. gives metal analysis of feedwater, condensate and demineralizer effluents obtained during the sampling audit. Dr. Brutchey stated that no abnormal problems had been observed regarding the water chemistry. He stated that the feedwater crud had been averaging less than 5 ppb except during transient periods. Dr. Brutchey stated that oxygen samples taken of the steam ranged between 12 and 20 ppm. He stated that no 02 samples had been taken for reactor water, however, he estimated it to be between 0.2 and 0.3 ppm (using Henry's Law). Dr. Brutchey stated that Dr. Henry Helmoltz, a radiochemist will be at OC-1 during the warranty run and will have special equipment to determine the amount of 02 in reactor water. Dr. Brutchey stated that there was approximately 70 cfm air in-leakage into the condenser at this time. He stated that he expects this in-leakage to be reduced to about 10 to 20 cfm during normal operations after leak repairs are completed. Two samples taken of condensate on November 19, 1969, were analyzed for oxygen and the results ranged between 15 and 50 ppb of 0 - 2 F. Reactivity Control and Core Physics 1. Control Rods A test was conducted to determine the assist effect that the control rod drive system's hydraulic pump has on scram times. The test was conducted at reactor conditions of 1000 psig and approximately 1000 F. The data shown are the average times for 24 control rod drives with and

/ 's A ( ) 4 !' without (accumulator only) the pump. L-Pump Assist (PA) vs. Without Pump Assist (W/0) Rod Insertion 10% 50% 90% PA W/O PA W/O' PA W/O Scram Time 0.40 0.39 1.47 1.41 2.60 2.50 (seconds) Control rod drive 18-23 w'hich has exhibited long scram times

• has improved in its scram time performance.

The specific reason (s) for the changes in scram times are unknown. It is speculated that the amount of pluggage on the inner filter has varied. The following tabulation lists scram times recorded for drive 18-23. Insertion Time (Seconds) Date Recorded 10% 50% 90% 11/13/69 0.39 1.99 4.02 11/14/69 0.37 1.89 5.52** 11/15/69 0.41 2.24 4.56 { 11/20/69 0.37 1.53 3.01 11/25/69 0.34 1.33 2.87 11/30/69 0.38 1.60 3.26 12/ 5 /69 0.38 1.50 3.21 JC-GE have continued a surveillance program of control rod scram times in keeping with the guide lines established by the DRL lettei to JC.*** Except for one drive (18-23), no problems have been detected involving scram times and all other monitored scram times meet the guide lines of the DRL l letter. The following tabulation gives data from five { different scrams for the 26 monitored control rods.****

• CO Report No. 219/69-10, Faragraph D.1.

This drive is one of 8 R&D drives installed with an inner filter; the mesh size is 100 mic ron.

• • Longt it Time Recorded.
    • Lette :

Morris to Ritter, dated November 6,

1969, (Docket No.

50-2 9).

• • • • CO Report No. 219/69-10, Paragraph D.l.

.g . _. _ ~ 1 Insertion Time (Seconds)** Date of Scram 11/9/69 11/13 11/14 11/25 12/5 Reason For Scram Turbine Trip Test Pre-Test Powgg,goss gpggrggg{ gteamPresgurg a CausedF[uxScram Coordinate 10% 50% 90% 90% 90% 90% 10% 50% 90% 06-43 0.40 1.33 2.33 2.24 2.36 2.25 0.37 1.31 2.42 06-27 0.38 1.36 2.48 2.43 2.48 2.41 0.39 1.45 2.62 06-19 a 0.41 1.37 2.38 2.36 2.38 2.27 0.37 1.32 2.34 10-39 a 0.41 1.40 2.40 2.33 2.46 2.34 0.38 1.35 2.42 14-19 0.43 1.52 2.81 2.54 2.66 2.48 0.39 1.40 2.58 14-27 0.41 1.43 2.53 2.59 2.61 2.55 0.42 1.50 2.57 14-35 0.40 1.38 2.38 2.44 2.46 2.37 0.39 1.40 2.51 14-43 a 0.42 1.46 2.60 2.75 2.71 2.61 0.42 1.51 2.68 18-07 a 0.42 1.40 2.46 2.40 2.50 2.41 0.39 1.37 2.46 18-23*a 0.42 1.64 3.24 4.02 5.52 2.87 0.38 1.50 3.21 18-47 a 0.43 1.49 2.64 2.65 2.65 2.58 0.39 1.45 2.66 22-27 0.45 1.42 2.42 2.37 2.43 2.34 0.40 1.37 2.42 22-35 0.42 1.43 2.50 2.53 2.62 2.35 0.40 1.41 2.48 22-43 0.41 1.36 2.32 2.34 2.39 2.28 0.39 1.37 2.40 30-19 0.42 1.42 2.46 2.48 2.50 2.44 0.39 1.38 2.47 30-27 0.42 1.48 2.50 2.48 2.57 2.41 0.40 1.40 2.47 30-35 0.44 1.41 2.40 2.35 2.40 2.30 0.40 1.35 2.37 30-43 0.43 1.42 2.49 2.50 2.55 2.39 0.40 1.36 2.43 30-51 a 0.40 1.42 2.49 2.53 2.60 2.66 0.42 1.53 2.82 38-11 0.40 1.43 2.52 2.40 2.52 2.37 0.38 1.34 2.43 38-27 0.41 1.39 2.44 2.39 2.46 2.63 0.40 1.47 2.56 38-35 0.40 1.41 2.50 2.42 2.51 2.47 0.38 1.38 2.52 38-43 0.38 1.38 2.44 2.50 2.45 2.35 0.37 1.35 2.48 46-11 0.37 1.31 2.34 2.40 2.45 2.38 0.36 1.32 2.42 f 46-19 0.42 1.40 2.53 2.49 2.56 2.49 0.38 1.38 2.57 46-35 0.40 1.37 2.40 2.39 2.45 2.32 0.38 1.38 2.40 )

• This drive has given erratic scram time performance.

a. Drives with GE R&D inner filters, see CO Report No. 219/69-10, Paragraph C.S.

• • Technical Specification 3.2.B.3. specifies for average times:

10% = 0.70 sec.; 50% = 2.05 sec.; and 90% = 5.00 sec.

3 ,e g, _s....-.. ( . Messrs. McCluskey and Hetrick were questioned regarding a potential change by GE in the nitrogen charging pressure for control rod drive accumulators.* It was stated that { JC had not at this time, been contacted by GE regarding any change for OC-1. Mr. Bibb was questioned concerning any proposed reduction j in the initial nitrogen charging pressure to the control-rod drive accumulators and the potential for damaging component parts of the drive with the present nitrogen charging pressure at OC-1. 4 Mr. Bibb stated that at OC-1 the technique used to-prevent damage to the individual control rod drive components-was to limit the accumulator charging water to a maximum safe pressure of 1470 psig. He stated that this.is accomplished by limiting the pressure in the accumulator charging water. header. He stated that at OC-1 this technique has been used successfully to do the job. The inspector asked Mr. Bibb whether or not GE intended to make.a change at 0C-1. He stated that GE may make the change at OC-1, but at this time it had not been specified. He stated that GE may make the change because GE desires to use the best method and that perhaps the N2 Pressure reduction was a better method than was currently being used at OC-1, i.e., to m1nimize the possibility for damaging the drive. Mr. Bibb stated that the specific part in the control rod drive that had a past history of being damaged if an over-pressure occurred was an area at the necked-down portion of the index tube near the upper end of the tube. He stated that the thin wall of the tube could be distorted from excessive pressure. Mr. Bibb stated that OC-1 drives at this time had exhibited no signs of any problems relating to over pressure. S. Experiments and Tests The results of the power ascension test program to December 4, 1969 (completion of 1200 Mwt tests) were reviewed with Mr. Diefenderfer. The review included an examination of appropriate charts and data. The following is a brief summary of the results of significant tests that were conducted at 1200 Mwt and other test results not previously reported by the inspector. The opinions expressed are those of the GE test engineer and/or Mr. Diefenderfer.

• A change was made at Nine Mile Point.

Inquiry Memorandum No. 220/69-H.

_r L -a... s: (; ) t .7 ! 1-f 1. Loss of Auxiliary Power .}- The loss of auxiliary power test was conducted at 0615 hours I on November 14, 1969.. Reactor operating conditions at the if time were 300 Mwt, 65 Mwe, reactor pressure - 1000 psig, and recirculation flow - 120,000 gpm. The purpose of the test-I was to determine reactor transient performance and to l I determine electrical system performance during a loss of auxiliary power transient. The results of the test were as follows:- t a. The reactor protection system motor generator coasted i down to the 54 cps trip. point in 9.9 seconds, scramming the reactor and causing-a " full" isolation. b. Reactor water level increased due to flow coast down abat 3 to. 6 inches and then decreased to an indicated low of 4.6 feet. [ c. Reactor pressure gradually increased to.1040 psig in 3 minutes and then dropped because of emergency condenser operation. d. The emergency condenser operated several times following the isolation. 1 e. Both diesel generators started automatically and were l syncronized with the two emergency buses in 14.5 seconds. i' f. All required auxiliary loads were picked up as specified in the tables for the test procedure. g. The reactor protection system buses were not picked up for approximately 10 minutes.because the operator failed to reset the " low frequency". breaker'. This did not effect operation of the plant as the reactor had scrammed and isolation had been initiated.- h. Natura1' recirculation flow following the test was. l 12,000 gpm. j i. Three local power range monitors (LPR' Ms) failed due to voltage transients on the reactor protection system j panels.

~ s f .. j. According to Mr. Diefenderfer, no significant anomalies were identified as a result of the test. 2. Main Steam Isolation valves The main steam isolation valves will be tested at 800 and ~ 1600 Mwt for proper operation and for steam leakage. Tests } to simulate a " steam line break outside the drywell accident" were performed at 800 Mwt and 1200 Mwt. The 800 Mwt isolation test was conducted at 0200 hours on November 9, 1969. The four isolation valves closed in 6.7 - 9.4 seconds. The steam leakage test showed that the pressure in the 24 inch downstream steam lines and 30 inch header increased from 440 psig to 660 psig within five minutes, apparently from leakage through steam isolation valve V-1-8 (NS 03B). Little or no leakage was experienced through the other three steam isolation valves. According to Mr. Diefenderfer, this test only indicated that there may be_ steam leakage and that the leakage rates would be measured during the isolation valve trip test at 1600 Mwt. JC was contacted on December 9, 1969 subsequent to the 1600 Mwt isolation valve trip test to obtain the results of these leakage rate measurements. It appears that there is definitely steam leakage through at least V-1-8. The results of the leakage tests are contained in CO Inquiry Memorandum No. 219/69-I, dated December 18, 1969. The results of the simulated steam line break tests at 800 and 1200 Mwt were as follows (no anomalies were identified during the tests): ) a. November 9, 1969 Reactor conditions - 886 Mwt; O Mwe; 1000 psig; 160,000 gpm recirculation flow; all steam to condensers. Reactor pressure - Increased to 1060 psig in 9.5 seconds. Neutron flux - Increased to 130% in 6.4 seconds; scram actuated at 120%. l

.y n swaw e-- a ~ \\ (y i \\ '" Reactor Water Level Decreased from an indicated 6 feet to 2 feet following the reactor scram. simulated Heat Flux Increased 5.3% (indication change from 38% to 40%) 'l f LPRM Increased from 36% to 56% before scram. b. November 30, 1969 Reactor Conditions 1302 Mwt; 200 Mwe: 1010 psig; ] 160,000 gpm. recirculation flow; and some steam to condenser. i Reactor pressure 1060 psig in 8 seconds; peaked ) at 1110 psig on a pressure increase rate of 27 psig/sec. Pressure rise terminated by scram ) and automatic isolation condenser operation at 1060 psig. Neutron Flux Flux peaked at 117%. Reactor Water Level Dropped from an indicated 6.2 feet to 1.4 feet in 15 seconds (void collapse, high pressure and scram). Fuel Surface Heat Increased 4%. Flux Isolation Automatic operation of steam isolation valves at 850 psig; performance satisfactory. 3. Recirculation Pump Trips The recirculation flow-reactor power transient resulting from recirculation pump trips were measured and evaluated at power levels of 800 and 1200 Mwt (also to be demonstrated at 1600 Mwt). The results of the tests were as follows:

hps.A. _.u..uwL.. u,.._,.. s _ m, e; +.> A: s +. -

• m i

, (g--. i t ) +, L 4 7 .16 - .l a. . November 15, 1969 (Rerun of test conducted on October 2,. j J '1969) 1 Y i + 1 i -Reactor Conditions - 870 Mwt; 246 Mwe; 1000 psig;lRe-1 A; l ' )t circulation flow - 165,000 gpm. c-Results - All five recirculation pamps. were tripped. p The MCHFR calculated"just' prior to the trip.was.6.7 at-1 core s positon 13-10.which is greater than the ' initial 1 predicted condition. A comparison. of the1 actual flow-power coastdown with predicted coastdown showed that the' } ~ actual power coastdown was ' greater'while-the flow coast-1 down was less 'than the pre-calculated results. Therefore,. :t'[ the minimum MCHFR was 6.2, again greater than predicted.

)

The natural circulation flow was about 16.8%, (27,000 j gpm) and the final power was about 285 Mwt (17.8%'of l rated)..The final value of generator. output was'un-obtainable.because of a turbine. trip caused byLhigh- 'l moisture separator' drain tank' levels. Closing the discharge valve on recirculation loop "C" decreased total-indicated core flow to.25,000 GPM. Power decreased to' approximately 280 Mwt (17.5% of rated). Evaluation.of' MCHFR'at-the four loop natural circulation condition yielded a value alio.0. l I Reactor power indicated by the APRM's drops faster than recirculation flow while heat fluxclags due to the time J constant of the fuel. After the pumpsEwere tripped, .i reactor pressure dropped approximately 13.5 psi causing an increase in' level with a consequent decrease in feedwater flow. Decreasing steam flow also contributed l1 to the decreaseLin feedwater flow. i b. November 25, 1969 Reactor conditions - 1220 Mwt; 395 Mwe; 160,000 gpm core flow. Results - The trip of all five recirculation pumps at' 1200 Mwt resulted in a transient that closely. resembled the re-evaluated predicted power and flow transient. j Both power and flow decayed at a slightly slower rate j . - ~..

g,,. _-w [. Ji:- u._ n, '~ . (j ') , than-anticipated.- The MCHFR values (4.. initial decreasing to a minimum value of.3.75) were:at all times equal to or greater than the predicted values. I i' Attachments 4, 5, 6 and 7 show the observed:results. t The natural circulation flow rate was'34,000 gpm or J 21% of rated' flow, whereas the power it:ve1ed off at ~ 1. 36% of rated or 575 Mwt-(150 Mwe). 4. Recirculation Flow Control' Plant response '(load following) to recirculation-flow control at an initial reactor, power level of 1200 Mwt'was determined with the following results: Date of Test: November 25, 1969 Initial operating Conditions: 1163 Mwt; 387 Mwer 1002 psig;1 ~ 158,000 gpm flow. Feedwater Flow Power Level Flow-Temperature m Mwt Mwe M lb/hr OF 156,000 100 1160 390 4.2 275 126,000 81 , 1010 330 3.5 265 110,000 70.5 885 295' 3.2 260 94,000 60 780 262 2.8-255 80,000 51 710 226 2.4 245' 58,000 37 620 196 1.9 235-5. Turbine Trip The turbine trip from 1200 Mwt' (1229 Mwt; '390 Mwe) ' operating : power level was performed on an unplanned basis on November 20, 1969 because of high levels in the moisture separator drain tank. Wie reactor steam flow-bypass capacity mis-- match caused a high flux spike of 129% (100% E 1260 Mwt) which exceeded the scram point for less-than 2.5 seconds.

w \\ ~_ j g . Reactor pressure peaked at about 1030 psig. The emergency condensers were not actuated during the transient. The reac-tor water fst reached the low level scram set point of 3.94 { feet about 4 seconds after the trip due to void collapse from pressurization. 1 -t' 6. Generator Trip 1l The generator trip test at 1200 Mwt (1150 Mwt) was performed on November 25, 1969. The neutron flux peaked at 127% (100% a 1260 Mwt) in 0.6 seconds, reactor scram at 120%. Reactor pressure peaked at 1028 psig. Heat flux only increased 3%. The steam flow increased momentarily from 4.2 to 7.0 M lb/hr when the bypass valves opened. No significant problems were identified as a result of the generator trip test. 7. Pressure Regulator Tests of the electrical and mechanical pressure regulators at 1200 Mwt (395 Mwe) and at reduced flow and power levels were performed on November 25, 1969. The tests did not dis-close any anomalies. The relative dampening factors during the pressure transients were always less than 0.1. 8. Bypass Valves The test to demonstrate that a bypass valve can be tested j for proper operation at 1200 Mwt (395 Mwe) was performed on November 25, 1969. Bypass valve No. 2 was opened in 3 seconds, once with the mechanical pressure regulator controlling pressure and once with the electrical pressure regulator controlling pressure. In both instances the reactor power level fluctuation was only 2% and the relative dampening factor was less than 0.1. No anomalies were identified during the test. 9. Feedwater Pumps j i Various transient tests of reactor water level changes (shutting off a feedwater pump and starting a feedwater pump) were conducted on November 25 and 30, 1969. The initial level change tests showed the relative dampening factor to i

t / s:s' ('l- ) / be higher than anticipated but still acceptable. How ever, after several adjustments to the level controller, the i-relative dampening factor for a 6 inch step level change was less than 0.1. 4 The shutting off of a feedwater water pump did not j significantly effect the reactor operating parameters as I all were stable within 30 seconds. The maximum level change was 6 inches and the maximum flow change was 10%. The restart of the pump was not discernible. 10. Flux Response to Rods The flux response to rod movement test performed at 1200 Mwt on November 25, 1969, did not disclose any anomalies. Control rod 22-07 was used for the test. The signal-to-noise ratio of adjacent LPRM's was about 6.5 (Criteria O). The relative dampening factor was less than 0.1- (Criteria 5 0. 2 5).. 11. Average Power Range Monitor (APRM) Calibration Calibration of the APRM's was being checked at least daily. A calibration performed on November 20, 1969, at an indicated level of 94-99% (1210 Mwt) demonstrated the true power level to be 98% based upon 100% being equivalent to 1265 Mwt. All eight APRM's were reset to indicate 98%. 12. Core Performance Evaluation A core performance evaluation performed on November 24, 1969, at reactor operating parameters of 1236 Mwt, 407 Mwe, 1002 psig, 140 F subcooling, and 59.5 M lb/hr core flow showed a minimum critical heat flux ratio (, CHFR) of 3.54 (2.94 at M 120% overpower scram point) for assembly 37-10. The maximum 2 heat flux was 78 watts /cm, 13. Axial Power Distribution TIP traces for LPRM calibrations at 1200 Mwt performed on November 24, 1969 showed the following axial power factors:

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i LPRM Axial Power Factor- . Channel A B C D I j.. 44-33 ' O.632 1.<68 1.03 0.565 { .j 44-09 0.598 1.92 1.10 0.398- { i 44-17 0.605 1.84 .l.02 0.430 i i-36-25 0.628 1.74 1.13 0.436 j 44-41 0.565-1.93 1.04 0.382 i i 12-41 0.630 1.91 1.07 0.407 20-33 0.677 1.61 1.23-0.462 l 28-09 0.607-1.57 1.11 0.626 j 12-09 0.575 1.99 1.11 -0.562 ) 20-17 0.685 1.63 1.15 0.432 i 20-09 0.624 1.68 1.02 0.559 28-25 0.680' 1.64 1.20 0.429 36-41 0.647 1.81 1.06 0.413-36-33 0.677 1.60 1.38 'O.456 j 44-25 0.612 1.57 1.11 0.663 36-17 0.677 1.59 1.14. 0.443 28-17 0.627 1.74 1.11 0.423 36-09 'O.579 1.85 1.04 0.445 l 28-33 0.616 1.76 1.19 0.435 36-49 0.694 1.75 .992 0.475 { 28-41 0.625 1.67 1.17 0.530 i 20-41 0.650 1.79 1.10 0.470 i 20-49 0.731 1.64 1.01 0.625 28-49 0.590 1.60 1.11 0.755 l 12-17 0.610 '1.82

1. ' 9 0.435 l

0 20-25 0.632 1.73 1.16 0.439 12-33 0.620 1.78 1.09-0.457 12-25 0.624 1.63 1.17 'O.535 04-33 0.731 1.59 1.02 0.627 04-17 0.701 1.76 .991 0.454 04-25 0.573 1.58 1.12 0.733 14. Steam Separator-DryeYi Steam separator-dryer tests performed at 1200 Mwt, 160,000 gpm 1 core flow, 4.55 M lb/hr steam performed-on November 30, 1969,. showed a water carryover content of d0.0026% and a ' steam carryunder of 0.14 - 0.16%. (Criteria: water carryover - 15 0.2%; steam carryunder X - 0.lti X $1.0). i 'll

.. ~.. - +. D ~d 4 mls= mism 1 SEPARATOR stp m g W NIST M 80857 W ' 1-3 34 D ARATOR SEPARAT0st l I-2 33 ~] y-t-99 v.3 103 v.3 102 - ppg W** TATER 5 nAsn ~ h' ~ v4-95 1A.3 gg.3 gg,3 ~ T 7 ~ 3 Qd { < tt ~~ ~ca 4 RDEATER = 14 v4-10 v4 3 g y.g g y 6 M 14 - ', A9 O

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7 '/ . ' i JERSEY CENTRAL POWER & LIGHT CO () ') CO REPORT NO. 219/69-13 4 ATTACHMENT 3 Date '11/19/69 11/29/69 12/1/69 12/3/69 s L . j System FN FW FW FW C DE . Iron ppb 34 16 27 48 13 35 Copper ppb <1 41 0.50 41 <1 <1 Nickel ppb <1 1.9 3.8 41 <1 4.1 Chromium ppb 1.5 <1 4.25 7.0 4.2 4.~ 2 Reactor Power Mwt 510 Reactor Power Mwe 110 Feedwater FW = Condensate C = Demineralizer effluent DE = i P f ) ) I I

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