ML20197A481
| ML20197A481 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/19/1978 |
| From: | Smarrella E AMERICAN ELECTRIC POWER CO., INC. |
| To: | |
| Shared Package | |
| ML20197A477 | List: |
| References | |
| NUDOCS 7810120067 | |
| Download: ML20197A481 (294) | |
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- I DONALD C. COOK I NUCLEAR PLANT I
.i l UNIT 2 l I l INDIANA & MICHIGAN POWER COMPANY g il '
' STARTUP lI ll TEST REPORT
- I I (Owgpgyggs I 781 p .2 2- FF6 7
i i I E l DONALD C. COOK NUCLEAR PLANT g UNIT 2 STARTUP REPORT E 9 PREPARED BY: Plant Technical Staff American Electric Power Service Corp. Nuclear Startup Services, Inc. DATE: September 19, 1978 I L' A U2 a : c .. _ ,^, COMPILED & REVIEWED BY: E. A. Smarrella I Technical Supervisor RECOMMEND APPROVAL: v 'A AEPj C -- Nuclear Division t .s d, M a_ln + - AEPSC -- Mechanical Division sw_/ - v I AEPSC -- Electrical Division E APPROVED BY: gW I D. V. Shaller Plant Manager
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TABLE OF CONTENTS
- REPORT SECTION PAGE NUMBER
1.0 INTRODUCTION
1-1 2.0
SUMMARY
2-1 0 3.0 INITIAL CORE LOADING 3-1 4.0 POST CORE LOADING 4-1 4.1 ICE CONDENSER TESTING 4.1-1 4.2 CONTROL ROD PERFORMANCE 4.2-1 1 4.3 REACTOR COOLANT SYSTEM TESTS 4.3-1 4.3.1 RCS HEAT LOSS & CAPACITY MEASUREMENTS 4.3.1-1 6 4.3.2 RCS THERMAL EXPANSION 4.3.2-1 4.3.3 RTD BYPASS LOOP FLOW VERIFICATION 4.3.3-1 4.3.4 RCS FLOW COAST DOWN MEASUREMENTS 4.3.4-1 5.0 INITIAL CRITICALITY 5-1 6 6.0 LOW POWER PHYSICS TESTIhd 6-1 I 7.0 POWER ASCENSION TESTING 7-1 7.1 PHYSICS TESTING 7. 3 - 1 7.1.1 POWER COEFFICIENT MEASUREMENTS 7.1.1-1 g 7.1.2 CORE POWER DISTRIBUTION 7.1.2-1 3 7.1.3 STATIC RCCA DROP TEST H-12 7.1.3-1 7.1.4 APDMS OPERATION 7.1.4-l 7.2 TRANSIENT TESTING 7.2-1 7.2.1 LOSS OF OFFSITE POWER TEST 7.2.1-1 7.2.2 10% STEP LOAD DECREASE (100 MWe) FROM 8 50% POWER 7.2.2-1 7.2.3 10% LOAD DECREASE-(100 MWe) FROM 100% POWER 7.2.3-1 i 7.2.4 50% LOAD DECREASE (550 MWe) FROM 100% POWER 7.2.4-1 7.2.5 NEGATIVE RATE TRIP TEST FROM 50% POWER 7.2.5-1
*7.2.6 TURBINE TRIP FROM 100% POWER 7.2.6-1 i *7.2.7 GENERATOR TRIP FROM 100% POWER 7.2.7-1 I
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TABLE OF CONTENTS (Cont'd) REPORT SECTION PAGE NUMBER 7.3 BALANCE OF PLANT TESTING 7.3-1 7.3.1 TURBINE GENERATOR START-UP TESTS 7.3.1-1 9 7.3.2 7.3.3 GENERATOR START-UP TESTS PLANT THERMAL POWER CALIBRATION 7.3.2-1 7.3.3-1 7.3.4 PLANT PARAMETERS AT POWER 7.3.4-1 8 7.4 CONTROLS / INSTRUMENTATION TESTING 7.4-1 7.4.1 FEEDWATER/ STEAM GENERATOR LEVEL CONTROL 7.4.1-1 5 7.4.2 7.4.3 AUTOMATIC REACTOR CONTROL INCORE/EXCORE DETECTOR CALIBRATION 7.4.2-1 7.4.3-1 7.5 MISCELLANEOUS TESTING 7.5-1 7.5.1 SHUTDOWN DEMONSTRATION FROM OUTSIDE I, 7.5.2 THE CONTROL ROOM PLANT RADIATION SURVEYS 7.5.1-1 7.5.2-1 7.5.3 PLANT CHEMISTRY HISTORY DURING POWER ASCENSION 7.5.3-1 8 7.5.4 7.5.5 LOOSE PARTS MONITORING POWER RANGE NOISE MEASUREMENTS 7.5.4-1 7.5.5-1 7.5.6 P250 COMPUTER CHECKOUT 7.5.6-1 7.5.7 EFFLUENT MONITORING SYSTEM CALIBRATIONS 7.5.7-1 7.5.8 REACTOR COOLANT FLOW MEASUREMENT. 7.5.8-1 APPENDIX A ACCEPTANCE / REVIEW CRITERIA A-1 APPENDIX B DISCUSSION OF APDMS TERMS B-1 9
- Indicates testing remaining to be completed.
t t I i i I 9
l i SECTION 1.0 ] INTRODUCTION The D. C. Cook Nuclear Plant consists of two pressurized water reactor units. The Nuclear Steam Supply Systems (NSSS) iI for both units are Westinghouse, supplied with a General j Electric turbine-generator on Unit 1 and a Brown Boveri turbine-t generator on Unit 2. The reactor for Unit 1 has fuel assemblies with a 15x15 fuel rod array, and is licensed for a rated thermal power of 3250 MWt. The reactor for Unit 2 has fuel assemblies 1 with a 17x17 fuel rod array, and is licensed for a rated thermal power of 3391 MWt. The nominal output of the turbine-generator on both units is 1100 MWe. The D. C. Cook Nuclear Plant is the , first nuclear facility to use the ice condenser reactor con-tainment system, which utilizes a heat sink of borated ice in a cold storage compartment located inside the containment for each unit. The construction permit was filed for on December 18, 1967, and was granted by the Atomic Energy Commission (AEC) on March 25, 1969. On October 10, 1972, a request fo'. extension of the construction permit was filed. The extension was granted by the 8 AEC on October 26, 1972. Subsequent milestones for each unit are as follows:
" Unit 1 Unit 2 Operating license issued for low i power tests 10/25/74 12/23/77 Operating license ammended for 81%
power' operation 12/20/74 NA Initial Criticality 1/18/75 3/10/78 5 Initial Turbine-Generator Syn-chronization 2/10/75 3/22/78 Operating license amended for 90% power operation (100% for tests) 3/30/76 NA Operating license amended for oper-ation at 100% rated power 5/28/76 4/28/78 First operation above 97% power 4/2/76 7/14/78 This report describes the startup/ testing activities com-mencing with Unit 2 initial core loading and continuing through initial criticality, low power physics testing and power ascension testing. The power ascension testing completed through August 1, a 1978, is described in this report. The report index indicates 1-1
.}
those portions of the power ascension test program that have yet to be completed. As the remaining tests are completed, the appropriate sections to this report will be issued. This report was prepared by the olant Technical Staff with t some assistance bv the American Electric Power Service Coroora-tion and Nuclear Startuo Services, Inc. The report has received internal clant review, AEPSC review and has been accroved by the Plant Manacrer. { i t n 1 i i i i t i e i i 1-2
SECTION 2.0
SUMMARY
8-Initial core loading began at 2225 hours on December 26, 1977, and loading of the 193 fuel assemblies was completed at 1643 hours on January 2, 1978. During core loading, problem I related delays were encountered in the following areas: Failure of the fuel transfer system air motor; containment airlock door interlock mechanism malfunction; reactor side upender limit switch I malfunction; high airborne activity measurements in the containment and auxiliary building; manipulator crane not always indexing properly; and an inadvertent signal causing a safety injection. Preoperational testing of the ice condenser system consisted f of an extensive program of component and controls checkout, combined with a continuing performance monitoring program. The h y Unit 2 ice loading was conducted utilizing the plant ice making and conveying equipment. The number of glycol chiller packages installed was increased from eight to ten to overcome the lack of system cooling capacity. Considerable equipment tuning was I required to produce ice of consistent density in the baskets. g During tests to measure control rod performance, it was g observed that the control rod drive mechanisms were misstepping. Exercising did not solve the misstepping problem. The reactor coolant was raised to about 5450F to promote thermal siphoning to clean the mechanisms. After the heatup the mechanisms no longer misstepped. The longest rod drop time of 1.6 seconds, which occurred during cold-full flow measurements, was well within the specified maximum of 2.2 seconds. Further post core loading tests were conducted to verify performance of the reactor coolant system and components. The heat loss and heat capacity of the pressurizer and the reactor I coolant system were determined. The main steam system heat loss, and maximum blowdown capacity were determined. The capacity of I the pressurizer spray valves, power operated relief valves, and a typical atmospheric dump valve were determined. During system heatup extensive thermal expansion measurements were taken. RTD h bypass loop flow was determined. Finally, reactor coolant flow coastdown, low flow alarm setpoints and reactor trip response 3 times were determined following the trip of one or more reactor coolant pumps. Initial criticality was achieved on March 10, 1978 at 1753 hours with an all rods out boron concentration of 1325 ppm. This was well within the design prediction of 1306 + 75 ppm. Nuclear heating was detected at about 3x10-7 amps on tee intermediate E l t 2-1
]
1 I l range channels, which corresponded to about 7x10-7 amps on the reactivity computer. The range for zero power testing was set i at 10-7 to 10-c amps on the reactivity computer. 8 Low power physics testing began at 0232 hours on March 11, j 1978, and was completed at 2350 hours on March 17, 1978. Para-I meters measured included: isothermal temperature coefficient versus boron concentration, critical boron endpoints, rod worths, minimum shutdown verification and boron worth. Also numerous flux maps were taken at a power level of about 2% with various I ~ control rod configurations, including an ejected RCCA (D-12) from the hot, zero power insertion limit. The all rods out moderagor coefficient at the beginning of life was positive (+1.30 pcm/ F). I With the Doppler coefficient of -1.86 pcm/ UF, the isothermal temperature coefficient was measured at -0.56 pcm/ F. The total worth of the control rod banks (CBA, B, C and D) was measured at I 4.36%, compared to the design value of 4.25%. The maximum F q with RCCA D-12 ejected frem the hot, zero power insertion limit was measured at 8.0, compared to a design value of 10.6 and the value of 13.0 used for the FSAR ejected rod accident analysis. The power ascension testing program was designed to provide data in the areas of core physics, plant transients, component startup, controls and instrumentation, chemical control and behavior and the plant's radiation environment. Power coefficient measurements were performed at 30, 50, 70 8 and 90% of full power. The power coefficient was determined directly from power changes resulting from reactivity insertions of control rods or boron and from data obtained by combining I' Doppler and moderator temperature coefficient measurements. The measured Doppler only power defect at 100% power was determined to be 1065 pcm, compared to the design value (BOL) of 1240 pcm. Although all data measured was within acceptance criteria limits, considerable scatter in the data was observed. Except for a core tilt of about 2% maximum at low power, the I core power distributions agreed well with des 3gn values. At a power level of 97.7% of full power the maximu.n assembly power deviation from design was 5.15% near the core periphery. The i average deviation was -0.07%. In order to minimize core axial offset and Fq, control bank D was operated in an almost completely withdrawn position. Two dropped rod tests were performed. Both tests were per-formed while the reactor was at 50% of full power. The first test was a static rod drop test (RCCA H-12) designed to demonstrate I that the most adverse effect on core power distribution from a single dropped RCCA would not result in an FAH of 1,58 or greater. 3 The measured FAH was 1.57. The worth of RCCA H-12 was measured E at about 0.07% compared to the design value of 0.12%. This test I 2-2 e ;
also demonstrated successfully that the incore and excore instrumentation could detect a single dropped RCCA. The second rod drop test was designed to demonstrate that release of two preselected RCCA (C9 and G3) would result in a negative rate trip of the reactor. Analysis of the data showed that at least three of the four excore neutron detector signals resulted in a negative rate bistable trip from the release of the two RCCA. In addition to the negative rate trip test a considerable I number of other plant transient tests were performed during power ascension testing. These consisted of a 10% power decrease from both 50% and 97% of full power, a 50% power decrease from 97% of full power and a loss of offsite power test from about 2% of full power. A turbine and/or generator trip test from full power have I yet to be performed. The data collected from all the transient tests verified that the plant was maintained in a stable, well I controlled condition during the transients and plant parameters were maintained within specifications. I Plant radiation surveys were performed at 0, 10, 30, 50, 70, 90 and 100% power levels to verify the effectiveness of the reactor biological shielding. Surveys were taken for gamma, fast neutrons and thermal neutrons. No unexpected radiation levels were encoun-I tered. Plant chemistry history was established by performing complete primary and secondary analyses at a maximum eight hour time interval i with increased frequency during power transients, reactor trips or other unusual circumstances. The higher than expected gross I activity in the reactor coolant of 0.003 pCi/cc at initial criticality was due to borating the system with recycled boric acid. Gross activity has increased to 0.65 pCi/cc at full power. The major isotopic constituent, Sodium -24 (greater than 90%), is believed I to have resulted from slippage through the plants demineralized makeup water plant. Iodine-131 activity, while not consistently present, was generally about 5.4 x lg-4 pCi/cc near full power, with a maximum activity of 1.2 x 10 pCi/cc following a power I transient. No primary to secondary leakage has been detected. A secondary water chemistry problem (high conductivity) was exper-I ienced when insufficient steam dump line baffling resulted in main condenser tube wastage, failure and leakage. The generator was sychronized with the AEP system on March 22, I 1978. output. Unit output has been about 70 megawatts short of the design Part of this deficiency has been attributed to low per-formance of the moisture-separator reheaters. Reasons for the remaining deficiency have not yet been identified. t 2-3 I l
SECTION 3.0 INITIAL CORE LOADING 8 PURPOSE The purpose of 2-PO-060-601 (Initial Core Loading) was to establish the conditions and sequence which governed the installation of the initial nuclear fuel (193 new fuel assem-blies) and their corresponding insert components in the vessel. I
SUMMARY
OF RESULTS Initial core loading operations began at 2225 hours on December 26, 1977. The core was completely loaded with 193 I fuel assemblies at 1643 hours en January 2, 1978. The total time for the loading was 162.25 hours. The maximum number of fuel assemblies loaded by one shift was 35 assemblies, with I most trouble free shifts loading between 22 to 27 assemblies. The average count rate, upon completion of core loading, for source range channel N31 was 11.1 and for N32 was 10.6 cps. Verification of the proper fuel assembly location was performed on January 2, 1978 using a video tape camera. DISCUSSION OF TEST , I Personnel assigned to the initial core loading were divided into two ten hour shifts. Each shift consisted of a reactivity engineer from I&M, three data takers (one each from AEPSC, Westinghouse, and I&M Operations), and two fuel I handling crews (one located in containment and the other located in the auxiliary building). The test engineer was located at the reactivity station in containment with one data I taker plotting ICRR curves. Two data takers were in the control room monitoring the source range channels and maintaining the reactor core and new fuel vault fuel location boards. I The containment fuel handling crew consisted of an SRO, a maintenance man operating the manipulator crane, a fuel transfer system operator, and a fuel inspector on the reactor vessel .g flange. The fuel inspector ensured that the assemblies were 3 located in their proper location, inspected for damage during transfer, and assisted the manipulator crane operator. In the auxiliary building, the fuel handling crew consisted of I a fuel handling foreman, a fuel transfer system operator, and four maintenance men. Initially the fuel assemblies were stored in the new fuel storage vault (144) and in temporary storage racks (49) in the auxiliary building. They were directly loaded from their I storage positions to the upender. This method of loading was used to save time by not having to load and unload assemblies into and out of the new fuel elevator. The auxiliary building crane was used to handle fuel in the auxiliary building. 3-1 Y
I Indexing of the manipulator crane was performed two weeks prior to the initial core loading using the Westinghouse supplied dummy fuel assembly. The indexing consisted of setting the dummy fuel assembly down in each core location around the core I periphery (against the core baffle) and each location in the cruciform through the center of the core running north-south and east-west. The manipulator crane and auxiliary building I cranes were load tested prior to core loading operations. Maintenance and operations personnel were trained in the oper-ation of the manipulator crane, auxiliary building crane, and fuel transfer system operation. Three temporary neutron detectors (BF3) were installed in core locations D2, M2 and L1 and their plateaus were determined I by use of an Americium-241 source. Background counts were then taken for the three temporary detectors (A, B and C) and the two source range channels N31 and N32. Core loading began with the insertion of the first primary source assembly into core position L-15. The rest of the initial I nucleus of fuel assemblies was then loaded. loading of the initial nucleus, ICRR data was Following the taken for each fuel assembly loaded (Fig. 3-3 to 3-7) for all detectors.1 The detailed sequence of loading steps is shown in Fig. 3-1 and I Fig. 3-2. g The boron concentration in the reactor vessel remained 5 above the 2000 ppm value required for the test. RHR loop water was sampled and analyzed at least once per shift,, The RHR temp-erature was also monitored by the plant process computer to insure that it did not exceed the 140'F limit. During core loading operations, the following significant delays were encountered:
- 1. The reactor side upender would not give a "down" interlock in spite of slack cable. A limit switch I had to be repaired.
- 2. High airborne activity in the auxiliary building, due to a gaseous release, caused a 26 hour delay.
- 3. The manipulator crane did not index properly all the I time., A loose set screw was found on the selsyn fine adjustment mechanism.
I
- 4. The fuel transfer system air motor failed. The transfer I canal was drained and the air motor was repaired. (
l
- 5. The interlock on an inner air lock door for contain- l I
ment malfunctioned. A guard was posted to insure containment integrity during core alterations while the repairs were made. J. Abrupt increases and decreases in ICRR are due to renormalizing detectors after repositioning of sources or changes in the boron concentration. 3-2
]
I 6. A spurious safety injection occurred on 1-2-78. The boron concentration after the safety injection was The detectors were renormalized and core I 2158 ppm. loading resumed. I f I 8 .I ' I i i ! i I l I I I I I f . I
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d I I I I I I I secr ou 4.o I ~ POST CORE LOADING I I I I I I I E 1 I I g 4-1
.._ _ _ _ _ __ _ _ . ___ _ _ _ _ _ __ _ J
I SECTION 4.1 ICE CONDENSER TESTING PURPOSE Startup testing of the ice condenser consisted mainly of I two applicable preoperational tests, one of the " Ice Condenser Refrigeration System" and the other of the " Ice Condenser Reactor Containment". The preoperational test procedure " Ice Condenser Refrigeration System" verified the ability of the refrigeration system to maintain the ice condenser at an acceptable temperature. The I test consisted of ensuring the acceptable operation of the various components by completing the following objectives: a) Verifying that the floor cooling pumps and defrost heater met or exceeded design specifications. I b) Assuring that the performance of the air handling units and the refrigeration system was adequate to maintain the ice bed at 15 F or less at ambient conditions. The primary purpose of preoperational test procedure " Ice Condenser Reactor Containment" was to verify the operability I condition of the system components required to function in the event of a loss of coolant accident. This was accomplished by verifying the following: a) The weight of the ice in the ice condenser satisfied Technical Specification 4.6.5.1.b.2. b) The lower inlet doors' initial opening torque and opening, closing and frictional torques when 40 degrees open satisfied Technical Speci fication 4. 6. 5. 3.1.b. c) The intermediate deck doors' initial opening force satisfied Technical Specification 4.6.5.3.2. ,I. The maximum opening force required to open each drain d) flapper valve satisfied Technical Specification 4.6.5.7.c. e) The lower personnel access door seal and alarm operated as required in Technical Specification 3.6.5.4. f) The instrumentation for monitoring the lower inlet doors l and ice bed temperatures properly verformed its indicating, alarming and switching functions. ,I i 4.1-1 l .
I g) The crane alarms and interlocks operated as required.
SUMMARY
OF RESULTS The performance of the floor cooling pumps was found to I exceed design specifications. In the first defrost, both the air handling units and floor defrost heaters performed quite well resulting in approximately 95% of the frost buildup removed with 4630 lbs. of condensate collected. The performance of the refrigeration system and its corresponding com-ponents was found acceptable for safety purposes for the existing I ambient conditions. However, when the ambient conditions are less than ideal, the system's effectiveness may decrease and the average ice bed temperature may be adversely affected to some extent. Two ice basket weighings were conducted. The first one was I to determine the ice basket weights immediately after loading; the second, shortly before initial criticality, was to verify compliance with the ice weight Technical Specifications and to establish a base of data obtained over a short time interval I for purposes of comparison with future ice weight data. original Technical Specification had required a higher minimum The ice weight than required for Unit 1, but allowed a longer I surveillance interval. The ice weights determined in the second weighing program failed to satisfy Unit 2's original Technical Specification. Consequently, a request was made and granted to I make the Unit 2 ice weight Technical Specifications similar to those for Unit 1. The ice weights determined in the second weighing satisfied the new Technical Specifications by wide margins. The operation of the lower inlet doors, intermediate deck doors and drain flapper valves were found to be within specifica-I tions after only minor adjustments. The lower personnel access door sealed and alarmed as designed.
.I All instrumentation switches and alarms performed properly after calibration and minor adjustments were complete. A modification to the equipment access doors was made which consisted
! of the removal of the inflatable seals and corresponding alarms l and switches and the installation of a permanent mechanical seal. DISCUSSION Ice was loaded into the Unit 2 ice condenser in 128 days, for l$5 the first time using the permanently installed ice making and conveying equipment. The only major problem encountered was the lack of cooling capacity in the system. Two additional glycol I 4.1-2 ' .I . .- . . _ 1
I chiller packages were installed making the total number available ten. Since the ice loading was being conducted during the summer when ambient conditions were far from ideal, the additional chiller capacity was required to maintain the average ice bed temperature within the desired limits. I Four chillers were utilized on each unit with the remaining two being used for ice making. Some minor changes were made to the ice loading equipment to increase its efficiency such as ice rake modifications and ice machine interlocks. This equipment I required an immense amount of tuning to produce ice of consistent density in the baskets. In an effort to minimize the detrimental effect of the construction activity in the ice condenser, the following pre-cautions were taken. a) All lower inlet doors were propped shut and their corresponding ports sealed off using temporary wood and herculite frames. b) Temporary airlocks were constructed on both upper plenum access doors and a selected lower inlet door I which was used for lower plenum access. Approximately eighty percent of the ice baskets were weighed I immediately after loading before the intermediate deck was installed. The initial load per basket averaged 1467.2 + 3.7 lbs. The purpose of this initial weighing was to supply immediate feedback to the loading crew, permitting the equipment to be I adjusted as required in order to maintain consistent weights. However, while the initial data was quite valuable for its intended purpose, it was found that these weights changed I appreciably after the cleanup operation, which called for clearing the flow paths around the baskets. Subsequently, another ice basket weighing program was conducted in February 1978. This program was conducted for two reasons: one, to obtain I a set of data consisting of the wieghts of 144 ice baskets obtained over a relatively short period of time to accurately provide a batis for comparison to the results of future weighing I programs; and two, to again verify that the ice bed weight was within Technical Specification limits. During the weighing, it was found that by weighing the desired sample of 6 baskets per I bay, 12 bays (one half of the ice condenser) were below specifica-tion. After increasing the number of samples per bay to increase the sample population and thereby reduce the statistical uncer-I tainity interval, all bays except 7 and 8 were within specifica-tions. Since water addition was the only means to satisfy the original Technical Specification requirements, its repercussions were weighed against the alternative of a Technical Specification I chante. The Technical Specification lower limit for Unit 2 of 1400 lbs. of ice per basket had been determined by adding a margin to the Unit 1 limit of 1220 lbs. in order to justify lengthening I the surveillance interval between weighings from 12 to 18 months, 4.1-3
I while retaining an acceptable confidence level. It was decided that the only practical resolution to this dilemma was to change the Technical Specification to a minimum ice I weight of 1220 lbs. and to include the spatial group ice weight requirements similar to those for Unit 1. The corresponding test interval was changed to 9 months. The I average ice load per basket determined by the second weighing program was 1430.7 + 3.8 lbs. I The lower inlet door seals also required additional attention. Since the original seals were installed years before cooldown took place and were badly damaged due to construction activities, all 48 seals were replaced with seals I of an improved design. After careful hand fitting, the final total lower inlet door air leakage measured in April 1978 was 78.7 cfm. I I I I I I i l I
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- I 4.1-4
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I SECTION 4.2 CONTROL ROD PERFORMANCE PURPOSE During cold control rod drops, it was found that many RCCA's were missing steps and not withdrawing completely. I This section discusses the CRDM testing program which was undertaken to define the problem and assist in the develop-ment of a solution.
SUMMARY
OF RESULTS The CRDM misstepping problet was found during the cold rod drop tests. After consultation with Westinghouse Electric Corporation personnel, an exercising / testing program was I developed and implemented. solved after the exercising. The misstepping problem was not It was then decided to heat up the RCS to promote thermal siphoning to clean the CRDM's. I After the heat up, the CRDM's did not misstep. All control rods were then dropped and their times met the Technical At no time, during the misstepping Specification limits. I problem was there any trouble with rod drop times of any RCCA's which had been verified to be at 228 steps by observing the glitch in the stationary gripper trace at 231 or 232 steps. DISCUSSION OF TEST Control rod drops started on 1-29-78 under cold no-flow conditions. After the drops, the fastest (N-ll) and slowest I (H-16) RCCA's were dropped 10 times each, for a repeatability test. During the repeatability tests for N-ll, it was noticed that the RPI was lagging behind the step counter and the drop times exhibited significant variability. After several in-I sertions and withdrawals, it was determined that the RCCA was misstepping and did not reach the stop prior to its drop. The failure to withdraw completely resulted in N-11 having I the shortest drop time. Upon consultation with Westinghouse Electric Corporation, I
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it was determined that the problem of misstepping rods was probably a mechanical (e.g. dirt in the CRDM), rat.hr. r than an electrical problem. A scheme was devised to chec:: for the number of missteps and to determine when the rod was at the I top of its travel. This test was dubb'ed the " glitch method". It consisted of driving the rods out to 228 steps and then taking a high speed visicorder trace of the stationary gripper I 4.2-1 a
I voltage as the rod was stepped out farther. Since the upper limit of travel, when the rods are cold, is 230 steps (?31 when the rods are hot, due to thermal expansion) an anomaly I occurs in the trace at 231 steps. This anomaly is a little spike in the stationary gripper voltage, which has been called a " glitch" (Fig. 4.2-1). Once the glitch appears, the rod is at 230 steps and any farther withdrawal is stopped. The glitch I will appear in subsequent steps, but the rods will remain at 230. Once the glitch method was devised, N-ll was stepped out to 230 steps, and checked for glitch. It was stepped back in to 228 steps, and dropped. Its time was then compared to the I other rods and it was not found to be the fastest rod. F14 was determined to be the fastest and it was dropped 10 times for repeatability. This test was repeated as described below. Cold full flow drops were then started. H-12 and H-14 were found to be misstepping. Rod drops were suspended while Westingbouse personnel explained the problem. Several of the I rods were .xercised by an automatic cycler brought by Westing-house reprecentatives. Then, all of the control rods were checked by the glitch method. Of the 53 control rods, five were found to be misstepping. A control rod exercise program was then developed to try to clear any possible dirt out of the CRDM's. A selected I bank was stepped out until the highest rod position indicator (RPI) read 200 steps. The rods were then stepped in until the first rod bottom light came on. This sequence was repeated I 10 tiraes. RPI readings vs. number of excursions was plotted (sample Fig. 4.2-2). Misstepping was observed by the decreas-ing top RPI reading on the misstepping RCCA, while the bottom I RPI readings of the other RCCA's increased. On the tenth excursion out, the visicorder was used to check for the glitch on each individual RCCA. It was found that the misstepping was of a random nature, i.e., RCCA's did not miss the same ,I number of steps each excursion and the same rods did not always misstep. Visicorder traces over the entire withdrawl of selected RCCA's also confirmed the random nature of the mis-I stepping. On 2-10-78, it was decided to proceed with the rod drops I and CRDM testing under hot conditions. Cold no-flow and full-flow rod drops were redone while checking for the glitch prior to dropping. The system was then allowed to heat up to I 547 F. It was hoped that the heat-up would promote thermal siphoning in the CRDM's to help clean out any possible dirt. The RCS was held at 547 F for 24 hours to maximize thermal siphoning prior to the closing of the reactor trip breakers. I Chemistry personnel sampled the RCS during this time for particulates, but could not definitely pinpoint any strange or high particulate concentrations. 4.2-2 I >
i 1 I l I After 24 hours ut 547 F, all of the control rods were checked for the glitch except C-9, which had a loose electrical connection at the Rer tor Vessel Head. All the other rods I showed the glitch at 232 steps - exactly where it should be. Westinghouse recommended that we exercise all the rods again as in the previous exercise. The C-9 connection was fixed and the exercise was completed with the data showing no missteps. I (Sample Fig. 4.2-3).
" glitch tests".
The RPI data was verified by visicorder I out. Hot full-fic*; and hot no-flow rod drops were then carried All rods were checked for the glitch and none showed any missteps. The rod drop times all met the Technical Specification I limit of 2.2 seconds from beginning of rod movement to dashpot entry (table 4.2-1). I I I I I I I I I I I 4.2-3 I o
Table 4.2-1: Control Rod Drop Times, 228 Steps to Dashpot Entry RCCA Cold No-Flow (sec) Cold Full-Flow (sec) Hot No-Flow (sec) Hot Full-Flow (sec) l H6 1.183 1.533 1.100 1.300 H10 1.183 1.533 1.108 1.317 1 F8 1.175 1.525 1.100 1.308 1.300 IK8 F2 1.167 1.183 1.542 1.567 1.117 1.117 1.300 B10 1.175 1.517 1.117 1.317 ; X14 1.175 1.567 1.108 1.325 l IP6 1.175 1.567 1.108 1.108 1.333 1.300 B6 1.183 1.542 1.167 1.567 1.117 1.325 IF14P10 X2 1.175 1.167 1.542 1.575 1.117 1.117 1.317 1.317 m H2 1.158 1.533 1.117 1.300
- 1. 75 1.525 1.117 1.300 5 88H14 1.183 1.525 1.117 1.308 P8 1.192 1.542 1.117 1.292 F6 1.167 1.517 1.117 1.292 IF10 1.183 1.533 1.117 1.292 K10 1.167 1.517 1.100 1.300 1.167 1.550 1.125 1.308 IK6 D4 D12 1.192 1.167 1.500 1.508 1.117 1.108 1.292 1.292 M12 1.183 1.517 1.100 1.283 1.183 1.517 1.108 1.292 IM4 H4 1.175 1.508 1.117 1.300 08 1.167 1.508 1.108 1.293 H12 1.167 1.517 1.108 1.292 i IM8 1.175 1.517 1.108 1.283 H8 1.167 1.525 1.100 1.292 D2 1.167 1.600 1.117 1.308 IB12 1.183 1.517 1.117 1.117 1.308 1.325 M14 1.175 1.567 P4 1.192 1.500 1.125 1.300 IB4D14 1.183 1.183 1.533 1.542 1.125 1.117 1.317 1.325 P12 1.175 1.508 1.117 1.300 1.183 1.500 1.117 1.342 IM2 G3 1.192 1.567 1.100 1.317 C9 1.175 1.525 1.117 1.292 1.107 1.525 1.117 1.300 I J13 N7 1.175 1.542 1.108 1.308 C7 1.158 1.517 1.108 1.283 !
G13- 1.183 1.508 1.108 1.317 IN9 J3 1.175 1.192 1.533 1.550 1.117 1.125 1.317 1.325 E3 1.167 1.567 1.108 1.300 1,183 1.508 1.108 1.308 I C11 L13 1.158 1.558 1.108 1.317 N5 1.158 1.550 1.108 1.325 I C5 E13 N11 1.175 1.175 1.192 1.533 1.550 1.550 1.117 1.117 1.117 1.300 1.308 1.308 1.120 1.600 1.133 1.333 I L3 I .
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I E i i i I SECTION 4.3 1 REACTOR COOLANT SYSTEM TESTS '4 I i .I I 1 I E I 4.3-1 g 1
I SECTION 4.3.1 RCS HEAT LOSS AND CAPACI't" !4EASUREMENTS i PURPOSE The purpose of reactor coolant system heat loss and capacity measurements was to provide data generally useful in thermal power determinations and to verify that system component thermal para-i meters were in accordance with expected design performance.
SUMMARY
OF RESULTS During the test the thermal data from system components was I measured and thermal response to system transients determined. The results are shown below and in Figures 4.3.1-1 through 7: Reactor goolant System Heat Loss at 547 F 1.0 MW Reactor Coolant System Effective Heat Capacity 0.68 MW/O F/hr 8 W Main Steam System Heat Loss Maximum Blowdown to Flash Tank 6.3 MW 342,000 lbs/hr (64.5 MW) Atmospheric Dump Valve Capacity (MRV-243) 522,000 lbs/hr (70 MW) Pressurizer Heat Loss 77 KW Pressurizer Heat Capacity 4.6 Amps / F/hr (18 Amps / psi / min) Pressurizer Spray Valve Capacity hW (NRV-163 and 164 Combined) Pressurizer Continuous Spray Valve 337,000 lbs/hr (916 gpm) Capacity (RC-122 and 123 Combined) 12,000 lbs/hr (34 gpm) g' Power Operated Relief Valve NRV-151 390,000 lbs/hr 5 Capacity Power Operated Relief Valve NRV-152 Capacity 422,000 lbs/hr i Fower Operated Relief Valve NRV-153 Capacity 441,000 lbs/hr DISCUSSION OF TEST During the first reactor coolant system heatup following initial I core loading, heatup data was taken with various numbers of reactor coolant pumps operating. Figure 4.3.1-1 shows the heatup of the reactor coolant using reactor coolant pumps with two-pump and four-8 pump operating modes. The heatup rates shown are 10.7 and 23.3 F/hr 3 respectively with two and four pump operation. The heat input of !' the coolant pumps minus the system heat loss equals the system heat-n up rate times its effective specific heat. With the measurement of heatup rates with two different pump heat sources (two and four pump l I 4.3,1-1 lt l . . . _ . .
I operation) we have two equations and two unknowns, namely system heat loss and effective specific heat. The RCS heat loss and heat capacity (corrected to 5470F) were 1.0 MW and 0.68 MW/ F/hr respec-I tively. Figure 4.3.1-2 shows the net reactor coolant system heat input 5 as a function of reactor coolant ternpera ture . This data provides a correction necessary for determination of core power from secondary plant thermal power data. With the main steam stop valves open and the main steam lines heated the system heat losses are somewhat larger. These losses were determined by adjusting the auxiliary feedwater flow to esta-blish constant water level in the steam generators, and then stop-ping feedwater flow and observing the decrease in steam generator water levels. The sum of the decreasing levels shown in Figure t 4.3.1-3 was 40%/hr, which corresponds to 6.3 MW for the main steam system heat loss. The levels in steam generators 2 and 3 decreased much more rapidly because only the atmospheric dump valves for those coolant loops were open. Another correction frequently necessary to determine plant thermal power is steam generator blowdown. Maximum blowdown to the start-up flash tank was determined by observing the decrease in steam generator water levels with maximum blowdown. The sum of the decreasing levels shown in Figure 4.3.1-4 was 409%/hr (after correction for one valve from steam generator no. 1 being closed), i which corresponds to 64.5 MW. Figure 4.3.1-5 shows the reactor coolant temperature and steam generator water level versus time when .tmospheric dunp valve MRV-243 was fully opened. The decreasing temperature corresponds to the
. equivalent of 522,000 lbs/hr at 1025 psia (70 Ma or 2% of full power).
t Using the two heat source method described above for the RCS measurements, the pressurizer heat loss was determined to b 77 KU (10.7 amps) and the pressurizer heat capacity was 4.6 amps /gF /hr (18 amps / psi / min). Using this data the capacity of the spray valves, NRV-163 and 164; the continuous spray valves, RC-122 and 123; and h the power operated relief valves NRV-151, 152 and 153 where deter-E mined. The combined capacity of NRV-163 and 162 was 916 gpm (337,000 lbs/hr). The combined capacity of RC-122 and 123 was 34 gpm (12,000 lbs/hr). I Capacities of the power operated relief valves were 390,000 lbs/hr for NRV-151, 422,000 lbs/hr for NRV-152 and 441,000 lbs/nr for NRV-153. The 34 gpm for the continuous spray valves is quite high. The flow was verified to be somewhat over I the nominal 2 gpm by observing the change in the boren concentration 5 in the pressurizer during a boration of the reactor coolant system. I Figures 4.3.1-6 and 7 show the response of the pressurizer pressure to opening of both spray valves and actuation of all heaters respectively. 4.3.1-2 i
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_.___ 4__ _ . . - _p. .__ _. m _. _ ap_ m _._.q y_. . _ _ ~} _p__. 8 . _a__ _ _ _ .. p _ . .__ _ M_S_d -l'3Mn..S,S3_8d._ . - . . + - _ - _ . _ . _ . _ . . - - _.._-...-._p y . .- y - 83218DS,S_3sd_ .. .- m ____. .__--.s.. y .....i- . .. __. --+y._- 4 - - . -__ . _.__L.__ . . . . _ - _ 1 . _.. _..{ .._.._ ..-......4-...... t . - o I SECTION 4.3.2 RCS THERMAL EXPANSION I PURPOSE
- 1. To verify that all reactor coolant components (reactor vessel, steam generators, reactor coolant pumps and pressurizer), piping and all auxiliary piping connected I to the main coolant piping expand and move without obstruction during initial heatup.
I 2. To obtain accurate expansion displacement measurements which provided information to calculate shim sizes to obtain minimum clearances on all major RCS support I structures and all rupture restraints.
SUMMARY
OF RESULTS During the initial reactor coolant system heatup, surveys were made at regular intervals to assure all piping and equipment were expanding without interferences. The surveys were conducted I up to and including 550 F. All interferences that developed were eliminated before the system temperature was raised to the next level. The final results show ghat the RCS could be at any temp-I crature between ambient and 550 F without interference problems. The second portion of the test consisted of taking cold and hot measurements between piping and its restraints and equipment and its restraints. These measurements were later used to I determine equipment and piping movements due to thermal expansion. From these movements, shim thicknesses were determined for all piping and equipment restraints. I DISCUSSION OF TEST The end result of the reactor coolant system thermal expansion preoperational test procedure was to assure that all piping and equiprent associated with the reactor coolant system could move I during system heatup without creating interference between piping and its restraints, and equipment and its supporting structure. I To assure that no obstruction developed during heatup, surveys were made before the initial heatup of all piping and equipment supports. All interference problems were corrected before the system was allowed to be elevated in temperature. Also during the I initial cold inspection, all possible interference problems were noted for closer observation as the system temperature was raised. The system temperature was raised at 100 increments starting at 250 F and a complete inspection was made at each temperature level up to 550 , the normal operating temperature. Each inter-ference problem was corrected as it was discovered, before the 1 4.3.2-1 I J
I I system temperature was allowed to continue up. No major com-I plications were encountered during the heatup procedure. discrepancies were corrected as they were found. All The second portion of the test consisted of determining I shim sizes for piping and equipment restraints. Piping systems monitored were main steam, feedwater, pressurizer surge lines, I safety injection and PER. Equipment restraints monitored were steam generator, reactor coolant pumps and reactor vessel. Before the initial system heatup, measurements were taken .I between the pipe and its associated restraint in four directions. Each restraint was marked in such a way that each measurement could be repeated at a later date. The equipment restraints were also measured prior to the initial heatup. After the system I had reached system operating temperature and stabilized, all restraint gaps previously measured were remeasured and this dimension was considered the hot position. This hot position would represent the location under normal operating conditions 8 for RCS piping. Following the completion of the hot functional testing, the I RCS was cooled down; at this point in time the restraint measure-ments were taken again. These measurements taken after the cool-down were considered the cold position, the position the piping I would maintain during a cold shutdown condition. The difference between the hot position and the cold position represent the movement of the RCS due to thermal expansion. The shim size was determined by leaving a pre-determined clearance and subtracting that clearance from the minimum clearance found during the hot and cold position measurements. Approximately 2000 measurements were taken during the completion I of this test. I I
- I I
't e 3 4.3.2-2 J
SECTION 4.3.3 RTD BYPASS LOOP FLOW VERIFICATION I PURPOSE The primary purpose of this test was to verity that the reactor ' I coolant temperature, as measured in the bypass loop, would be sufficiently representative of the reactor coolant loop temperature (based on the time required to detect temperature transients) to satisfy system control functions 8 and emergency safeguard actuations. More specifically, the objectives of the test were as follows:
- 1) to determine the flow rate necessary to ensure that the reactor coolant transport time in each RTD bypass loop (the amount of time I required for one molecule of reactor coolant water to travel from the point where the bypass loop leaves the reactor coolant piping to the last downstream RTD; including spares) is less than 1.0 second,
- 2) to measure each cold leg and hot leg RTD bypass loop flow rate to ensure that the transport time was less than 1.0 second and
- 3) to verify that the RTD bypass loop low flow alarm setpoint (at NFA 210, 220, 230 and 240) was 90% (*2%) of the lowest flow rate I observed at each reactor coolant loop's RTD bypass loop flow instru-ment (NFA 210,220,230,240) while the unit was in a normal hot standby condition.
I
SUMMARY
OF RESULTS Tables 4.3.3-1 through 3 below serve to suninarize the calculations made to determine the required flowrates to achieve a transport time of 5 1.0 second, the measurements taken to verify that a transport time of 5 1.0 second was being obtained and the verifications of the low flow alarm setpoints. Table 4.3.3-1
" Determination of Flowrate Necessary To Achieve A Transport Time of 1.0 Second" Cold Leo Bypass Loop Hot leg Bypass Loop Loop Total 1" Total 2" Total Required Total 1" Total 2" Total Required
# Pipe (ft) Pipe (ft) Vol.(ft 3) Flow (apm) Pipe (ft) Pipe (ft) Vol.(ft 3) Flow (gem) l i 1 0.0 10.5 .1634 73.4 14.7 6.3 .1512 67.9 2 0.0 13.5 .2101 94.3 14.9 6.0 .1473 66.1 16.3 9.7 .2099 94.2
- 84 3 0.0 0.0 11.1 12.3
.1727
.1914 77.5 85.9 14.7 7.6 .1715 77.0 l
4.3.3-1 lI l
Table 4.3.3-2
" Measurements of the Actual RTD Bypass Loop Flowrates" Cold Leo Bypass Hot Lea Bypass I Measured Actual Measured Actual Loop Flowrate Flowrate Flowrate Flowrate
# Fc (gpml F'c(gpm) Fs (gom) F'h(gom) 1 204 184 144 130 2 210 188 160 144 I 3 4
200 202 182 187 154 152 140 140 I Table 4 . 3. 3 -_3 I "RTD Bypass Loop Low Flow Alarm Verifications" Loop Total Bypass Computed Setpoint Actuation Point Reset Point Flow (gpm) (gpm) (gpm) I # (com) 1 314 283 283 283 2 332 283 283 286 I 3 322 283_ 282 285 4 327 283 285 290 I DISCUSSION OF TEST _ As a prerequisite to actual testing, field measurements were made to detennine the amount of one inch and two inch piping that was present in each hot and cold leg RTD bypass loop. These measurements were made prior to the I installation of the reactor coolant system insulation. As is summarized in Table 4.3.3-1, these measurements made it possible to compute the overall I bypass loop piping volume and, hence, the flow rate required to achieve a 1.0 second loop transport time. I Af ter the required flow rates were computed, actual hot and cold leg bypass loop flowrates were measured to verify a transport time of s 1.0 second. During this portion of the test, each reactor coolant loop's hot leg bypass loop flowrate was measured with its associated cold leg loop valved out since the flow I measuring instrument (NFA 210,220,230,240) is downstream of the point where each bypass loop's discharge ties together. Similarly, each cold leg loop measurement was made. These measured, individual, loop flowrates were then corrected to the actual flowrates that would be realized when the hot and cold I loops were valved into service together using the following equation: 8 4.3.3-2 l I J l
I I F'h = Ft I (1 +"Fc) F'c = Ft - F'h l 8 Where, F'h = actual hot leg flowrate Fh = measured hot leg flowrate F'c = actual cold leg flowrate I Fc = measured cold leg flowrate Ft = observed total bypass flow (hot , loop plus cold loop). I As can be seen in Table 4.3.3-2, the actual cold and hot leg loop flowrates were greater than the associated required flowrate (Table 4.3.3-1) to achieve I an acceptable transport time. The required RTD bypass loop low flow alarm setpoints were determ.i.ied by measuring each reactor coolant loop's total bypass flowrate on flow meters NFA 210, 220, 230 and 240. The setpoint was then determined to be 90% of the lowest of the four total flowrates. As Table 4.3.3-3 reveals, the setpoint was computed to be 283 gpm. In attempting to verify the required low flow I setpoint, either the hot or cold leg bypass loop flow was throttled to decrease the total flow seen at the NFA flow meter. During the throttling process, the alarm setpoint was verified. I I I I I
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4.3.3-3 : I j
I SECTION 4.3.4 RCS FLOW COAST DOWN MEASUREMENTS PURPOSE Flow coastdown measurements were performed to determine I the reactor coolant flow through the core versus time following the trip of one or more reactor coolant pumps. In addition, the measurements provided data to verify reactor coolant system I low flow alarm setpoints and reactor trip response times, and to verify that flow data are comparable to a prototype design for which natural circulation tests have been successfully completed.
SUMMARY
OF RESULTS Figures 4.3.4-1 and 4.3.4-2 show the time response of reactor coolant flow following the trip of one and four reactor coolant pumps respectively. When one reactor coolant pump was tripped (RCP#4), the reactor trip and low flow trips occurred at 0.097 and 1.483 I seconds after the pump trip respectively. When all four reactor coolant pumps were tripped, the four low flow trip signals occurred between 1.5 and 1.7 seconds after the pump trip. The I reactor coolant flow through the core during this time was between 90 and 91% of the full flow value. DISCUSSION OF TEST Figure 4.3.4-1 shows the time response of reactor coolant I flow following the trip of one reactor coolant pump. BGcause of the shape of the characteristic curve for the reactor coolant pumps, the flow through the remaining three operating pumps I increases as the pressure drop decreases as total core flow decreases. The flow through the loop in which the pomp has been tripped reaches zero 24 seconds after the pump was tripped. After I equilibrium conditions are reached the backflow through the idle loop reaches nearly 30% of the normal loop coolant flow. The pump was observed to stop turning at 58 seconds after the pump was tripped. Reactor core flow as a fraction of the full initial flow is shown in Figure 4.3.4-2 following the trip of all four reactor I coolant pumps. The measurement shows that the core flow is higher than shown by the design curve. Some core flow is available until about 140 seconds after the pumps were tripped. 4.3.4 ;
I Figure 4.3.4-3 shows the time and core flow fraction at which the low flow trips occurred.- All trips occurred between about 1.5 and 1.7 seconds following the pump trips, and at a I core flow between about 90 and 91% of the full initial core flow value. I The Trojan Nuclear Plant is a Westinghouse PWR nuclear steam supply system rated at 3411 MWT with 17x17 fuel assemblies, and as such has been accepted as a prototype design for Donald I C. Cook - Unit 2. The flow data obtained during the flow coast-down measurements were compared with the Trojan Nuclear Plant flow coastdown data. Natural circulation tests were conducted as well at the Trojan Nuclear Plant. Figure 4. 3.4-4 shows a comparison of the flow data obtained. Since the flow during I coastdown at the Donald C. Cook Plant - Unit 2 exceeded both the Trojan and Donald C. Cook acceptance curves, natural cir-culation flow measurements were not required for Unit 2. I I I I I I I I I I I . 4.3.4-2
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