ML20212H510
| ML20212H510 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 03/03/1987 |
| From: | Cantlin J CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | |
| Shared Package | |
| ML20212H480 | List: |
| References | |
| NUDOCS 8703060129 | |
| Download: ML20212H510 (162) | |
Text
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THE CLEVELAND ELECTRIC I ILLU MIN ATING COM PANY '
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FORWARD This Startup Report for Perry Nuclear Power Plant, Unit 1, covers the period from initial fuel load through the nine months following initial criticality (June 6, 1986 to March 6, 1987) as required by Regulatory Guide 1.16 and Perry Technical Specifications Section 6.9.
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I i TABLE OF CONTENTS Title Page SECTI N 1 - DESCRIPfIN 1 1.1 Introduction 2 1.2 Perry Plant Design 3 1.3 Startup Test Program 7 1.4 Startup Test Program Scope 8 1.5 Startup Test Program Organization 9 1.6 Startup Test Program Administration 11 I 1.7 1.8 Test Plateaus and Test Conditions Quality Assurance Coverage of the Startup Test Program 13 18 SECTION 2 -
SUMMARY
19 2.1 Summary of Testing to Date 20 2.2 Summary of Key Events 21 2.3 Startup Test Program Chronology 22 SECTION 3 - STAR'IUP 'IEST INSTRUCTIONS 26 3.1 Startup Test Instruction Format and Content 27 3.2 Startup Test Instruction Abstracts 30 SECTION 4 - STAR'IUP TEST RESULTS 49 4.1 Chemical and Radiochemical 51 I 4.2 4.3 4.4 Radiation Measurements Fuel Loading Full Core Shutdown Margin 56 63 67 4.5 Control Rod Drive System 69 4.6 SRM Performance and Control Rod Sequence 80 t 4.7 IRM Performance 81 l 4.8 LPRM Calibration 90 l 4.9 APRM Calibration 91 4.10 Process Computer 94 4.11 Reactor Core Isolation Cooling 101 4.12 Selected Process Temperatures 103 I. 4.13 Water Level Reference Leg Temperatures 105 4.14 Vibration and Thermal Expansion Tests 106
- 1) System Expansion I 4.15
- 2) Drywell Piping Vibration
- 3) BOP Piping Expansion and Vibration Feedwater Control System 122 4.16 MSIV Function Test 123 4.17 Safety Relief Valves 126 4.18 Recirculation Flow Control - Valve Position Loop 127 4.19 Reactor Internals Vibration 129 I 4.20 4.21 4.22 Reactor Water Cleanup System Off-Gas System 133 138 ERIS (Emergency Response and Information System) 142 4.23 Turbine Building Closed Cooling System 146 ii
TABLE OF CONTDES (COfC. )
Title Page SECTICN 4 - STAR 1UP TEST RESULTS (CONT. )
4.24 SPCU Performance Test 150 4.25 Concrete Temperature Survey at Low Power 151 l 4.26 Steam Seal 152 l 4.27 Condenser Air Removal System 155 4.28 Loose Parts Monitoring System (LPMS) 157 Baseline Data I
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SECTION 1 .
l DESCRIPTION ,
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1.1 INTRODUCTION
tis report consists of a summary of the Startup Test Program performed at Unit 1 of the Perry Nuclear Power Plant in compliance with Regulatory Guide 1.16, Revision 4, and Perry Technical Specifications paragraphs 6.9.1.1 thru 6.9.1.3. This report covers the nine acnth period following initial criticality but also includes Startup Testing beginning with pre-tuel load activities.
This report addresses each of the Startup Tests identified in Chapter 14 of the FSAR which have been completed to date and includes a description of the measured values of the operating conditions or characteristics obtained during the test program with a comparison of these values to the Acceptance Criteria. Also included is a description of corrective actions required to obtain satisfactory operation.
h is report also provides a brief description of the plant, a description of the Startup Test organization and administration and a brief abstract of each Startup Test Instruction.
Since the Startup Test Program is still in progress, Supplemental Startup Reports will be submitted at least every three months until the program is completed.
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1.2 PERRY PLANT DESIGN 5
U 2e Perry Nuclear Power Plant is a two unit nuclear powered electrical generating station with Unit 1 cmplete and undergoing Startup Testing.
The plant is operated by the Cleveland Electric Illuminating Ccrnpany (CEI) and is located near Lake Erie in Lake County, Ohio. The plant site is approximately 35 miles northeast of Cleveland, Ohio and 21 miles southwest I of Ashtabula, Ohio.
a Unit I has a boiling water reactor nuclear steam supply system as designed I and supplied by the General Electric Cmpany and designated BWR/6, with a Mark III containment. We rated core thermal power is 3,579 Mwt. The unit rated net electrical output is 1,205 Mwe.
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The balance-of-plant was designed by Gilbert Associates, Inc., Reading, -
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Pennsylvania, as architect-engineer. The reactor building ccmplex includes the drywell, containment vessel, and shield building. We drywell is a reinforced concrete structure enclosing the reactor pressure vessel and the main reactor coolant recirculation loops. Outside the drywell, there is a dcmed cylindrical steel containment vessel supported by a reinforced concrete, steel-lined foundation mat, surrounded by a reinforced concrete shield building.
The Unit I reactor core consists of 748 GE 8x8 fuel bundles of which 656 are new " barrier" fuel designed by GE to minimize fuel failures due to pellet-clad interactions.
The recirculation loops consist of two 7935 hp notor-driven recirculation pumps. Rese pumps can be operated at 1800 or 450 RPM. Recirculation flow is continuously variable using hydraulically operated flow control valves located on the discharge of each pump. A total of 20 jet pumps 6
located inside the reactor vessel are used to develop 104 x 10 lbn/hr of coolant flow through the core at rated power. Rated steam flow is 15.4 x 6
10 lbm/hr.
Four main steam lines route the steam output of the reactor vessel to the E main turbine, a cE turbine rated at 1250 Mwe. Each steam line has two E main steam isolation valves (Atwood-Merrill) for fast-acting isolation.
The full-arc admission turbine has one high pressure stage, the discharge of which is routed to four moisture / separator reheaters. Three low pressure turbine stages receive the output of the moisture / separator reheaters which is then exhausted to the main condenser.
Fast acting steam bypass valves can bypass up to 35% of rated steam output directly to the condenser.
Unit 1 uses a 510 foot tall natural draft cooling tower to cool the circulating water in the main condenser. Makeup water is taken from Lake Erie by service water pumps.
The condensate /feedwater systems use four stages of pumps and six stages of feedwater heating. The main feedwater pumps consist of two 60%
capacity turbine driven pumps and one 20% capacity motor driven pump. The fourth stage of feedwater heating is a direct contact heater, the other five stages consist of shell/ tube heat exchangers.
Independent emergency core cooling systems consist of:
- 1) One high pressure core spray system which uses a motor driven pump to deliver coolant into a spray header directly over the core. A dedicated and independent diesel generator (General Motors) supplies emergency backup power to this system.
- 2) Three low pressure coolant injection systems which use motor driven pumps to deliver coolant directly into the core shroud.
- 3) One low pressure core spray system which uses a motor driven pump to deliver coolant into a spray header directly over the core.
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Two divisional diesel-generators (DeLaval) provide emergency backup power E to tha inw pramtre emargency core cooling systeras. 'h>o of the three low E pressure coolant injection systems can also function to remove decay heat as loops A and B are equipped with heat exchangers.
An additional core cooling system is the Reactor Core Isolation Cooling System (RCIC) which uses a single-stage turbine pump driven by reactor steam to deliver coolant to the reactor vessel.
There are nineteen safety relief valves (Dikkers) connected to the reactor vessel of which eight are used by the Automatic Depressurization System to rapidly depressurize the plant, if needed, to a pressure within the capability of the low pressure emergency core cooling systems. A large (approximately one million gallons of water) suppression pool inside con-tainment is used to quench the discharge of these relief valves. In the event of a Loss of Coolant Accident, high pressure steam is diverted through vent holes in the drywell to this pool for quenching.
To monitor the flux level in the core there are four channels of Source Range Monitors, eight channels of Intermediate Range Monitors and eight channels of Average Power Range Monitors. The power range monitors use a total of 164 local power range detectors to develop a representative measure of core flux. A moving " gamma-tip" Traversing Incore Probe is used in each of the 41 local power range monitor instrument strings to measure the axial flux shape.
l l A total of 177 GE locking-piston control rod drives, using boron as the control poison, are used, in addition to changing core flow, for power changes and for scram shutdowns.
Perry is equipped with two VAX 11/780 computers to drive the Emergency l Response Information System (ERIS) of which the Safety Parameter Display l
System is a part. They provide operating information and are used for transient monitoring (such as startup testing).
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I A third VAX 11/780 is interfaced with the plant process enmputer (Honey-fl wil 4400) and is usM for running the NSSS core monitoring programs and U for off-site dose projections during an emergency.
A restricted license from the NRC permitting operation of Unit i up to 5%
power was received on March 18, 1986 and a full-power license permitting operation up to 100% power was received on November 13, 1986.
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1.3 STAR 7UP TEST PROGRAM E
a The Startup Test Program includes those plant testing activities to be performed during and following initial fuel loading. It also includes those startup tests used to gather baseline data for the Startup Test Program prior to fuel loading. The Startup Test Program culminates with the completion of the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, full-power Warranty Run.
The Startup Test Program is one of three phases of testing described in I Chapter 14 of the Final Safety Analysis Report (FSAR). The other two phases of testing are the Initial Checkout and Run-In phase and the Preoperational Test phase which were performed following completion of construction and construction-related inspections and tests. They demonstrate, to the extent practical, the capability of structures, systems, and components to meet performance requirements and to satisfy design criteria.
The Startup Test Program consists of Startup Test Instructions (STI's) which cover fuel loading, precritical tests, initial criticality, low-pcwer tests, and power ascension tests. These tests confirm the design bases and demonstrate, to the extent practical, that the piant will operate in accordance with design and is capable of responding as designed to anticipated transients and postulated accidents as specified in the FSAR.
I In addition to the three phases of testing described above, the overall test program for Perry includes Surveillance Test Instructions (SVI's).
These instructions are used to demonstrate the proper operation of interlocks, setpoints, and other protective features, systems, and equipment as required by the Perry Technical Specifications. Selected surveillances used to support startup testing are included and scheduled in plateau instructions used for administrative control of testing.
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1.4 STARTUP TEST PROGRAM SCOPE i a !
O The Perry Startup Test Program was designed to comply with the require-ments set forth in the following Regulatory Guides as described in the Perry Final Safety Analysis Report:
Regulatory Guide 1.68 - Rev. 2, Initial Test Program for Water-Cooled Nuclear Power Plants Regulatory Guide 1.68.1 - Rev. 1, Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants Regulatory Guide 1.68.2 - Rev. 1, Initial Startup Test Program to Demon-strate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants The Acceptance Criteria for the Nuclear Steam Supply System Startup Tests were based on General Electric supplied Startup Test Specifications, MPL Item Number A42-3610.
The Acceptance Criteria for the Balance of Plant Startup Tests were based on Gilbert supplied Test Specifications. These are now controlled by CEI's Nuclear Engineering Department.
Since Perry is regarded as a prototype plant for the BWR/6 238 inch diameter reactor vessel, the test program includes a comprehensive reactor internal vibration program in compliance with Regulatory Guide 1.20, Com-l prehensive Vibration Assessment Program for Reactor Internals during Pre-l operational and Initial Startup Testing.
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1.5 STAR'IUP TEST PROGRAM ORGANIZATICH E
a The Managers, Perry Plant Operations and Technical (PPOD/PPTD) Depart-ments, are responsible for approving Startup Test procedures, instruc-tions, and results and for authorizing entry into Test Plateaus and Test Conditions. The Manager, PPTD, is responsible for the Startup Test Program beginning at fuel load and continuing through the Startup Test Phase.
The Plant Operations Review Committee (PORC) is responsible for review and recommendation of approval of Startup Test procedures, instructions, and results and for recommending entry into Test Plateaus and Test Conditions.
The Startup Test Program Director reports to the Technical Superintendent and is responsible for the development and implementation of the Startup E st Program. He is responsible for preparation and approval of the startup test schedule; preparation, review and approval of startup test procedures and results; conduct of startup testing; and coordination of interfacing organizations in support of the Startup Test Program.
The Startup Test Element Supervisor, under the direction of the Startup Test Program Director, is responsible for: the preparation and revision of Startup Test Instructions and procedures; Startup Test Change Notices; maintaining and updating the test schedule; reviewing and submitting the test results; tracking the resolution of test exceptions; and for directing and coordinating the activities of test personnel.
The Test coordinators are responsible for coordinating the preparation and performance of Startup Tests with other plant activities, for directly supervising the Test Directors, and for ensuring that startup tests are completed and analyzed expeditiously.
The Test Directors are responsible for the conduct and performance of individual Startup Tests, preparing and analyzing test results, main-taining the Test Directors Chronological Test Log and the Startup Test I
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1.5 STARW P TEST PROGRAM ORGANIZATION - (CONT.)
E E Performance Log, obtaining Unit Supervisor permission to start a test, and for keeping the Test Coordinator and Unit Supervisor informed of test and equipment status.
The Test Engineers are responsible for assisting the Startup Test Element Supervisor, Test Coordinators, or Test Directors as required.
The General Electric Site Operations Manager (SOM) provides advice to the Startup Test Program Director during the Startup Test Program. The SOM reviews and approves Startup Test Instructions, assists in evaluating test results, and advises the Startup Test Program Director in all matters related to the Startup Test Program.
General Electric Startup Personnel under the direction of the GE Site Operations Manager, will provide assistance and support to the Perry Plant departments (PPOD/PPTD) commencing at fuel load. The Startup Test Op-erations (SIO) Engineers, under the direction of the GE Operations Super-intendent, interface with and provide technical direction to the Operations staff. The Startup Test Design and Analysis (STD&A) Engineers, under the direction of the Lead STD&A Fngineer, interface with and provide technical direction to the Startup Test Organization. The Control and Instrumentation Services Engineers (C&I) provide technical direction on
{ matters within their area of expertise.
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l.6 STARTUP TEST PROGRAM ADMINISTRATION E
E Listed below are a few of the administrative controls to arrive at a consistent and thorough Startup Test Program. These activities are covered in detail in the Startup Test Program procedure, Plant Administrative Procedure (PAP)-1104.
l To the extent practical, the administration of the Startup Test Program utilizes normal plant procedures and organizations for its implementation.
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- a. A test exception occurs whenever test results fail to satisfy an acceptance criteria or whenever a step or steps of the test instruction cannot be satisfactorily completed during the conduct of the test.
- b. The Test Director may initiate a TER for any plant or equipment problem discovered during testing if he feels it is appropriate.
Test Exception Reports (TER) are written to document the description and resolution of all test exceptions as well as the subsequent I actons required to close out the exception.
- 2. Test Serial Numbers (TSN) - Each individual test performance has an unique Test Serial Number (TSN) assigned to the test.
Test Directors are responsible for completing the Startup Test Per-formance Log and assigning serial numbers. The log is completed, and a serial number assigned prior to the start of the performance of a test.
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I 1.6 STAR'IUP TEST PROGRAM ADMINISTRATION - (CONT.)
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" 3. Startup Test Results Packages - At the completion of the test, all associated information is assembled into a package and submitted to the Startup Test Element Supervisor for review. Copies are then sent to Quality Assurance, Nuclear Engineering, and the General Electric =
Site operations Manager, and then to the Plant Operating Review Committee for approval review. The PPOD/PPTD Managers provide final approval. -
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I 1.7 TEST PIATEAUS AND TEST CONDITIONS
" Testing and power escalation sequences are divided into the following distinct test plateaus:
Test Plateau Definition Open Vessel (OV) Plant Conditions cannot exceed those for TC OV.
Heatup (HU) Plant Conditions cannot exceed those for TC HU.
Low Power (LP) Plant Conditions cannot exceed those for TCl.
Mid Power (MP) Plant Conditions cannot exceed those for TC2 and TC3.
High Power (HP) Plant Conditions cannot exceed those for TC4, 5, 6, 7, and 8.
The plant Test Conditions are specified regions on the Power / Flow Operating Map within which most startup testing is performed. These regions are defined in Figure 1.7-1, Sheets 1, 2, and 3, Power / Flow Map and Test Condition Region Definitions. The Power / Flow Operating Map is provided as an aid to understanding these Test Condition Region Definitions.
A Test Plateau and Test Condition review is performed prior to escalating above the maximum power associated with that condition. The following items must be completed prior to each review:
I a. All Startup Tests scheduled for the current Test Plateau or Condition j have been implemented, the analyses have been completed, and the test results have been reviewed and approved.
- b. All Test Exception Reports affecting tests scheduled for the current Test Plateau have been resolved.
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1.7 TEST PLATEAUS AND TEST CONDITIONS - (COfff.)
- c. Exceptions to the above requirements must be authorized by the Plant Operating Review Committee.
Listings of all tests approved to be run during each Test Plateau are con-tained in specific plateau Startup Test Instructions. These instructions ara the primary means to document that all required tests for a particular test plateau are completed.
Maximum Extended Operating Domain The Maximum Extended Operating Domain (MEOD) is defined by the expansion of two regions of the power to flow operating map (See Figure 1.7-1, Sheet 1 of 3). The Extended Load Line Region (ELLR) boundary allows full power operation at less than rated flow conditions. ELLR allows operation with I a flow reduction to 75% core flow at 100% power and its corresponding power / flow constant rod line. The Increased Core Flow Region (ICFR) allows for testing with the maximum allowable core flow. This flow is based on a plant operational (such as recirculation pump motor amps, winding temperature, core plate delta pressure) or vibrational limit.
After the startup test program, the ICFR will be bounded by the maximum allowable core flow or 105% core flow, whichever is smaller. This expansion of the operating envalope permits improved power ascension capability to rated power.
l l MEOD boundaries are determined based on meeting thermal and reactivity i
margins, plant recirculation system capability, and acceptable flow induced vibration on vessel internal components. MEOD impact on accident analysis is discussed in Chapter 15 of the FSAR, Appendix 15E.
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o ee !eae e a eaelt iet lt n'selt iea!eeaeIa ea n lt iaelt iea o io ao ao 4o so so .7o so so too tio 120 PERCENT CORE FtOW P6wer-Flow Operating Map
Test Condition (TC) Region Definitions T oi,. Condition
('IC) Power Flow Map Region and Notes Open vessel (OV) Anytime the reactor vessel head is not fully tensioned or if tensioned, reactor coolant temperature is less than or equal to 200*F.
Heatup (HU) Anytime the reactor vessel head is fully tensioned and the plant is less than 5% thermal power and reactor coolant temperature is greater than 200'F.
1 Before or after main generator synchronization from 5%
to 32% thermal power and operating on recirculation pump low frequency power supply. Upper limit is the I 50% rod line.
2 After main generator synchronization from 50 to 75%
control rod lines, at or below the analytical lower limit of Master Flow Control mode and with the lower power corner within bypass valve capacity.
3 From 50 to 75% control rod lines above 80% core flow, and within maximum core flow.
4 On the natural circulation core flow line within 5%
of the intersection with the 100% power rod line.
l 5 From the 100% loadline to 5% below the 100% loadline and between minimum flow at rated recirculation pump speed (minimum valve position) to 5% above the analytical lower limit of the automatic flow control range.
Figure 1.7-1, Sheet 2 of 3 l
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Test Condition
('IC) Power Flow Map Region and Notes 6 Within 0 to -5% of rated 100% thermal power, above 95%
core flow and within maximum core flow.
7 From the bounding licensed loadline to 5% below the bounding licensed loadline with flow between minimum at rated recirculation pump speed (minimum FCV position) to the core flow which results in 90%
thermal power on the bounding licensed loadline.
8 Within 0 to -5% of rated 100% thermal power and within the bounding licensed loadline and 85% of rated core flow.
NOII: Core flow will exceed 100% rated core flow rate only for testing associated with the reactor internals prototype vibration measure-ments program and MEOD related testing.
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Figure 1.7-1, Sheet 3 of 3 l
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I 1.8 QUALITY ASSURANCE COVERAGE OF THE STARTUP TEST PROGRAM The Nuclear Quality Assurance Department has contributed to the Startup Test Program chiefly through surveillance coverage of testing activities I and additionally by providing QA review of Startup Test Instructions.
Beginning with Fuel Load, the Operational Support and Program Unit has provided on-shift coverage. Surveillances were conducted on approximately 80% of all tests performed. The primary intent was to verify compliance with the administrative program and individual test steps.
. The surveillance reporting program has been an effective means for communicating concerns and helping to maintain a high degree of awareness on the importance of program compliance. To complete the coverage, QA reviews all Test Results Packages for appropriate completion and program compliance prior to their approval by the Plant Operating Review Committee.
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b SECTION 2
SUMMARY
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-I 2.1
SUMMARY
OF TESTING 'IO DATE Since the receipt of the low power operating license the following significant activities have been successfully completed within the scope of the Startup Test Program:
I a. Fuel loading
- b. Initial Criticality
- c. Non-nuclear heatup for reactor internals vibration testing,
- d. All Test Condition - Open-Vessel and - Heatup testing.
- e. Entry into Test Condition One and the start of TC1 testing.
Testing on the following systems has been performed with very satisfactory results: Safety Relief Valves, Control Rod Drives, Nuclear Instrumentation, Main Steam Isolation Valves, NSSS and BOP piping, Plant chemistry, Radiation Monitoring, Recirculation System, the Emergency Response Information System, Water Cleanup, Feedwater level control and Loose Parts Monitoring.
Testing of the Reactor Core Isolation Cooling System has not been completed due to reactor vessel level anomalies observed during RCIC injection to the vessel.
Also, several other equipment and licensing problems have occured which have prevented the completion of the Startup Test Program in the initially scheduled time frame.
At the time of this report, the plant is continuing TC-1 testing.
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SUMMARY
OF KEY EVENTS I 3-18-86 Receive Low Power Operating License 3-21-86 Commence Fuel Loading 6-06-86 Initial Criticality at 0246 7-07-86 Complete Open vessel Testing 7-25-86 Commence Heatup Plateau Testing 10-03-86 Initial Nuclear Heatup achieved rated pressure and temperature 11-13-86 Receive Full Power Operating License from NRC 12-03-86 Complete Heatup Plateau Testing 12-15-86 Commence Test Condition 1 Plateau Testing 12-19-86 Initial Turbine Generator Synchronization I
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i 2.3 STARIUP TESTING PROGRAM CHRONOII)GY OPEN VESSEL PLATEAU TESTING PERIOD l
3-18-86 Receipt of the Low Power License from the NRC.
3-20-86 Commence Open Vessel Plateau testing. Several baseline
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information tests have been completed at this time.
3-21-86 Commence Fuel Loading operations. First assembly loaded at 0718, 3-31-86 SCRAM 86 Cause: IRM noise spike. RPS was in non-I coincident mode, i.e., the shorting links removed. Reactor not critical in Operating Mode 5; Performing Partial core S/D Margin Verification.
4-24-86 Complete Fuel Loading and start Control Rod Drive testing.
I 4-30-86 Load Californium Neutron Sources for Source Range Monitoring.
5-02-86 Enter Maintenance Outage for surveillance testing, system operability verification for Initial Criticality, Containment and Drywell Airlock Seals repair, and completion of a Fire Protection supply water modification.
6-06-86 Initial Criticality established at 0246.
6-06-86 Continue Control Rod Drive Testing.
6-10-86 Start RPV Internals and Head Installation.
6-20-86 Smoldering charcoal was discovered in the Off-Gas Charcoal Adsorbers during the performance of an acceptance test.
6-29-86 Reactor Coolant System Leakage Test completed.
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HEATUP PLATEAU TESTING PERIOD I 7-06-86 Off-Gas Charcoal Adsorbers re-ignited during the continuation of the test resulting in the changeout of charcoal from all I filter beds.
7-07-86 Complete all requirements for Open Vessel tests.
7-08-86 Enter outage for completion of work required for Non-Nuclear I Heatup. Repairs including MSIV leakage control and Drywell Airlock were completed and Division III Diesel Generator surveillances were performed.
, 7-25-86 Commence Heatup Plateau with Non-Nuclear Heatup testing for reactor internal vibration testing.
8-08-86 Complete Non-Nuclear Heatup testing and commence maintenance outage to complete Nuclear Heatup prerequisites.
I 8-31-86 Commence Nuclear Heatup.
9-02-86 SCRAM 86 Cause: Spurious upscale trip on the Intermediate Range Monitors.
9-04-86 U.S. Court of Appeals ruling barred the NRC's vote on the Perry Full Power License.
9-08-86 Testing held for 20 days to resolve RCIC equipment problers, and to rework throttle valve and steam flow instrumentation.
10-03-86 Initial Nuclear Heatup achieved rated temperature and pressure.
10-14-86 SCRAM 86 Cause: Personnel error while performing a surveillance test.
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'I HEA'IUP PLATEAU TESTING PERIOD - (CONT. )
I 10-14-86 U.S. Court of Appeals lifted the stay on issuance of the '
operating License.
10-23-86 Entered planned Pre-TC-1 outage.
11-13-86 Received Full Power License from the NRC.
11-13-86 U.S. Court of Appeals issued another stay against Perry operating above 5% power.
11-23-86 Pre-TC-1 Outage complete.
'I 11-26-86 SCRAM 86 Cause: excessive Feedwater flow while changing Feedwater pump alignment.
12-03-86 Complete Heatup Plateau Testing.
I TEST CONDITION 1 PLATEAU TESTING PERIOD 12-15-86 Commence TC-1 Plateau Testing (under 5% power due to license restriction).
12-19-86 Initial Generator Synchronization at 1503 resulting in a net output of 18 MWe while holding at less than 5% reactor power.
12-23-86 U.S. Court of Appeals vacated the stay on Full Power operation.
'I 12-30-86 Generator Synchronization was achieved at 0235 producing an average net 130 se and a high gross of 162 se at 17% power.
12-30-86 SCRAM 85 Cause: Manually initiated due to Technical i specification action in response to high conductivity in the feedwater. This resulted in tube repair and other modifications to the condenser.
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L TEST CCNDITION 1 PLATEAU TESTItG PERIOD - (CCNr.) >
l-08-87 Returned to Mode 2*and ccrnpleted the Control Tcan Ieak Test and other plateau testing before the RCIC steam inlet valve failed toci open resulting in susperriing testing and shutting down for valve maintenance. '
I l-11-87 Manual Shutdown for repairs to RCIC Steam Inlet Valve anh' Steam Bypass Valves.
l-21-87 Planc Startup ccxFnencM to test conditions until problems '
I were incurred on the F063 RCIC inlet valve operator and the plant was shutdown on January 24.
2-05-87 The plant ccmnenced startup and was shutdown tne same day due to a nonconfornance report (NR) on the (L31) Isak Detection k' ,
I system that needed resolution. A thormgh n4iew of alb open i NR's was completed by management subsequent to the shutdcyn
! resultirrj in minar eviditional work.
l L, ..
2-07-87 Upon the resolution of tne NR related work, criticality was achieved. During this time period startup testing includM ,
E feedwater low flow valve tests using thb notor driven feed- ;
- pump, and tuneups on the turbine driven fudpump mininum flow valves.
{ 2-13-87 SCRAM 87 Cause: An inadvertent redundant reactivity control system initiation. Criticality was achieved on February 14 at 1430.
2-14-37 When performing a PCIC injection to the vessel, a vessel level instrument transient occurred. While pursuing pcssible Technical Specification relief on the RCIC system, the plant continued to generate electricity while maintaining less than 12% thermal power.
25
I I
i SECTION 3 ;
I STARIUP TEST INSTRUCTIONS l
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26
3.1 STAR'IUP TEST INSTRUCTION FORMAT AND CONTENT Startup Test Instructions (STI's) are used for testing performed during the Startup Test Program. These instructions detail the step-by-step method used to perform and record the results of each test. STI's are also used to ensure and document that the tests are completed in the correct sequence. STI's also document that authorization is obtained prior to proceeding to the next test plateau or test condition, l
Startup Test Instructions are developed by the Startup Test Organization using Test Specifications provided by the Architect-Engineer (AE), Gilbert Associates Inc. (GAI) or the Nuclear Engineering Department (NED) or Nuclear Steam Supply System (NSSS) supplier, General Electric (GE); the guidelines of Regulatory Guide 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plants, as described in Section 1.8 of the FSAR; and as a minimum, all FSAR (including SER) startup test objectives.
STI's were prepared by the Startup Test Directors, Test Engineers, plant staff, GE, or other consultant personnel using project drawings, vendor manuals, design criteria, test specifications, and the FSAR.
Plant procedures and in.,, ructions will be used to the maximum extent practi' cal. The STI shall refer to the plant document when required. The l
reference shall be clear and specific.
The following sections comprise the body of an STI.
l 1.0 PURPOSE
2.0 DESCRIPTION
3.0 ACCEPTANCE CRITERIA
4.0 REFERENCES
5.0 PRECAUTIONS l
6.0 PREREQUISITES AND INITIAL CONDITIONS 7.0 TEST EQUIPMENT AND TEMPORARY INSTALLATION INSTRUCTIONS 27
l I
8.0 TEST INSTRUCTIONS AND ANALYSIS 9.0 SUPPORTING INFORMATION 10.0 ATrACHMENTS I Applicable test criteria shall be clearly identified in each test in-struction. If no Level 1 or Level 2 criteria are applicable, "None" is listed under the Level 1 or 2 titles as the criteria. The criteria levels, as listed below, are used. This assures that results different from expected or predicted receive proper attention while avoiding unnecessary cessation of testing. The levels are:
Level 1 If a Level 1 test criterion is not satisfied, the plant must be placed in a hold condition that is judged to be satisfactory and safe, based upon prior testing. Plant operating or test procedures, or the Technical Specifications, may guide the decision on the direction to be taken. Resolution of the problem must immediately be pursued by appropriate equipment adjustments or through engineering support. A Test Exception Report shall be prepared. Following resolution, the applicable test portion must be repeated to verify that the Level l requirement is satisfied. A description of the problem resolution must be included in the report documenting the successful test.
Level 2 If a Level 2 test criterion is not satisfied, plant operating or startup test plans would not necessarily be altered. The limits stated in this category are usually associarert with expectations of system transient performance, and whose chara.:teristics can be im-proved by equipment adjustments. An investigation of the related adjustments, as well as the measurement and analysis methods, would be initiated. A Test Exception Report shall be prepared. If equip-ment adjustments or corrective actions are specified in the TER resolution, then the applicable test shall be repeated to verify that the Level 2 requirement is satisfied.
Level 3 I If Level 3 test criterion is not satisfied, plant operating or startup test plans would not necessarily be altered. The numerical limits stated in this category are associated with expectations of plant subsystem, individual component or inner control loop transient performance. Level 3 performance is also not specified in fuel or vessel protective systems. If all Level 1 and 2 criteria are satisfied when a Level 3 criterion is not satisfied, then it is not required to repeat the transient test to satisfy Level 3 performance.
The occurrence must be documented and resolved in a Test Exception Report.
I iI l
29 l
?
I 3.2 STARWP TEST PROCEDURE ABSTRACTS I The abstracts on the following pages provide general information on the l content of each Startup Test Instruction. These abstracts in no way modify or replace those contained in Section 14 of the Final Safety Analysis Report. Refer to Section 14 for more information.
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I 30
I -
I STAR'IUP TEST ABSTRACTS I
Chemical and Radiochemical I STI-P35-001
'Ihis test describes how to collect chemical and radiochemical data for use in analyzing plant operations. During various plant conditions data is collected from the following systems: Reactor Water, Condensate, Feed-water, Control Rod Drive Water, Condensate Demineralizer, Reactor Water Cleanup, Main Steam and Liquid Radwaste. Using established methods and approved procedures, the analysis of the samples will be conducted and the results reviewed.
Chemistry data will be collected prior to fuel load, prior to heatup, during heatup, power ascension and during a "No RWCU" test. The radiation doses at selected locations on Recirculation Piping and RWCU piping will be measured after plant shutdown to identify any buildup of radioactive crud in the piping.
General Chemistry data will be collected at various test conditions to gain baseline data to be analyzed and used for future plant operation.
The purpose for testing the chemical and radiochemical parameters of the plant are first; to verify that chemical parameters of the reactor coolant and selected support systems meet acceptable limits and second; to deter-mine, using approved plant procedures, the adequacy of sampling equipment and analytical procedures / techniques for sampling. Additional objectives for this test are to evaluate fuel performance, confirmation of condenser I integrity, demonstrate proper steam separation-dryer operation, and to measure and calibrate certain process instrumentation.
I 31 l
I STARTUP TEST ABSTRACTS - (LDTI. )
STI-D21-002 Radiation Measurements I 'Ihis test will determine the background radiation levels in the plant environs prior to operation for baseline data on activity buildup and to monitor radiation at selected power levels to assure the protection of personnel during plant operation.
STI-J11-003 Fuel Loading The purpose of this test is to load fuel safely and efficiently to the full core size.
I STI-Jll-004 Full Core Shutdown Margin The purpose of this test is first to demonstrate that the reactor will be subcritical throughout the first fuel cycle with any single control rod fully withdrawn and second to determine quantitatively the shutdown margin of the as-loaded core.
STI-C11-005 Control Rod Drive System The purposes of the control rod drive system test are to demonstrate that the control rod drive (CRD) system operates properly over the full r,'.nge of reactor coolant temperatures and pressures from ambient to operating, and to determine the initial operatit.g characteristics of the entire CRD system.
32
STAR'IUP TEST ABSTRACTS - (CONT. )
I STI-C51-006 SRM Performance and Control Rod Sequence This test will demonstrate that the neutron sources, SRM instrumentation and rod withdrawal sequences provide adequate information to achieve l criticality and increase power in a safe and efficient manner.
l STI-C11-008 Control Rod Sequence Exchange This test performs a representative sequence exchange of control rod patterns at a significant power level.
STI-C51-010 Intermediate Range Monitor Performance This instruction adjusts the IRM system to obtain an optimum overlap with the SRM and APRM systems.
I STI-C51-011 Local Power Range Monitor Calibration l
l The purpose of this test is to calibrate the local power range monitoring system and to verify the LPRM flux response.
STI-C51-012 Average Power Range Monitor Calibration l This test calibrates the average power range monitor system.
l 33
STARTUP TEST ABSTRACTS - (00NT.)
STI-C91-013 Process Computer The purpose of this test is to verify the performance of the Process Computer and on-line NSSS computer programs under plant operating conditions.
STI-E51-014 Reactor Core Isolation Cooling System The purpose of this test is to verify the proper operation of the reactor I core isolation cooling (RCIC) system over its expected operating pressure range and to demonstrate reliability at power conditions and during RCIC startup.
STI-B21-016A Selected Process Temperatures This instruction will assure that the measured bottom head drain temper-I ature corresponds to bottom head coolant temperature during normal op-eration, to identify any reactor operating modes that cause temperature stratification, to determine the minimum position of the flow control valves which will prevent coolant temperature stratification in the reactor pressure vessel bottom head region, and to familiarize plant personnel with the temperature differential limitations of the reactor system.
STI-B21-016B Water Level Reference Leg Temperature The purpose of this test is to measure the reactor water level reference i leg temperature at rated temperature and pressure and steady-state, and ensure recalibration of the instruments if the measured temperature is different than the value assumed during the initial calibration.
I 34
I STARIUP TEST ABSTRACTS - (CONT. )
STI-B21-016B (CONT.)
Temperature data will also be taken during heat up testing to verify acceptance criteria for the shutdown range water level instrumentation.
The fuel range reference leg temperatures will not be verified since this range is calibrated at 0 psig with saturated water and steam conditions in I the reactor vessel and drywell.
STI-B21-017 System Expansion The purpose of this test is to confirm that safety-related pipe suspension systems and other systems as identified are working as designed and that I the pipe is free of obstructions that could constrain free pipe movement.
STI-C91-018 Core Power Distribution The purpose of this test is to determine the reproducibility of the TIP system readings.
STI-C91-019 Core Performance fois test will evaluate the core thermal power and core flow and evaluate the following core performance parameters:
- 1. Maximum linear heat generation rate (MLHGR).
'I 2. Minimum critical power ratio (MCPR).
- 3. Maximum average planar linear heat generation rate (MAPLHGR).
35
I l
i STARTUP TEST ABSTRACTS - (CONT.)
STI-B21-020 Steam Production To demonstrate that the reactor steam production rate is sufficient to satisfy all appropriate warranties as defined in the contract.
I STI-J11-021 Core Power-Void Mode Response The purpose of this test is to measure the stability of the core power-void dynamic response and to demonstrate that its behavior is within specified limits.
STI-C85-022 Pressure Regulator This instruction will demonstrate the following:
- 1. The optimum settings for the pressure control loop by analysis of the transients induced in the reactor pressure control system by means of the pressure regulator;
- 2. The takeover capability of the backup pressure regulator via simulated failure of the controlling regulator;
- 3. Smooth pressure control transition between the control valves and bypass valves when reactor steam generation exceeds steam used by the turbine;
'I 4. That other affected parameters are within acceptable limits during pressure regulator' induced transients.
I 36
L STARTUP TEST ABSTRACTS - (CONr.)
c STI-N27-023A Feedwater Control System The purpose of this test is to verify that the feedweter system has been adjusted to provide acceptable reactor water level control.
- STI-N27-023B Loss of Feedwater Heating This test will demonstrate adequate response to feedwater temperature loss.
I STI-N27-023C Feedwater Pump Trip The purpose of this test is to demonstrate the capability of the automatic core flow runback feature to prevent low water level scram following the trip of one feedwater pump.
I STI-N27-023D Maximum Feedwater Runout Capability The purpose of the test is to determine the maximum feedwater runout capability.
STI-N31-024 Turbine Valve Surveillance This test will demonstrate acceptable procedures and maximum power levels for recomended periodic surveillance testing of the main turbine control, stop and bypass valves without producing a reactor scram.
l 37
I STAR'IUP TEST ABSTRACTS - (CONT. )
STI-B21-025A Main Steam Isolation Valves Function Tests The purpose of this test are:
- 1. To functionally check the main steam line isolation valves (MSIVs) for proper operation at selected power levels,
- 2. To determine isolation valve closure times, and
- 3. To determine maximum power at which full closures of a single valve can be performed without a scram.
STI-B21-025B Full Reactor Isolation This test determines the reactor transient behavior that results from the simultaneous full closure of all MSIVs.
STI-B21-025C Main Steamline Flow Venturi Calibration This instruction calibrates the main steam flow venturis at selected power levels over the entire core flow range, the final calibration taking place with the data accumulated along the 60 percent and 100 percent rod lines.
STI-B21-026 Safety Relief Valves This test will verify the following:
- 1. That the primary system relief valves function properly (can be opened and closed manually).
I 38
(
l STAR'IUP TEST ABSTRACTS - (CONT. )
STI-B21-026 (CONT.)
- 2. That the discharge piping contains no major blockage.
- 3. Proper seating following operation.
STI-B21-027 Turbine Trip and Generator Load Rejection The purpose of this test is to demonstrate the response of the reactor and its control systems to protective trips in the turbine and generator.
I STI-C61-028 Shutdown From Outside the control Room To demonstrate that the reactor can be brought from a normal initial steady-state power level to the point where cooldown is initiated and under control with reactor vessel pressure and water level controlled from outside the control room. The reactor shall be brought from power operation to shutdown cooling mode from outside the control room.
I STI-B33-029A Recirculation Flow Control - Valve Position Loop The purpose of this test is to demonstrate that recirculation flow control system's capability while in the loop flow manual mode.
I STI-B33-029B Recirculation Flow Control I The purpose of this test are:
- 1. To demonstrate the core flow system's control capability over the entire flow control range, including both core flow neutron flux and load following modes of operation, and 39
I STARTUP TEST ABSTRACTS - (CONT.)
s STI-B33-029B (CONT.)
- 2. To determine that all electrical compensators and controllers are set for desired system performance and stability.
I STI-B33-030A One Recirculation Pump Trip and Restart The purposes of this test are:
- 1. To obtain recirculation system performance data during the pump trip, <
flow coastdown, and pump restart, and
- 2. To verify that the feedwater control system can satisfactorily control water level without a resulting turbine trip / scram.
STI-B33-030B RPT Trip of Two Pumps The purpose of this test is to record and verify acceptable performance of the recirculation two-pump trip circuitry and to demonstrate satisfactory recirculation loop flow coastdown.
STI-B33-030C Recirculation System Performance This test records recirculation system parameters during the power test program.
STI-B33-030D Recirculation Flow Control Valve Runback The purpose of this test is verify the adequacy of the recirculation runback to mitigate a scram upon the loss of one feedwater pump.
40
STARTUP TEST ABSTRACTS - (CONT.)
STI-B33-030E Recirculation System Cavitation The purpose of this test is to verify that no recirculation system cavitation will occur in the operable region of the power-flow map.
I STI-R43-031 Loss of Turbine-Generator and Offsite Power This test determines the reactor transient performance during the loss of the main generator and all offsite power, and to demonstrate acceptable performance of the plant electrical supply system. Loss of offsite power will be maintained for sufficient time to demonstrate that necessary equipment, controls, and indications are available following loss of offsite power to remove decay heat from the core using only emergency power supplies and distribution systems.
STI-B21-033 Drywell Picing Vibration The purpose of this test is to verify that the main steam, recirculation, and RCIC steam piping vibration is within acceptable limits and to verify during operating transient loads that pipe stresses are within code limits.
STI-F41-034 Reactor Internals Vibration To obtain vibration measurements on the reactor internal components to demonstrate the mechanical integrity of the system to flow-induced vibration.
I 41
STARTJP TEST ABSTRACTS - (CONT. )
STI-B33-035 Recirculation System Flow Calibration The purpose of this test is to perform complete calibration of the installed recirculation system flow instrumentation.
STI-G33-070 Reactor Water Cleanup System The purpose of this test is to demonstrate specific aspects of the mechanical ability of the reactor water cleanup system. (This test, performed at rated reactor pressure and temperature, is actually the completion of the preoperational testing that could not be done without nuclear heating).
STI-E12-071 Residual Heat Removal System
'Ihe purpose of this test is to demonstrate the ability of the residual heat removal (RHR) system to:
- 1. Remove heat from the reactor system so that the refueling and nuclear system servicing can be performed.
- 2. Condense steam while the reactor is isolated from the main condenser.
STI-N64-074 Off-Gas System This test will verify the proper operation of the off-gas system over its
! expected operating parameters and to determine the performance of the activated carbon adsorbers.
1 l
l 42 l
STARIUP TEST ABSTRACTS - (CONT.)
I STI-C95-099 ERIS (Emergency Response and Information System)
These tests verify that the BASIC ERIS and scram timing software and hardware have been correctly installed and calibrated. The test will also ensure that certain data needed from plant operation is incorporated into the plant specific data bases. Together these tests will verify that the system as a whole, is capable of meeting the design requirements for the BASIC ERIS and scram timing functions and is performing within the limits specified in the acceptance criteria. ,
I STI-M99-100 Integrated HVAC I To decunstrate the ability of ventilation systems to maintain specified Unit 1 and common area temperatures and relative humidity within specified limits during plant operation.
STI-P41-ll3 Service Water System The purpose of this test is to demonstrate that the service water system can provide a sufficient amount of cooling water to the heat loads it supplies.
STI-P42-ll4 Emergency Closed Cooling System To demonstrate that the emergency closed cooling system can provide sufficient heat removal capability for those components specified by GAI.
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43
I STAR'IUP TEST ABSTRACTS - (CONT. )
STI-P43-ll5 Nuclear Closed Cooling System To demonstrate that the Nuclear Closed Cooling System (NCCS) can provide sufficient heat removal to the heat loads it supplies.
I STI-P44-ll6 Turbine Building Closed Cooling System To demonstrate that the Turbine Building Closed Cooling System (TBCCS) can provide sufficient heat removal to the heat loads it supplies.
I STI-P45-ll7 Emergency Service Water System To demonstrate that the Emergency Service Water System can provide sufficient heat removal capability for these components specified by GAI.
I STI-N71-ll8 Circulating Water System To demonstrate that the Circulating Water System can provide sufficient heat removal to its heat loads.
STI-G42-119 Suppression Pool Cleanup System To ensure that the Suppression Pool Cleanup System (SPCU) can maintain water quality in the suppression Pool and at the SPCU demineralizer outlet within specified limits.
I 44
STARTUP TEST ABSTRACTS - (CONT.)
STI-N27-120 Feedwater System To demonstrate operation of the feedwater system and to demonstrate the automatic start capabilities of the motor driven feedwater pump.
STI-N36-121 Extraction Steam System To demonstrate that the Extraction Steam System can supply steam to its heat loads.
I STI-P99-122 BOP Piping Expansion and Vibration I The purpose of this test is to verify transient induced pipe vibrations and steady state vibrations are within acceptable limits, that piping /
piping supports can expand thermally without obstruction for selected BOP piping and that snubber and spring hanger expansion are within acceptable limits.
STI-T23-123 Concrete Temperature Survey The purpose of this test is to demonstrate the ability of natural heat transfer to cool the concrete surrounding selected pipe penetrations in I the secondary containment wall and drywell wall.
I 45
r STAR'IUP TEST ABSTRACTS - (CONT. ) l STI-Nil-124 Main and Reheat steam System The purpose of this test is to demonstrate that the Nuclear Steam Supply System and the Moisture Separator / Reheaters maintain a balanced steam flow to the high pressure and low pressure turbines during steady-state operations and during turbine valve testing; and to demonstrate that the appropriate valves function properly subsequent to a main turbine trip.
I STI-N21-125 Condensate System The purpose of this test is to demonstrate proper operation of the condensate system.
STI-N22-126 Main, Reheat, Extraction, and Miscellaneous Drains The purpose of this test is to demonstrate that the drains for the first MSIV before seat drain, second MSIV before seat drain, shutoff valve before seat drain, and main steam line drains are operating properly.
STI-N25-127 LP/HP Heaters, Drains and Vents The purpose of this test is to demonstrate that the low pressure and high I pressure heater drains and vents systems are capable of maintaining the water levels in their respective components within limits.
STI-N24-128 Condensate Demineralizer The purpose of this test is to verify that the condensate Demineralizer System can provide condensate in sufficient amount and sufficient quality.
46
I I STAR'IUP TEST ABSTRACTS - (CONT. )
STI-N33-129 Steam Seal The purpose of this test is to demonstrate operability of the steam seal evaporator when supplied with steam from the main steam system and extraction steam systen.
STI-N62-130 Condenser Air Removal System To demonstrate automatic actions of the steam jet air ejector (SJAE) air suction valves. To demonstrate that one SJAE can maintain sufficient vacuum in the condensers to operate the ttnit at approximately 100 percent power.
STI-N64-131 Off-Gas Vault Refrigeration System The purpose of this test is to demonstrate that the Off-Gas Vault I Refrigerations System can properly cool, and maintain cooled, the main off-gas process stream and the off-gas charcoal vaults.
I STI-P33-132 Turbine Plant Sampling To compare conductivity, dissolved oxygen, pH and sodium concentration readings with grab sample analysis.
I I 47
t j -
STARIUP TEST ABSTRACTS - (CONT.)
STI-R63-133 Loose Parts Monitoring System To obtain a full range of baseline data for the Loose Parts Monitoring System (LPMS).
I STI-M99-134 Equipment Area Cooling The purpose of this test is to verify that the RCIC, LPCS, HPCS, and RHR "A", "B" and "C" room coolers are capable of removing the postulated post-accident design heat loads.
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48
I STARTUP TEST RESULTS - (CONT. )
STI-P35-001 (Cottr. )
- 5. Off-Gas:
Results Criteria Isotopic Analysis Pretreatment ( Ci/sec) None Detected < 3.58E5 Off-Gas Vent ( Ci/cc) Alpha < 1.0E-11 None 4.2 STI-D21-002 Radiation Measurements (TSN-002)
Transfer Tube Survey:
The purpose of this test was to determine the adequacy of the shielding at accessible locations adjacent to the Inclined Fuel Transfer System (IFTS).
The test was performed prior to initial fuel loading utilizing the gamma
" pins" used for the startup neutron sources. These fourteen (14) pins each contained an average of 118 Curies of Antimony - 124 (Sb-124). The pins are approximately 19 inches long and were transferred together. The pins were arranged symmetrically around the circumference of a seven (7) inch diameter storage basket. It is estimated that the combined sources strength of these pins due to gamma radiation (in air) at the time of transfer was 1400 R/hr at one meter.
No shielding problems were observed during the test. The transfer tube mechanism was stopped at three points during the transfer of the sources from the Fuel Handling Building to the Reactor Building. These points were selected to correspond to points where the potential for "ctreaming" was highest. Detailed radiation surveys were performed at these points.
Additional surveys were performed with the sources near each end of the transfer tube.
56
I STARIUP TEST RESULTS - (CONT.)
I STI-P35-001 (CONT.)
The results of the heatup chemical analyses were as follows:
- 1. RNCU Influent:
Results Criteria Conductivity ( mho/cm 0 25' C) .452 5 2.0 Chloride (ppm) 0.0131 f 0.1 pH (0 25' C) 6.3 5.6 - 8.6 Isotopic Analysis ( Ci/ml) 2.30E-03 7.0 Boron (ppm) < 1.0 $ 5.0 Silica (ppm) 0.105 f 5.0
- 2. Feedwater:
Conductivity ( mho/cm 0 25' C) 0.0630 f 0.2 pH (0 25' C) 7.3 5.7 - 7.5 Metallic Impurities (ppm) 0.0280 f 0.10 Cu Impurity (ppm) 0.000267 1 0.002 Oxygen (ppm) 0.02 0.02 - 0.2
- 3. CRD Water:
Conductivity (umho/cm 0 25' C) 0.0760 f 0.1 0xygen (ppm) 0.02 f 0.05 I 4. Condensate:
Conductivity ( mho/cm 0 25' C) 0.0714 N/A Chloride (ppa) < 0.0010 N/A suspended Iron oxide (ppm) 500 N/A 55
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I STARTUP TEST RESULTS - (CONT.)
STI-P35-001 (CONr.)
The results of the Pre-heatup chemical analyses were as follows:
I 1. RWCU Influent:
Results Criteria Conductivity (vmho/cm 0 25' C) 0.367 5 10 Chloride (ppm) 0.0011 f 0.5 pH (0 25' C) 6.6 5.3 - 8.6 Boron (ppm) <1 f 5.0 Silica (ppm) < 0.01 f 5.0
- 2. CRD Water:
s.
Conductivity ( mho/cm 0 25' C) 0.0776 N/A oxygen (ppm) 0.063 N/A
- 3. Condensate:
Conductivity (vmho/cm 0 25' C) 0.131 N/A Chloride (ppm) 0.002 N/A suspended Iron oxide (ppm) 0.025 N/A 1
l l
s 54
STARIUP TEST RESULTS - (CONT.)
STI-P35-001 (CONT.)
- 3. Water quality must be known at all times and must remain within the guidelines of the Water Quality Specifications.
Test Results:
The results of the Pre-fuel load chemical analyses were as follows:
- 1. RNCU Influent:
Results Criteria
- a. Conductivity ( mho/cm @ 25' C) 0.80 $10.0
- b. Chloride (ppm) < 0.01 f 0.5
- c. pH (@ 25' C) 6.0 5.3 - 8.6
- d. Turbidity (NIU) 0.07 N/A
< 1.0 I e.
f.
Boron (ppm)
Silica (ppm) 0.011
$ 5.0 f 5.0
- g. Isotopic Analysis ( Ci/ml) None N/A Detected
- 2. Fuel Fool Cooling and Cleaning Influent:
- a. Conductivity (pmho/cm @ 25' C) 0.91 f 3.0 I b.
c.
Chloride (ppm) pH (@ 25' C)
< 0.01 6.2 5 0.5 5.3 - 7.5
- d. Metallic impurities 1.26E-3 f 0.1 (Fe, Cu, Ni, Cr) (ppm)
- e. Suspended solids (ppm) < 0.5 5 1.0 I
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I STAR'IUP TEST RESULTS - (CONT. )
STI-P35-001 (CONT.)
Grab samples were obtained in accordance with approved plant procedures from the Reactor Water Cleanup (RWCU) influent, Control Rod Drive (CRD) water, and from the discharge of the hotwell pumps with the RNCU, CRD, and condensate demineralizers in operation. The samples were analyzed using various plant chemistry procedures. All of the results were within the limits specified by the Acceptance Criteria and no Test Exceptions were generated.
During Test Condition Heatup, with reactor operating at 3% power, grab samples were obtained from the Reactor Water cleanup influent, Feedwater, CRD water, Condensate, and Off-Gas systems. For Feedwater, a 24-hour composite sample was obtained and analyzed for total metallic and copper impurities. In addition, radiation monitors for the Plant Vent Gas (Units 1 & 2), Heater Bay / turbine Building (HB/TB) Gas, Off-Gas gas and Radwaste to Emergency Service Water (ESW) were verified to not be in their alarmed states with the setpoints set at 10% (Alert) and 70% (High) of the I respective Technical Specification limits. All test samples / data were obtained with the reactor at rated pressure and the RNCU and Condensate Demineralizer systems in operation. The chemical samples were obtained and analyzed by the plant chemistry department as part of their normal daily rounds, to the extent practical, and the results evaluated.
Acceptance Criteria Level 1 I 1. Chemical factors defined in the Technical Specifications and Fuel Warranty must be maintained within the limits specified.
- 2. The activity of gaseous and liquid effluents must conform to license limitations. (Not applicable to Pre-fuel and Pre-heatup since no radioactive effluents existed.)
I 52 l
l STARIUP TEST RESULTS I
4.1 STI-P35-001 Chemical and Radiochemical (TSN-003, 023, 058)
Test Description Prior to fuel load, chemistry analyses were performed on Reactor Water I Cleanup Influent and Fuel Pool Cooling and Cleanup Influent. At this time the drywell head was removed, the reactor cavity was flooded and the Reactor Building and Fuel Building fuel pools were filled. The neutron sources were installed in the reactor vessel. 180 new fuel bundles were stored in the reactor building fuel pools and 568 fuel bundles in the fuel handling pools.
With the RWCU and Fuel Pool Cooling and Cleanup Systems in operation, grab samples were obtained from the Fuel Pool F/D inlet and the RWCU influent in accordance with approved plant procedures SOI-P35 (Reactor Plant sampling) and SOI-P34 (Nuclear Sampling System) respectively. In addi-tion, a 24-hour integrated flow sample for metallic impurities was obtained using CHI-42, Miscellaneous Sampling Systems. The samples were analyzed using plant chemistry procedures. All of the results were within the limits specified by the Acceptance criteria and no Test Exceptions were generated.
I Prior to Heatup, chemistry analyses were performed on Reactor Water Cleanup Influent, CRD Water, and Condensate Hotwell Pump Discharge. At this time the reactor head was installed, fuel loading was complete and Reactor Coolant System Leakage Pressure Test was in progress.
I I
I 51
tore: The following Startup Test Instruction Results section reflects those STI's that have been completed and approved by Plant Management. The remaining tests will be included in subsequent reports.
I
- I 50
I l
f I
SECTION 4 i
I s_ zes, mers I
lI 49
l STAR'IUP TEST RESULTS - (CONr. )
I STI-D21-002 (CONT.)
All general areas surveyed measured less than or equal to 0.2 milliR/hr throughout the test. It was necessary to insert a Teletector probe five to ten feet inside the small (approximately 2 inch) gap between the Fuel Handling and Reactor buildings and practically adjacent to the transfer tube in order to measure radiation levels in the 100 to 300 milliR/hr range. A contact reading on the tube shielding in the rattle space of 31Vhr was observed with a fully inserted Teletector.
The test Acceptance Criteria was satisfied and no Test Exceptions were generated.
Acceptance Criteria Level 1 The radiation doses of plant origin and the occupancy times of personnel in radiation zones shall be controlled consistent with the guidelines of the standards for protection against radiation as outlined in 10CFR20,
" Standards for Protection Against Radiation".
Test Results:
The following table summarizes the data from the HP surveys during the I transfer of the Startup sources:
I 57
I I STAR'IUP TEST RESULTS - (COffr. )
STI-D21-002 (CONT.)
Transfer Tube Survey Data I Gamma - mR/hr Carrier Position Note 1 Point No.
Note 2 A B C D E F G l l '
l'%l%l%l%l%l -
l l
I l"" l l
l%l%l%l ""
l l
, j eeeml -l -lxl<glx l -l-S j l "" l%l7l' y,l
, l*~ l l l l*" l l
"" l % l'"K ? % ,l ~
l l
! ! l<0 6 l7 !
X63101 N/S N/S N/S N/S l < <0 < I
- X642 N/S N/S N/S N/S l .1 21 2 2 1 N/S = No Survey values are general highest ( in mr/hr unless )
area reading (specified otherwise)
I 58 1
STARTUP TEST RESULTS - (CONT.)
STI-D21-002 (CONT.)
liighest Readings were at contact with shielding in inaccessible areas using Teletector.
NCTIES
- 1. Carrier Position - indicator on Reactor building panel 1F42-P001 and Fuel Building panel 1F42-P004.
A. 1034, Lower Pool B. 745, Fill / Drain Position C. *541, Fuel fiandling Building, Mid-support Room D. *349, Shield Building Annulus Room E. *347, Containment Building Valve Room F. 269, Flap valve Closed / Fill Valve Closed G. 11, Upper Pool
- Source Pins at position of potential streaming
- 2. Survey Area Locations.
F62001 - Fuel IIandling Building West, Elev. 620' I59912 - FPPC Surge Tank Area, Elev. 599' 159914 - Access Area Under Unit 1 IFTS, Elav. 599'
,g - Containment Equip. liatch area. E11v. 654', near IFTS g X654 X620 - Containment Equip. Ilatch area, Elev. 620', near IFTS FX63101 - IFTS Valve Room, Elev. 631' X642 - Access Area / Cable Chase area inside Containment, Elev. 631'/642' X664 - RNCU F/D llolding Pump Access Area, Elev. 664' l
- 3. See Figure 1 for IFTS carriage locations and elevations.
I 59
I I STARIUP TEST RESULTS - (CONr.)
STI-D21-002 (CONT.) g-. , , _
U PPer"P[1
'?
. S f, '
! .. k l
I l .
l g
Ij l
- Upender
~
I .-
=i
- 1. miJ-Elev. 646' I
e 8
~
l Shield Bldg. ,- ,
e Cncmc Bldg.
Annulus Room e
- s Valve Room
~
f- i Elev. 631' I f /
.. i e'
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i, _ . . sun.omol.,
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.<// ,! ~, ;,,
. Upender I /
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Figure 1
/ Inclined Fuel Transfer System 60 I
STAR'IUP TEST RESULTS - (CONT )
STI-D21-002 (CONT.)
General Surveys: (TSN's -001, 021, 057)
A total of 203 radiation base points were established and " radiation survey tags" were installed in the plant to identify these points. Prior I to fuel load, general area radiation measurements and radiation measurements at the base points were taken. All measurements were beta-gamma only.
Subsequent to initial criticality, a set of radiation measurements were made throughout the plant. In addition to the required beta-gamma readings, Health Physics obtained neutron readings at various radiation survey tags on the outer wall of the Containment. The 203 radiation survey points were monitored, and general area radiation measurements were taken during this test.
After achieving operating reactor pressure during nuclear heatup, a set of radiation measurements were made throughout the plant. Ilealth Physics obtained gamma readings at all locations and neutron readings at various survey points on the outer walls of the Drywell and Containment. The 203 posted radiation survey points were monitored during this test.
All test acceptance criteria was satisfied and no Test Exceptions were generated.
t 1
61
STARTUP TEST RESULTS - (CONT.)
STI-D21-002 (CONT.)
Acceptance Criteria Level 1 The radiation doses of plant origin and the occupancy times of personnel in radiation zones shall be controlled, consistent with the guidelines of the standards for protection against radiation as outlined in 10CFR20,
" Standards for Protection Against Radiation".
Results:
- a. Radiation Measurement Survey Prior to Fuel Loading. (TSN 001)
The results of this survey show that the criteria is satisfied. All areas of the plant were less than 0.2 mr/hr except for the !!ealth Physics' instrument calibration room in the Auxiliary Building. Due to the calibration sources in this room, the general survey results were 0.4 to 0.8 mr/hr.
- b. Post Fuel Loading Radiation Measurements Survey following the first Reactor Critical. (TSN 021)
The results of this survey show that the criteria is satisfied.
There was no significant change from the data taken prior to fuel loading.
- c. Reactor Power at 3% and Rated Reactor Pressure during fleatup Test Plateau. (TSN 057) 62
STARTUP TEST RESULTS - (CONT.)
I STI-D21-002 (Cotfr. )
All areas surveyed were inside the posted Radiological Control Area (RCA)
(required for areas > 0.5 meewbr). All areas were measured at less than or equal to 0.2 mrem /hr gamma and less than or equal to 0.5 meewhr neutron with the exception of the Reactor Water Cleanup Rooms (RNCU). The gamma readings were 1.2 mreW hr (Room A) and 0.8 mrem /hr (Room B).
Neither room was greater than the 5 mrewhr requirement for posting as a Radiation Area.
The results of the radiation surveys were compared with the maximum radiation fields described in FSAR Figures 12.3-1 through 12.3-8. In all cases the measured values were well below the FSAR maximum values, as expected for this low power level.
4.3 STI-Jll-003 Fuel Leading (TSN -007, 011, 012)
Section 8.1 Fuel Loading The initial core of Perry Unit 1 was successfully loaded with 748 fuel assemblies in 14 days (March 21, 1986 to April 24, 1986). Adequate shutdown margin was demonstrated after 144 bundles were loaded. Control rod functional tests and friction tests were performed to show that all fuel assemblies were properly loaded, oriented, and seated in the core.
I The Level 1 Acceptance Critoria was satisfied.
l Acceptance Criteria Level 1 The partially loaded core must be suberitical by at least 0.38% delta K/K with the analytically determined highest worth control rod fully withdrawn.
63
STARTUP TEST RESULTS - (COtfr.)
STI-Jll-003 (CONT.)
Section 8.1 After 144 fuel assemblies were loaded, nine control rods (the reactivity equivalent of the highest worth rod) were withdrawn one rotch at a time while observing the nuclear instrumentation. During the first attempt, a reactor scram occurred due to a spurious IRM liigh Flux signal on two IRM Channels (the reactor protection system was in the non-coincident mode due to the shorting links being removed). At the time, four control rods were fully withdrawn and a fifth rod was partially withdrawn. All rods scram-med successfully. After an investigation and review by plant management, the shutdown margin demonstration was successfully reperformed later the same day. The nuclear instrumentation did not indicate a continuous positive period thus demonstrating subcriticality.
Prior to the start of fuel loading, four fuel loading chambers (FLC) were assembled, placed in the core, and connected to the permanent SRM pre-amplifiers. The scram setpoint was set at 20,000 cps and the rod block I was set at 10,000 cps. The reactor protection system was placed in the non-coincidence scram mode (shorting links removed)'. liigh voltage and discriminator curves were obtained for each FLC and signal to noise ratios for each rLC were verified periodically throughout fuel loading. In general, the FLC's performed very well during the entire fuel load.
Prior to the start of fuel loading, 180 fuel bundles were transferred to the upper pool storage racks from the ruel llandling Building. Most of the delays during fuel loading were due to equipment problems with the Refuel Bridge (FIS). The Refuel Bridge main power cable failed on two occasions causing a total delay of about 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br />. Also, delays were caused by failure of the refuel bridge grapple air hose (13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />) and ruel llandling I Bridge Interlock malfunctions (22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />).
i e4 l
g
STAR WP TEST RESULTS - (CONT.)
t STI-Jll-003 (CONT.)
Section 8.1 Installation of the 7 SbBe Neutron Sources was a prerequisite to this procedure and was controlled by another instruction. Due to low neutron count rates, four additional sources (Cf-252) were installed after the core was fully loaded. The SbBe sources remained in the core.
A plot of inverse count rate (lA) was taken during fuel load to verify suberiticality through the entire fuel load. The plot was taken after loading each fuel assembly until 16 assemblies were loaded. Subsequently, 1/M plots were taken every 4 assemblies until 256 fuel assemblies were loaded. After 256 assemblies were loaded 1/M plots were taken every 16 assemblies. Plotting frequencies were increased if the current 1/M plot indicated that criticality would occur prior to the next scheduled 1/M plot.
All fuel moves were done per the fuel movement checklist. The fuel movement checklist coordinated all major activities including moving fuel from the lower pool to the upper pool and into the vessel, various Technical Specification activities (coupling checks, verification of an operable SRM in the required quadrants, etc.), subcritical checks, and visual verification of dry tube engagement.
On two occasions fuel bundles had to be inspected for possible damage. On April 20, bundle LY0042 was bumped against the lower fuel pool wall while being moved from its storage location to the transfer tube upender. The bundle was inspected and no damage found. On April 23, bundle LY0059 became temporarily stuck nt approximately 40 inches into the core during loading. Continued attempts to lower the assembly were not successful.
This assembly and the surrounding assemblics were inspected for damage and none was found. The assemblies were reloaded successfully, ruel Loading was completed on April 24, 1986.
65
STARRIP TEST RESULTS - (CONT. )
STI-J11-003 (CONT.)
Section 8.2 - Verification of Core Loading Core verification was performed in three parts. The first was to check for proper fuel assembly seating by lowering the fuel grapple to just above the assembly bail handles and slowly traversing the core, ensuring that the fuel grapple did not come into contact with any bail handles.
The second phase of core verification was a check to ensure that each fuel assembly was in the correct location. This was performed by video taping each fuel assembly serial number while traversing the core in a prescribed order and comparing the serial numbers to a verified core map. The video tape was later independently verified. The third and last phase was to check the fuel assemblies for proper orientation. This was performed by video taping the core a cell at a time and checking the orientation using five prescribed methods of orientation verification. This video tape was also independently verified.
I section 0.3 - Installation of Cf-252 Neutron Sources The purpose of this test was to install four Californium neutron sources into the reactor vessel. This was done after the completion of initial fuel loading. The four neutron sources, consisting of one Californium and two antimony pins in a beryllium sloove, are identical to the previously loaded antimony-beryllium sources except for the Californium pins.
Sixteen fuel bundles (four around each new source location) were unloaded from the core and temporarily stored in the upper pool storage racks. The sources were then loaded into the core. Next, the sixteen fuel bundles were reloaded into the reactor.
I 66
l I I STARRIP TEST RESULTS - (COTE.) l STI-Jll-003 (Corn. )
Essentially the same procedural controls used during initial fuel loading were used to perform this test.
The increase in S M count rates observed following the loading of the new sources were:
BEFORE AFTER SM Channel At 7 cps 30 cps SM Channel B: 2.5 cps 10 cps SM Channel C: 5 cps 25 cps SM Channel D: 3 cps 25 cps 4.4 STI-Jll-004 rull Core Shutdown riargin (TSN-019)
I The purpose of the rull Core Shutdown tiargin Test was to demonstrate that the reactor has adequate shutdown margin at any time during the first fuel cycle. Adequate SDt1 was demonstrated theoretically by withdrawing Groups 1 t. 2 of Sequence "B" and verifying the reactor remained subcritic31. The shutdown margin was then calculated to be 2.70% delta k/k by pulling control rods to criticality, obtaining the appropriate data, and perform-ing the shutdown margin calculation. This shutdown margin satisfies the Level I criterion of being at least 0.38% delta k/k. Predicted criticality was also compared to the actual point of criticality. The difference was -0.097% delta k/k, which satisfies the Level 2 criterion of t 1.0% delta k/k.
67
I STARTUP TEST RESULTS - (CONT.)
STI-J11-004 (CONT.)
Acceptance Criteria Level 1 The shutdown margin of the fully loaded, cold (60*F) xenon free core occurring at the nost reactive time during the cycle must be at least 0.30% delta k/k with the analytically determined strongest rod (or its reactivity equivalent) withdrawn. If the shutdown margin is measured at some time during the cycle other than the most reactive time, compliance with the above criterion is shown by demonstrating that the shutdwn margin is 0.30% delta kA plus an exposure dependent increment which adjusts the shutdown margin at that time to the minimum shutdown margin.
Level 2 Criticality should occur within 1.0% delta k/k of the predicted critical.
Results
'Ihe shutdown margin was calculated to be 2.70% delta k/k which satisfies the Level I criterion of 0.30% delta k/k. Criticality occurred within
-0.097% delta k/k of the predicted value. This value satisfies the criterion of 1.01 delta k/k. All test critoria were therefore satisfied.
Discussion I rull coro shutdown margin demonstration began by withdrawing Groups 1 & 2 of Sequence "B" to the fully withdrawn position (40). Theoretically, if the core remained suberitical with Groups 1 & 2 fully withdenwn, adequate SDM was demonstrated throughout cycle 1. This also takes into account a higher moderator temperature of up to 150'r. The core romalnod shutdown in this configuration with a modorator temperature of approximately 06'r.
60
STARTUP TEST RESULTS - (COfTF. )
STI-J11-004 (CONT.)
To determine an analytical value for SDM, rod withdrawal continued until the reactor became critical. Criticality occurred on the 9th rod in Group 3 at Notch 12. The reactor period was stable at approximately 100 sec.
Data was collected for approximately seven (7) minutes af ter criticality to obtain an accurate number for the reactor period. Then the teactor was made subcritical by inserting the rod to Notch 8. Using the data collec-ted and core reactivity data provided by GE San Jose Engineering, the core shutdown margin was calculated.
I 4.5 STI-C11-005 Control Rod Drive System (TSN -005, 006, 000)
The control rod drive system was friction tested during open vessel and rated pressure conditions to verify no significant binding existed with the fuel or within the control rod drive mechanisms. Following the friction testing, all rods were individually scrammed to verify rod insertion times were sufficiently fast. Additional data was taken for insertion withdrawal times, ganged rod timing, and system hydraulic performance.
Acceptance Criteria Level 1
- 1. Each CRD must have a normal withdraw speed less than or equal to 3.6 inches per second, indicated by a full 12-foot stroke in greater than or equal to 40 seconds.
- 2. For vessel pressures between 950 psig and 1050 poig the maximum Jcram times of individually fully withdrawn CRDs shall comply with the
( following tables (Note: Performance rated with charging headarn at 1750 poig). ,
t 69
i I
STARWP TEST RESULTS - (COtif.)
STI-Cll-005 (Cot 1T.)
The scram insertion time of each control rod from the fully withdrawn position, based on opening of contacts of main scram contactor (deenergization of scram pilot valve solenoids) as time zero, shall not exceed the following criterion:
Maximum Insertion Times (sec)
From opening of contacts of main scram contactor (de-energization of Reactor Pressure scram pilot valve solenoids) to psig Notch Position
- 43 29 13 950 0.31 0.01 1.44 1050 0.32 0.06 1.57 I *For intermediato vessel domo pressure, the scram timo criteria are determined by lineas interpolation at each notch position.
- 3. If the maximum scram insertion time of one or more control rods exceed Criterion 2 (above), then the following criteria are applicables
- a. The individual scram timos of a drive exceeding the times of critorion 2 (above shall not exceed the fo:. lowing table:
I Maximum Incertion Times (sec)
From opening of contacts of main scram contactor (do-onorgization of scram pilot valve solonoids) to I Reactor Pressure psig Noten Ponition 43 29 13 950 0.38 1.09 2.09 1050 0.39 1.14 2.22 70
I I STARWP TEST RESULTS - (C0tir.)
STI-C11-005 (CONT.)
- b. The total number of drives failing Criteria 2 but meeting criterion 3.a shall not exceed 7.
- c. The average scram times of the remaining (i.e., those that meet criterion 2) individual control rod drives shall be less than the following table:
Maximum Insertion Times (sec)
From opening of contacts of main scram contactor (de-energization of Reactor Pressure scram pilot valve solenoids) to psig Notch Position 43 29 13 950 0.30 0.70 1.40 1050 0.31 0.84 1.53
- d. A drive failing Criterion 2 but meeting the criteria under 3 shall not occupy an adjacent location in any direction, including the diagonal, with another slow or inoperativo drive.
Note that a drive that fails Criterion 3 is considered to be inoperative.
Level 2
- 1. Each CBD must have a normal insert or withdraw speed of 3.010.6 inches per second, indicated by a full 12-foot stroke in 40 to 60 seconds.
71
I STAR *IUP TEST RESULTS - (CONT.)
STI-Cll-005 (CONT.)
- 2. With respect to the control rod drive friction tests, if the differential pressure variation exceeds 15 psid for a continuous drive in, a settling test must be performed, in which case, the differential settling pressure should not be less than 30 psid nor should it vary by more than 10 paid over a full stroke, tore The differential settling pressure should be nominally 5 paid higher at the 00 position than at any other position along the CR due to the proper functioning of the spring actuated buffer piston located at the top of the drive.
- 3. The CRD's total cooling water flow shall be between 0.28 and 0.34 gpm times the total number of drives.
- 4. For vessel pressures below 950 psig tne maximum scram time of individual fully withdrawn CRDs shall comply with the criteria given in rigure 1. This is the time from the opening of the main scram contactor (de-energization of scram pilot valve solenoids) to notch 13.
- 5. Duffer time (defined as pickup of position indicator probe switch "52" to drop out of "52") shall not be less than 10 milliseconds when scram testing at nominal accumulator conditions with the reactor open to the atmosphere and 15 milliseconds at nominal accumulator conditions with the reactor at rated pressure.
I 6. In the continuous ganged rod mode, the rods shall always move j together so that all rods are within two notches of all other rods in the gang.
72
STARIUP TEST RESULTS - (CONT.)
STI-C11-005 (CONT.)
Level 3
- 1. Upon receipt of a simulated or actual scram signal (maximum error),
the FCV must close to its minimum position within 10 to 30 seconds.
I
- 2. The CRD system flow should not change by more than 3.0 gpm as reactor pressure varies from 0 to rated pressure.
- 3. The decay ratio of any oscillatory controlled variable must be less than or equal to 0.25 for any flow setpoint changes or for system disturbances caused by the CRD's being stroked.
I i
(Irm2rrIcemLY mum) l
!l 73
o STARWP TEST REStETS ABSTRF, - (CCNr. )
STI-C11-005 (carr.)
1.s
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l 32
- soo soo ooo 1000 a too PRESSURE VESSEL DOME (psig) rigure 1 Range of 75% Scram Incertion Times - tteasured to the Pick-up of Reed Switch at Notch Position 13. .
74
I I STARTUP TEST RESULTS - (COtfr.)
STI-Cll-005 (CONT.)
Results:
All control rods were friction tested during open vessel testing and half during rated pressure testing with no significant binding being indicated.
I All control rods were scram timed during open vessel testing and rated pressure testing. All times were found to be acceptable, although several control rods required minor adjustments and maintenance. All additional data for stroke times and system hydraulic performance was acceptable.
Discussion Cll-005-8.1: Insert /Nithdraw Timing The purpose of this test was to ensure satisfactory control rod insert and withdrawal times. All 177 control rods were verified to meet the criteria of 40 to 60 seconds for a full insert or withdrawal.
Cll-005-8.2: reiction Testing Control rods are friction tested to determine if any significant binding exists between the fuel bundles and the control rod or within the CRD mechanism. If the variation in dP for the friction test exceeds 15 psid, then a settle test is performed. This involves inserting a CED full stroke, one notch at a time, and analyzing the settle dP's. If the settle dP is less than or equal to 30 psid or the variation exceeds 10 paid, with a nominal extra 5 psid allowed at notch 00, then the rod fails. The rod in then exercised by stroking fully five times and then retested.
75
STARRIP TEST RESULTS - (CONT.)
STI-cll-005 (cont.)
Following fuel loading, all control rods are friction tested during open vessel. For all control rods, drive differential pressure ranged between 32 and 95 psid and the variation ranged between 4 and 16 psid. Wo rods required a settle test be performed and these met the settle test criteria. All 177 control rods met the acceptance criteria for friction testing for the open vessel condition.
At rated pressure of 920 psig, half of the CRD's (80) were friction tested, one rod, 22-59, exhibited a higher dP than the others at notch 00 and was documented on TER 059-1. This was evaluated by GE and found to not exhibit excessive friction. Following the maintenance outage, a c settle test of rod 10-15 was satisfactorily performed. 21s was necessary when a review of the test data found that the rod had missed the screening criteria for the full insert friction test by 1 psid, thereby requiring a settle test.
Cll-005-0.3: Single Rod Scram Timing once the friction testing was completed, individual scram testing of all the control rods was performed. This was performed both for open vessel testing and rated pressure testing.
2e scram times for this test were based on the time from the de-energization of the scram pilot valve solenoids to the pickup of the position indicator probe (P!P) "36", L.o., rod position 13, and additionally for the rated pressure testing, to rod positions 29 and 43.
The buffer time is based on the pickup of PIP "52" to the drop out of "52". This data was obtained by connecting a Dranetz 2000 TJme Events Analyzer to the RC&IS test panel (Illl3-P610). The Time Events Analyzer collected the times for scram initiation, all odd rod positions, and the I
76
v I STARWP TEST RESULTS - (CONT. )
STI-Cll-005 (CCNr. )
buffer position. After being fully withdrawn, the control rod is scrammed by placing the local test switches on the HCU in the test position.
For the open vessel testing, the scram times varied between 0.688 seconds and 0.882 seconds. The mean scram time was 0.771 seconds with a standard deviation of 0.0226 seconds. The buffer times were between 0.008 and 0.026 seconds. 'Ihirty-four (34) rods had buffer times of less than 0.011 with twenty-three (23) being 0.010 seconds. As a result, the design specification and the startup test specification was changed from 11 to 10 milliseconds acceptance criteria for buffer time. The eleven rods with buffer times of less than 10 milliseconds were later vented and satisfac-torily retested for buffer time.
The open vessel test was concluded by determining which rods were outside the 2 sigma confidence bands for scram times. Nine rods were outside this band. All nine rods were retested three times each, except for rod 54-27, which had a faulty test switch. Later, fc11owing repair of the switch, this rod was also retested the required three times. In sumary, all 177 rods were satisfactorily tested for scram times for open vessel conditions.
The rated pressure test required reactor pressure to be raised to greater than 950 psig to match the pressure required for scram timing to satisfy Technical Specifications. This test required the plant to fully insert all rods once the segaence A rods were tested, and than return to test conditions on sequence D. Five rods were found to exceed the scram timing criteria at position 43. One, 30-51, was declared inoperable and the other four were classed as " slow" rods per Technical Specifications.
I Since all five satisfied their acceptance criteria to position 29 and 13, the slow times to position 43 was likely a result of either the scram pilot solenoid valve not bleeding off air fast enough, or the scram valves net opening fast enough.
77
STARItJP TEST RESULTS - (CONT. )
I STI-Cll-005 (Cotfr. )
For the rated pressure testing, the scram times varied between 0.903 seconds and 1.304 seconds. The mean scram time was 1.042 seconds with a standard deviation of 0.0395 seconds. The buffer times were between 0.015 and 0.113 seconds. Ivelve rods, five faster and seven slower, were outside the two sigma confidence bands for times to position 13. Since only sequence a rods could be tested, only five rods, one of the faster and four of the slower rods, could be retested the required three times.
During a subsequent maintenance outage, the proper operation of the scram valves for the slow rods was verified. Minor adjustments were made if required. The rcram pilot solenoid valves for the slow-inop rod (38-51) were replaced. Also, a total of 24 control rod drive mechanisms were replaced for unrelated reasons. These CRD's had experienced elevated temperatures cause' primarily by too little cooling flow when the CRD was full-out. The feature of the rod designed to allow adequate cooling flow when the rod was backseated was not working properly so the affected CRD's mechanisms were replaced. One of the rods whose mechanism was replaced was one of the five slow rods. All 24 CRD's plus the four remaining slow CRD's were rotested for scram times during an operational hydrostatic test at greater than 950 psig following the maintenance outage. All rods ratested satisfactorily except one (02-39, the slowest rod) of the slow rods, which could not be withdrawn due to a faulty directional control valve (DCV). The ' slow-inop' rod, 38-51, was retested three times satisfactorily.
l!
I 70
i l
I 1 i STARIUP TEST RESULTS - (CONT.)
l l
STI-Cll-005 (CONT.)
Cll-005-8.4: Ganged Rod Timing This test was performed during open vessel testing immediately after initial criticality. The purpose of this test was to demonstrate that the control rods will move properly in gang mode, i.e., stroke within two i notches of each other during gang insertions and withdrawals. All 52 gangs were tested satisfactorily.
Cll-005-8.5: Scram Timing of Selected Rods Prior to reaching rated conditions, five rods were individually scrammed,
- I three times each, at both 600 and 800 psig. These were selected based on the scram times obtained during open vessel testing. All five rods passed their acceptance criteria at both pressures.
Cll-005-8.6: CRD Hydraulics Test I
CRD hydraulic data was taken every 50 psig during the heatup to verify -
proper steady state response throughout the operating pressure range. At rated pressure, transient response of the flow controller was verified through the use of flow steps, gang single notch insertion, and a flow
( controller step to minimum value. System response was satisfactory with l one exception. TER 038-1 and 038-2 documented the difference between CRD total flow and cooling flow being 5 gpm versus 3 gpm allowed. This condition was found to be acceptable, but the flow indicators will be recalibrattd when plant conditions will allow.
l1 79
STARIUP TEST RESULTS - (cot 7f.)
4.6 STI-CSI-006 SRM Performance and Control Rod Sequence (TSN -010, 040)
- 1. 8.1 (OV) SRM Functional Testing I The Source Range Monitor (SRM) System demonstrated its capability of adequately monitoring 1cw neutron flux levels in the reactor core.
The mechanical operability of the system was also demonstrated.
The following Acceptance Criteria were verified during this test.
Level 1
- 1. The Signal-to-Noise ratio for each detector was greater than 2 to 1.
- 2. The minimum count rate of each detector was greater than 0.7 cps.
I Other
- 1. The mechanical operability of each SRM drive was demonstrated.
The test was performed with 652 bundles loaded in the core once all four detectors had been placed in service. Readings were taken for each detector at its fully inserted and fully withdrawn position and
, the signal-to-noise ratios were calculated. All signal-to-noise ratios were greater than or equal to 29. The insert and withdrawal times were recorded for information purposes. All Acceptance Criteria were satisfied.
I I
80
I I STAR'IUP TEST RESULTS - (CONT. )
STI-C51-006 (CONT.)
- 2. 8.2 (HU) SRM Non-saturation The Source Range Monitor (SRM) detectors were verified not to saturate below 300,000 cps.
The following Acceptance Criterion was verified during this test.
l l
Other I No SRM detector underwent saturation prior to registering a count rate of 300,000 cps.
With the plant operating on range 7 of the Intermediate Range Monitoring System each SRM was individually bypassed and inserted into the core. The maximum readings for each detector were recorded and all were greater than 300,000 cps thus satisfying the Acceptance criterion.
4.7 STI-C51-010 IRM Performance (TSN -039, 041, 065)
The purpose of this test is to adjust the Intermediate Range Monitor (IRM)
System to obtain an optimum overlap with the Source Range Monitor (SRM) and Average Power Range Monitor (APRM) Systems. The three tests described
(
l below were successfully completed by the end of 'IC-HU and the required IRM l system overlap with the SRM and APRM systems was obtained.
81
I I STARHJP TEST RESULTS - (CONT.)
STI-C51-010 (CONT.)
Initially the IRM detector gains were set to values determined during preoperational testing. During the initial startups when the flux level was sufficient, the response of-all IRMs to neutron flux was verified.
Neutron flux continued to be increased over a half decade on all IRMs without exceeding the SRM rod block. This verified adequate overlap between the SRMs and IRMs. Additionally, the mechanical operability of each IRM drive was verified at this time.
During reactor heatup to rated temperature, the IRM Range Correlation was verified. The IRM Range 7 readings were adjusted to read within 3% of the Range 6 readings. This adjustment was necessary to compensate for the change in amplifiers between Ranges 6 and 7 in the IRM system.
During the increase to 5% of rated power in Test Condition HU, proper overlap between the IRMs and APRMs were verified. The IRM gains were adjusted for one full decade of overlap with all the APRMs reading greater than 4% of rated power. This adjustment required the SRM-IRM overlap portion to be performed again. No additional adjustments were needed as acceptable overlap with the SRM's was demonstrated.
I The final IRM-APRM overlap test will be performed after the LPRM calibration in Test Condition 1.
Section 8.1 - SRM and IRM Overlap This test was performed a total of three times. It was first performed during initial criticality in conjunction with the full core shutdown margin demonstration. The second and third time were during Test I Condition Heatup.
I n
g
l I
I STARIUP TEST RESULTS - (CONT.)
STI-C51-010 (CONT.)
During initial testing IRM channels A, B, E, F, G, and H were verified to have responded to neutron flux before the SRM's exceeded their rod block setpoint. Additionally, they were verified to have at least a half-decade overlap with the SRM's. Two (2) IRM channels (C & D) did not respond to
- I neutron flux and, therefore, SRM-IRM overlap could not be verified for those channels. This problem was addressed in Test Exception Report (TER-020-1). The resolution of the TER was to repair the affected IBMs and then retest during the next startup.
I IRM channels C & D were retested on September 1, 1986, after replacing the IRM de'cectors. All IBM's responded to neutron flux before the SRM's exceeded the normal rod block setpoint and had at least a half-decade overlap with the SRM's.
I On October 18, 1986, the IRM amplifier gains were adjusted to assure one decade overlap with the APRMs. (This is discussed later in the summary of IRM-APRM overlap testing.) The third SRM-IRM overlap test was performed on November 23, 1986. During this test all acceptance criteria were satisfied without making any additional gain adjustment.
I Acceptance Criteria Level 1 Each IRM channel must be on scale before the SRMs exceed their rod block setpoint which is 1 x 10 5counts per second.
I I
H 83 ll
l I STARIUP TEST RESULTS - (CONT.)
STI-C51-010 (CONT.)
Level 2 I
Each IRM channel must be adjusted so that a half decade overlap with the SRMs and one decade overlap with the APRMs are assured, the IRM I
i.e.,
initial readings on Range 1 show an increase of 50% of scale (1/2 decade) l before exceeding the SRM rod block setpoint.
I only the final set of data obtained during the third performance of this l section is presented below as they reflect the final gain settings of the IRMs.
The table below shows the data related to the IRM-On-Scale verification which indicates that all IRMs satisfy the Level 1 acceptance criterion.
l Reading Values **
I l I
l I
l ! IBM A l IRMBlIRMClIRMDl l l
! 42/2 I
! 56/2 I
l38/2 I I
38/2 I
I I I I I I
l I I I I I 38/2 ! 45/2 ! 49/3 ! 52/2 !
l I I I i l l SRM A l SRM B l SRM C l SRM D l l l (cps) l(cps) l(cps) l (cps) I ** For IRM readings, note reading ,
8000 ! 18000 ! 20000l in Percent and range, e.g.,
l 6000 ! '
l l l l l reading / range.
84 l
STARTUP TEST RESULTS - (CONT.)
STI-C51-010 (CONT.)
The SRM and IRM overlap data are shown below. Part A indicates the initial response of the IRMs on Range 1. Part B shows data recorded at a later time when the IRMs were reading mid-scale on Range 2 (IBM G was on l Range 3). These readings are equivalent to a value greater than 120% on 5
Range 1. Also note that the SRM readings remained below 1 x 10 s. We data in Part B indicate more than a 50% of scale increase without exceeding the SRM rod block. Consequently, the Level 2 acceptance criterion was met for all IRMs.
PART A PART B l
l Reading Values ** l l Reading Values ** l l
l l l l l
! IRM A ! IBM B l IRM C ! IRM D ! ! IBM A IRM B IRM C IRM D !
I I I I I I I I l l
! ! ! ! ! ! ! 56/2 ! 38/2 ! 38/2 !
10/1 12/1 G/1 8/1 42/2 l l l 1 I I l l l l l
! IRM E IRM F i IRM G l IRM H ! l IRM E IRM F IRM G IRM H l l l l l l l l l l l
! ! ! ! ! ! ! 45/2 ! 49/3 I 52/2 !
6/1 9/1 24/1 12/1 38/2 a l l l l l l l l l l lSRMAlSRMBlSRMClSRMDl l SRM A l SRM B l SRM C l SRM D l l(cps) l (cps)9 I (cps) I(cps) l l(eps) l(eps) l(eps) l(cps) l
! 9000 ! 7000 ! 20000 ! 23000 I ! 6000 I 8000 ! 18000l20000 !
l l l l ] l l l l l
- For IRM readings, note reading in percent d range, e.g.,
reading / range.
E 85
STARnJP TEST RESULTS - (CONT. )
STI-C51-010 (CONT.)
The final SRM readings in PART B appear lower than the initial readings in PART A. This is because the SRM mechanism was partially withdrawn to a lower neutron flux region in the core. Withdrawing the SRMs is normal operating practice during a reactor startup.
Section 8.2 - IRM Range Correlation Adjustment During the startup on September 1, 1986, the IRM range 6 to 7 correlation adjustments were made per this startup test section. The purpose of this test was to establish good correlation between range 6 and 7 indications for the eight IRM channels. This is necessary, because, in the IRM circuitry, signals pass through one pre-amplifier for ranges 1 through 6 and another for ranges 7 through 10. The range 7 pre-amplifier gains were adjusted until agreement was obtained between the indications of ranges 6 and 7 for each IRM.
Comparison of Results with Acceptance Criteria:
There is no Level 1, 2, or 3 Acceptance Criteria associated with this test l section. However, the criteria for the adjustment is that range 6 and 7 agree within 3%. The "As-Left" data for the IRMs are shown below:
i I
86 1
STARRIP TESk' RESULTS - (CONT. )
I STI-C51-010 (CONT.)
Readings Equivalent Percent Range 6l Range 7 Range 7 Reading Difference IRM (1) l (2) (2)
- 3.1623 (2) - (1) x 100 l (1) l A 74 l 24 75.9 2.6 B 77 l 25 79 2.67 C 47 l 15 47.4 0.9 D 67 l 21 66.4 -0.9 E 71 l 23 72.7 2.4 F 91 l 29 91.7 0.8 G 75 l 24 75.9 1.2 H 77 l 24 75.9 -1.4 As shown above all IRM range 6 and 7 indications were adjusted to correlate to within 3%. There were no problems related to IBM performance noted in this test.
Section 8.3 - IRM-APRM Overlap On October 18, 1986, the IRM amplifier gains were adjusted to assure one decade of overlap with the APRMs per this Startup Test Section. Reactor power was increased to > 4% on all APRM channels. At this point each IBM was adjusted so they read below 108% on range 8.
- These amplifier gain adjustments made do not affect the pre-amplifier associated with the IRM range 6 and 7 correlation. However, they do influence the SRM-IBM overlap results which required a third performance of that test.
87
I .
STARIUP TEST RESULTS - (CONT.)
STI-C51-010 (CONT.)
Comparison of Results with Acceptance Criteria:
Level 1 Each APRM must be on scale reading greater than or equal to 4%, before the IRM's exceed their rod block setpoint of 108% on Range 10.
Level 2 Each IRM channel must be adjusted so that one decade of overlap with the APRM is assured. -
One decade of overlap is defined as the IRM readings are not greater than 108% on range 8 at the point that all APFMS are reading greater than 4%.
The data below shows all the IRM readings immediately following their adjustments on range 8. Also shown are the maximum and minimum readings for each APRM that was recorded during the adjustment period.
I APRM Reading ll IRM readings I APRM (% of Rated Power)
Max Min ll ll IRM
(%, scale is 0-125%)
Range 8 ll A 4.7 4.45 ll A 100
- B 4.8 4.6 ll B 100 C 4.2 4.1 ll C 100 D 4.3 4.15 ll D 94 E 4.5 4.3 ll E 104 F 4.7 4.5 ll F 100 G 4.3 4.2 ll G 100 H 4.4 4.2 ll H 99 I
88
I STARIUP TEST RESULTS - (CONT.)
I STI-C51-010 (CONT.)
Each IRM circuitry to the Reactor Protection System was bypassed before its adjustment was made. Before placing the adjusted IRM back in service, it was set on range 9 as dictated by normal operating practice.
Consequently, the last readings obtained after all IRMs had been adjusted are shown below.
! IRM A IRM B l IRM C IRM D l l l l l
! ! 34/9 ! 34/9 I 30/9 !
34/9 l l l l l I 1RM E IRM F IRM G l IRM H i l I I I I
! ! I I 33/9 I 34/9 33/9 33/9 l l l l l I MMA ! MMB ! MMC ! MMD l l l l l l
! 4.45 I 4.70 ! 4.10 l 4.15 !
l l l l l
! MME l MMF MMG l MMH I I I I I
! 4.30 ! 4.60 I 4.25 ! 4.20 !
l l l l l NCyrE: 1. For IRMs, is shown as the reading in percent and the IRM range, e.g., reading / range.
- 2. APRM readings are in % Power.
- 3. 34 on range 9 is equivalent to 107.5 on range 8.
I 89
I STARETP TEST RESULTS - (CONT. )
I '
STI-C51-010 (CONT.)
As shown above, the Acceptance Criterion for all IRMs were met. There were no major problems encountered during this Test. This test will be repeated once more in Test Condition 1 after the LPRMS are calibrated at power.
I 4.8 STI-C51-011 LPRM Calibration (TSN-066)
This test verified the flux response of the LPRM's when a control rod I adjacent to a particular LPRM string was withdrawn from the full in position to the full out position.
Acceptance Criteria There is no applicable criteria for this test.
This test was performed in conjunction with STI-C11-005 Section 8.3 titled Single Rod Scram Timing. The various rods moving past the LPRM's in the core supplied the changes in flux needed to verify the LPRM's response.
If a rod about to be withdrawn was near an untested LPRM, the switch at the appropriate APRM meter was selected to feed the LPRM signal to the I ERIS-TRA computer. The computer recorded the LPRM response as the rod was continuously withdrawn per STI-C11-005. When the rod was fully withdrawn the recording stopped. A plot of the LPRM response was done for each LPRM string to depict the response.
It was expected that at the low power level at which the test was performed (less than 5% power), a number of LPRM's would not show any response. Out of the total number of 164 LPRM's there were 65 LPRM's that did not show any response. TER 066-01 was initiated to track the testing of all 65 LPRM's. They were generally in the lower core levels. Four (4) of the 65 LPRM's did not respond due to APRM switches apparently not l
1 l 9 lE3 l
I STARTUP TEST RESULTS - (CONT.)
STI-C51-Oll (Covr.)
making contact during the test. This would have prevented the signal from going to the computer to be recorded. In the case of one of those four LPM's, the wrong APM switch may have been selected. These four LPM's were retested satisfactorily. The remaining 61 LPM's will be retested at a higher power in TC-1.
4.9 STI-C51-012 APM Calibration (TSN-048, 072)
I This test adjusted the APM gains at low reactor power based upon a constant heatup rate heat balance. Prior to this test the APM gains were set at maximum.
Acceptance Criteria Level 1
- 1. The APM channels must be calibrated to read equal to or greater than actual core thermal power.
In the start-up mode, all APM channels must produce a scram at less I 2.
than or equal to 15% of rated thermal power.
l l Level 2 None j Three constant heatup APM calibrations were performed during Test I Condition Heatup. The first test collected heat balance data with the reactor in control rod sequence "A" and a reactor heatup rate of l 45'F/ hour. The heat balance indicated a core thermal power of 0.51%. The 91
I I STAR'IUP TEST RESULTS - (CONT. )
I STI-C51-012 (CONT.)
APRM Gain Adjustment Factors (GAF) were determined and the APRM gains were later adjusted. In addition, the start-up scram setpoint for each APRM was verified to be less than 15%.
I In order to accommodate Control Rod Drive testing, the reactor was shut down then restarted in the "B" control rod sequence. Due to the sequence change another heat balance was performed with a reactor heatup rate of 45'F/ hour. The heat balance indicated a core thermal power of 0.58%. The APRM Gain Adjustment factors were calculated, the APRM gains were later adjusted and the scram setpoints were verified to be less than 15%.
The third APRM calibration test was performed to ensure the APRM's were i calibrated immediately prior to the planned main generator synchronization at less than St core thermal power. A heat balance was performed on l
December 15, 1986 with a reactor heatup rate of 50*F/ hour. The heat balance indicated a core thermal power of 0.58%. The APRM Gain Adjustment factors were calculated, the APRM gains were later adjusted and the scram setpoints were verified to be less than 15%.
All Acceptance Criteria for the three Constant Heatup Rate APRM Calibra-tion tests were satisfied. The following is a table of significant test results.
SIGNIFICANT TEST RESULTS APRM Reading Scram
_GAF As Found As Left Setpoint September 30, 1986 Results APRM - A 0.402 4.3 2.0 14.25 i B 0.589 3.75 2.3 14.88 C 0.378 4.3 2.4 14.81 l
D 0.422 4.6 2.0 14.75 l E 0.378 4.6 2.0 14.38 92 l
I I STARTUP TEST RESULTS - (CONT.)
STI-C51-012 (Cottr. )
SIGNIFICANT TEST RESULTS (CONT.)
APRM Reading Scram GAF As Found As Left Setpoint September 30, 1986 Results (Cont.)
APRM - F 0.429 4.5 2.0 14.26 G 0.390 4.4 2.0 14.18 H 0.378 4.2 2.0 14.50 October 14, 1986 Results APRM - A 0.79 3.85 3.60 14.34 B 1.03 3.80 4.00 14.90 C 0.90 3.30 3.60 13.96 0 0.81 3.90 3.70 14.76 E 0.80 3.80 3.60 14.13 F 0.82 4.10 3.70 14.13 G 0.74
- 3.90 3.60 14.00 H 0.73 4.50 3.70 14.40 December 15, 1986 Results*
l I APRM - A 0.76 1.50 1.20 14.88 B 0.97 1.50 1.50 14.88 C 0.673 1.80 1.25 14.50 D 0.83 1.72 1.45 14.94 E 0.834 1.50 1.30 14.73 F 0.86 1.60 1.40 14.68 G 0.713 1.52 1.10 14.88 H 0.83 1.55 1.30 14.69
- Gain adjustments were done on December 16, 1986.
93
STAR'IUP TEST RESULTS - (CONT. )
4.10 STI-C91-013 Process Computer (TSN-024, 025, 088)
TIP Alianment After TIP System Installation (8.1)
The purpose of this test is to hand probe all channels of the five (5)
Traversing Incore Probe (TIP) machines to establish the full-in limits of detector travel and then to calculate the initial (cold) settings of core top and core bottcm limits for the Process Computer OD-1 Program.
Forty-two (42) of the total forty-nine (49) channels were found to be in agreement with the as-left values of the core limits determined during preoperational testing. Any differences were one inch. There is no test criterion applicable to this test.
Acceptance Criteria Level 1, 2, and 3 There is no test criteria related to this test.
Results l The TIP alignment data at open vessel conditions are tabulated in Attachment 1.
Discussion TIP alignment was performed by using the hand crank for each drive mechanism located in the containment. The TIP's were carefully driven into the core using this method in order to avoid damage to the TIP detector when it touched the top of the TIP tube and to determine the TIP
- panel reading at this location (NCFI). The core top (NCCT) was then defined by subtracting a turnaround margin of 2 inches. Core bottom (NCCB) for each TIP location is defined by subtracting 144 in. from NCCT.
l NCCT and NCCB were then set into the TIP panel electronics. These values 1
94 l
l
STAR'IUP TEST RESULTS - (CONT. )
STI-C91-013 (CONT.)
were determined in order to define a Core Top and Core Bottom limit between which the TIP detectors will scan when performing a TIP trace.
The gama flux will be transmitted to the process computer between these limits when performing an 0D-1 or OD-2. This test was repeated during Test Condition Heatup to allow for any system expansion during hot vessel conditions.
TIP Alignment After Reactor Startup (8.2)
The objective of this test was to hand probe the TIP channels with the plant operating at rated temperature and pressure to compare with the respective readings obtained when the plant was in the cold condition and then to adjust core top and bottom limits for subsequent TIP traverses, as necessary, the purpose of which is to prevent damage to the TIP detectors while they are operated in automatic mode.
All TIP channels registered a zero or positive increase in the travel from their respective reference positions to the tops of their dry tubes in going from the cold to the hot condition. Consequently, it was unnecessary to adjust any channel's upper and lower limits in this test.
l Acceptance Criteria There is no test acceptance criteria applicable to this test.
1 Results j The following table shows, for all TIP channels, the Veeder-Root counter l readings for the detector full-in positions and the corresponding core top l and the core bottom indications for the cold plant condition. In each l case, the hot - or "As-Left" - core top and bottom indications were the same as the cold.
95
STAR'IUP TEST RESULTS - (CONT. )
STI-C91-013 (CONT.)
TIP Alignment Data Attachment 1 Core Index NCFI NCFI Delta NCCT/NCCB TIP Machine Coordinates Number Hot Cold NCFI Cold A 24-25 1 949 948 +1 946/802 16-33 2 935 935 0 933M89 16-25 3 962 961 +1 959/815 16-17 4 981 980 +1 978/834 08-17 5 994 993 +1 991/847 08-25 6 963 963 0 961/817 08-33 7 950 949 +1 947/803 I 08-41 08-49 8
9 926 901 925 900
+1
+1 923M79 898/754 32-33 10 930 929 +1 927M83 B 24-41 1 931 930 +1 928M84 24-33 2 957 957 0 955/811 24-25 3 986 986 0 984/840 24-17 1004 1003 +1 1001/857 I
4 24-09 5 1027 1026 +1 1024/880 16-09 6 1044 1043 +1 1041/897 16-17 7 1011 1010 +1 1008/864 l
16-41 8 947 946 +1 944/800 l
l 16-49 9 926 925 +1 923/779
,- 32-33 10 966 965 +1 963/819 l
j C 32-57 1 852 851 +1 849/705 32-49 2 875 874 +1 872M28 32-41 3 911 910 +1 908M64 32-25 4 968 967 +1 965/821 1
96
I STARRIP TEST RESULTS - (CCtfr.)
STI-C91-013 (CONT.)
C (CCNT.) 32-17 5 969 968 +1 966/822 32-09 6 991 990 +1 988/844 24-49 7 897 896 +1 894/750 24-57 8 873 872 +1 870/726 32-33 10 939 938 +1 936/792 D 48-49 1 872 871 +1 869/725 l
48-09 2 983 982 +1 980/836 40-57 3 865 864 +1 862/718 40-49 4 890 889 +1 887/743 40-41 5 928 927 +1 925M81 40-33 6 945 944 +1 942M98 40-25 7 965 964 +1 962/818 40-17 8 979 978 +1 976/832 40-09 9 1004 1004 0 1002/853 l 32-33 10 969 968 +1 966/822 E 56-41 1 846 845 +1 843/699 l
56-33 2 870 869 +1 867/723 56-25 3 896 895 +1 893/749 l
48-17 4 959 959 0 957/813 48-25 5 933 933 0 931/787 48-33 6 910 909 +1 907/763 48-41 7 854 853 +1 851/707 48-49 8 842 842 +1 840/696 40-41 9 924 923 +1 921/777 32-33 10 972 971 +1 969/825 Where Delta NCFI = NCFI (hot) - NCFI (cold) l 97
STARTUP TEST RESULTS - (CONr.)
STI-C91-013 (CCNr.)
Discussion This test was performed in essentially the same manner as was done when the plant was cold (TC-open vessel): all channels of all TIP machines were probed (using the hand crank) to place the detector at the physical stop of the full-in position. At each such point, the Veeder-Root counter reading was recorded and set equal to the Number of Counts Full-In, or
! NCTI (hot). These values were then compared with their respective counterparts, NCFI (cold) obtained during open vessel testing, and the I difference, NCFI (hot) - NCFI (cold) (= Delta NCFI) determined for each machine / channel.
I For values of Delta NCFI which might be negative, an abnormal characteristic of thermal expansion, i.e., a shortening of the TIP traverse from the reference (indexer) position to the top of the dry tube, would be indicated, and it would be necessary to decrease the core limits Number of Counts at Core Top /Bottem (or NCCT/NCCB) - accordingly, i.e., by Delta NCFI, so as not to risk damaging the detector (s) by reducing the amount of turnaround at the top of the traverse stroke. For values of Delta NCFI which were zero or positive, however, no increase in NCCT/NCCB would be required since the TIP plumbing is presumed to expand a greater amount (owing to its length) than vessel internals hardware in the course of heatup. Since all values of Delta NCFI proved to be either zero or positive, no adjustments to NCCT/NCCB for any TIP/ channel were necessary, and the cold settings were left as they were set in TC-OV.
After the completion of the NCFI (hot) determination, the TIP plotter horizontal gains and zeroes were adjusted for the individual TIP
! machine / channels such that the recorder set down on the paper at core top (NCCT) and picked up at core bottom (NCCB). During this phase of the l
i test, it was determined that the governing plant procedure for these 1
98
STARRIP TEST RESULTS - (COffP. )
STI-C91-013 (Carr.)
adjustments could not be performed correctly as written. TER-088-1 was initiated and the plant procedure was changed after consulting the appropriate GEK manual. The entire plotter set-up step was repeated with satisfactory results.
DSTC - Prestartup Operation (8.5) l The purpose of this test was to demonstrate that process computer programs OD-3, OD-6, 00-7, CD-8, CD-9, OD-10, CD-11, CD-13, CD-16, OD-17, CD-20, I P1, P2, and P3 can be successfully demanded and run on the MONICORE (VAX) computer. A file of dummy input data was used as the basis for the calculations performed. In addition, programs 00-3, OD-7, and CD-8 were demonstrated to run successfully on the Honeywell computer, using actual plant input data.
Several software problems were encountered during preparations to run the test and were resolved with GE-NEBO computer group assistance.
No problems with any of the programs were noted in the performance of this test.
I Acceptance Criteria No acceptance criteria is applicable to this test section.
Results:
All edits were obtained for all process computer programs accessed in this test, and dummy CD-1 data was successfully transferred from the Honeywell to the VAX.
I I
l 99 1
I STARTUP TEST RESULTS - (CONT.)
STI-c91-013 (CONT.)
Discussion This test was done as a preliminary software checkout for the eventual performance of the Dynamic System Test Case (DSTC) and showed that the available OD- and P- programs installed in the Perry 1 process computer software acutally run when requested, that the resulting edits contain no obvious errors, and that edited values are consistent from program-to-program (where it was possible to check). Additionally, OD-10 was used to I alter and restore selected data in several of the editable arrays. The data used was provided by GE-NEBO as part of the process computer software. The only actual plant data used in this test was in CD-3, OD-7, and OD-8 edits accessed from the Honeywell 4400. Dumy OD-1 data was demonstrated to have been successfully transferred from the Honeywell to the MONICORE VAX via the Megastore bulk storage device. Finally, after resolution of mi tt software problems, 0D-20 was verified to have initialized thu vartous single-valued and array data in the manner I prescribed for a new reactor.
All edits were requisted from the system manager's I/O console (in the case of the MCNICOF$ VAX) and the I/O typer and CRD console (for the l Honeywell 4400).
Tnere were no problems noted in either the execution of the various software programs accessed during the test or in the consistency of values of certain arrays from edit-to-edit.
1 1
100
STARTUP TEST RESULTS - (CCtTP. )
4.11 STI-ESl-014 Reactor Core Isolation Cooling (TSN-043, 068)
The Reactor Core Isolation Cooling System (RCIC) demonstrated proper operation at the minimum and rated pressures and flow ranges when operating in the recirculation mode to the condensate storage tank (CST).
Reliability in the automatic quick starting mode was also demonstrated with the reactor at rated conditions and at 150 psig. Only quick starts from hot conditions have been performed; hcwever, cold quick starts are planned for later Test conditions.
Acceptance Criteria Level 1
- 1. The average pump discharge flow must be equal to or greater than the 100% rated value after 30 seconds have elapsed from automatic initiation (or manual push button start) at any reactor pressure between 150 psig and rated.
NorE: If Level I criteria is not met, the reactor will only be allowed to operate up to a restricted power level defined by Attachmnt 1 (of STI-ESl-014) until the problem is resolved. Also consult the plant Technical Specifications for actions to be taken.
- 2. The RCIC turbine shall not trip or isolate during auto or manual start test.
l*
101 i
(
STARTUP TEST RESULTS - (CONT.)
STI-ES1-014 (CONT.)
Level 2
- 1. In order to provide an overspeed and isolation trip avoidance margin, for transient starts, the first and subsequent speed peaks shall not exceed 5% above the rated RCIC turbine speed.
- 2. The speed and flow control loops shall be adjusted so that the decay ratio of any RCIC system related variable is not greater than 0.25.
- 3. The turbine gland seal system shall be capable of preventing steam leakage to the atmosphere.
The RCIC system demonstrated its reliability by always satisfying the Level 1 acceptance criteria. Section 8.1 of STI-E51-014, RCIC, was performed at 140 psig, 162 psig, and rated reactor pressures. The testing at 162 psig was performed in response to operational problems encountered during curveillance testing of the RCIC system at 150-165 psig. The only problem encountered during the performance of section 8.1 was that for the 140 psig and rated pressure testing, valve lE51-F046 (RCIC PUMP DISCH TO LUBE OIL COOLER) failed to open automatically upon RCIC initiation and had to be manually opened. The problem was later corrected by adjusting the torque switch on the 1E51-F046 operator.
Section 8.2 of STI-E51-014 was performed and demonstrated satisfactory operation of the RCIC system over a two hour continuous period with pump flow at 700 gpm or greater. This test was conducted in conjunction with the quick start demonstration of Section 8.1 at rated pressure. A small amount of water was observed dripping from the bottom of the RCIC turbine casing during the two hour run. The insulation was removed and this area is to be inspected for leakage during the next performanco of STI-t51-014.
102
STARWP TEST RESULTS - (CONT. )
STI-E51-014 (CONT.)
Dates, conditions, and results of RCIC testing is shown in the table below.
Rx TIME 'IO TRIP OR SPEED OSCILLA- SEAL DATE TEST PRESS RATED FIIM ISOIATE (4777) TIONS LEAKS?
9/15/86 8.1 140 30see no 2712 accept -
9/29/86 8.1 162 28 no 2850 none 10/8/86 8.1 920 16 no 4370 ncne -
10/8/86 8.2 920 -- - - yes 4.12 STI-B21-016A Selected Process Temperatures (TSN-075)
This test was run to determine the flow control valve position that causes temperature stratification in the bottom head region of the reactor.
I Steady state temperature data from reactor recirculation loops and bottom head drain were taken while the A and B flow control valves were closed in 10% increments. Data was collected until the rcy's are at their mininum position, the delta T between the steam dome and bottom head drain reaches 100 degrees Fahrenheit, or recirculation flow is unstable, whichever occurs first.
Acceptance Criteria Level 1 None 103
STARTUP TEST RESULTS - (CONT. )
STI-B21-016A (CCNT. )
Level 2 None Other Due to flow oscillations occurring at a FCV position of 20%, it was recomended that a precautionary note be added to SOI-B33. The precaution warns the operator not to run the recire. system on the LFMG with the FCV f 201.
During data collection, it was noted that the bottom head drain tempera-ture utilizing thermocouple NO22 was reading low. It was compared to a second thermocouple reading bottom head drain temperature approximately 3 feet farther downstream (thermocouple N021), and the difference in the two readings was approximately 60*F. Even with this low bottom head drain temperature, the 100*F delta T between the steam dome and the bottom head drain was never exceeded.
A work order was generated to fix thermocouple N022. After inspection, it was discovered that the thermocouple was not properly grounded and one wire of the thermocouple was broken in the well. The dual element thermo-couple has since been replaced. The thermocouple will again be used and checked in TC-3, for Section 8.2-Drainline Thermocouple Data.
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STAR'IUP TEST RESULTS - (CONT. )
4.13 STI-B21-0168 Water Level Reference Leg Temperature (TSN-032, 061)
The Shutdown Range Level Instrument System (section 8.2) was demonstrated to be calibrated correctly based on an evaluation between assumed initial calibrated conditions and actual measured conditions. Temperature data was c-Ilected near e e instruzent reference leg during cold, shutdown conditions with stable Containment and Drywell temperatures and 0 psig Reactor pressure.
Acceptance Criteria Level 1 None Level 2 The difference between the actual reference leg temperature and the value assumed during initial calibration shall be less than that amount which results in a scale end point error of 1% of the instrument span for the Shutdown Range level instrumentation.
l The Shutdown Range Level Instrument System was demonstrated to completely meet the Level 2 Acceptance Criteria with a scale end point error well below 1%, and therefore no additional calibration was required.
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STARRIP TEST RESULTS - (CONT.)
STI-B21-016B (CONT.)
W e Narrow, Wide and Upset Range Level Instrument Systems (section 8.1) -
were demonstrated to be calibrated correctly based on an evaluation between assumed initial calibrated conditions and actual measured conditions. Temperature data was collected near the instrument reference legs during low power operation (- 4.5%) with stable Containment and Drywell Temperatures and rated Reactor temperature and pressure.
Acceptance criteria Level 2 The difference between the actual reference leg temperature (s) and the value(s) assumed during initial calibration shall be less than that amount which results in a scale end point error of 1% of the instrument span for the Narrow, Wide and Upset Range level instrumentation.
The Narrow, Wide and Upset Range Level Instrument Systems wero denenstra-ted to completely meet the Level 2 Acceptance Criteria with scale end point errors below 1%, and therefore, no additional calibrations were required.
4.14 Vibratien and Wermal Expansion T g STI-021-017 NSSS hermal Expansion (TEN -033, !)a2, 006)
STI-B?.1-033 NSSS Vibration (TSN -027, 028, 034)
STI-P99-122 BOP W ermal and Vjbration (TSN -054, 055)
The Startup Test Piping Program is governed by three separate Startup Test Instructions. @e combined requirements of the STI's account for transient / steady state vibration and thermal expansion testing, on Nuclear
)
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STAR'IUP TEST RESULTS - (CONT. )
Piping Tests (Cont.)
Steam Supply System (NSSS) and Balance-of-Plant (BOP) Piping / Suspension systems. h e NSSS scope is the Main Steam and Reactor Recirculation I Piping Systems, including small attached branch piping. The BOP scope is made up of Main Steam, Main Steam Drain Lines, Fuel Pool Cooling and Cleanup, RHR, RCIC, MSIV Leak Detection and Feedwater Piping. The BOP scope of pipe testing completes a test program that was started during the preoperational phase of plant testing. The systems included in the Startup Test Program could not be adequately tested without the plant in operation. The individual systems will be discussed as part of the specific test in this synopsis. Test methods for this program include visual and remote monitoring. Visual inspections were performed by:
Qualified Piping Inspectors (QPI) and Qualified Piping Engineers (QPE). '
The inspection reports were evaluated by the QPE. These individuals were primarily engineers and designers from the Nuclear Engineering Department s (NED), who were trained and supervised by the Test Director. Remote monitoring devices used through the ERIS computer were lanyard potenio-meters for measurement of thermal and vibration displacement, strain gages, RTD's and accelerometers. ,
Thermal Expansion Testing The purpose of thermal expansion testing is to confirm that the pipe l suspension system is working as designed and that the piping is free to I
expand without obstruction. Data was collected using lanyard potentio- ,
meters to measure expansion of the piping, and visual inspection to check for adequate clearance and measure cold and hot positions of support I components: spring cans, snubbers and whip restraints. termal expansion testing on NSSS pipe is performed during three separate cycles, from ambient to operating temperature. Each test is evaluated individually, and the results from all three are compared for repeatability and 107
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STARIUP TEST RESULTS - (CONT.)
>?
Piping Tests (Cont.)
\
consistency. Thermal expansion testing for BOP piping was scheduled for performance 'only one time per system, however, two retests due 1:o failure to meet criteria were also performed.
1 Visual inspection commenced as a prerequisite to actual test performance.
I, Prior to heatup the piping, equipment, and supports were inspected at ambient temperature to determine if any interference existed based on projected design thermal movements, including a clearance envelope, at rated temperature. At this time, cold readings were also recorded for
' \ m, snubbers and spring cans to serve as a baseline for later inspections to calculate pipe growth. Pipe whip restraints were also inspected during l3 this program but were treated as structural items that required adequate clearance. Additionally, the ERIS computer program, EXPAND, was I initialized, that is, the lanyard pot settings were recorded by the computer at ambient conditions to serve as a "zero" or baseline for remote expansion measurements.
g Test performance during the first thermal cycle required running the ERIS i1
/ , ' EXPAND" program at 50*F intervals throughout the heatup. At each interval the program was executed and a printout was provided for each remote sensor. Information on the printout included the initial and g present position of the respective pipe, the temperature from a related remote temperature element for that particular reading, and the adjusted Level 1 criteria for that intermediate temperature. The criteria was adjusted, or approximately interpolated from the rated temperature Level 1 criteria by the EXPAND program. This data was reviewed by the QPE's and I Test Director for appropriate expansion. An intermediate temperature (250'F - 350'F) visual walkdown was performed during the first heatup.
, [ The visual inspection did not require the recording of any data for hanger positions, but was strictly a clearance check at the half-way point to 108
STARIUP TEST RESULTS - (CONT.)
Picing Tests (Cont.)
ensure proper pipe response. At rated temperature, a final EXPAND run was executed to record operating temperature position and also the visual l inspection was repeated to record the hot hanger position and check for l the required clearance envelope. This data was then compiled and evaluated to verify proper piping response at operating temperature conditions. BOP systems for this test were E51 (visual) and B21, Cll, E32, E51, and N22 (remote).
As mentioned previously, NSSS piping was tested for thermal expansion I during the second and third thermal cycles. Test performance consisted of remote ERIS data collection only. EXPAND runs were executed at three temperature intervals: ambient, intermediate and rated. Data was again evaluated to verify proper pipe response based on expansion criteria.
Additionally, following completion of the third cycle, the three sets of NSSS expansion data were compared to verify consistency and repeatability.
Although scheduled for performance only once, selected BOP piping was retested during the second and third cycles in a similar fashion. Retests were required due to failure to meet criteria in previous tests and were performed subsequent to rework to correct hardware associated with the violation or reanalysis that revised the Acceptance Criteria.
l Thermal testing was also performed on BOP piping outside of the overall nuclear heatup testing previously described. At a reactor pressure of 250 psig, a functional test of the safety / relief valves (SRV) was performed.
During this test, remote thermal expansion measurements were recorded on the B21 SRV tailpipes and attached piping during and after discharge.
l Due to the transient nature of the heatup, displacements were analyzed l using a time history plot of the recorded potentiometer signal, rather than the EXPAND routine previously described. Visual and remote thermal testing was also performed during the RCIC two-hour run at rated pressure.
Methodology was similar to that described for plant heatup.
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I I STARTUP TEST RESULTS - (CONT.)
I Piping Tests (Cont.)
Thermal Expansion Test Results Test results from NSSS and BOP piping thermal expansion testing were successful in verifying the analytical model of pipe expansion and also in identifying and correcting interferences to thermal expansion. Prere-quisite ambient temperature walkdowns identified several hundred existing and potential problems based on projected design thermal movements and required clearance envelopes. These problems were routed to NED for analysis and corrective measures prior to the first heatup. During the heatup, test data was imediately reviewed by the QPE as it became availa-ble. The primary document used to record communication between the team
( of QPI/QPE's and the Test Director during test performance was the Test Evaluation Form (TEF). The TEF was used by the QPI to identify items which were generated during the performance of these tests. The TEF played an important role during the plant heatups because, per the Test Specifications, application of acceptance criteria is not appropriate until reaching rated temperature. As a result, potential criteria j violations which were identified during the heatup did not necessarily warrant a Test Exception. Instead, Engineering evaluated the condition to determine acceptability of continued heatup and operation, and documented the evaluation on a TEF.
Following resolution of problems identified during the prerequisite walkdown, the first nuclear heatup commenced. ERIS EXPAND runs were performed, as required, and demonstrated proper pipe expansion up to the intermediate temperature walkdown. At that time, an impending inter-ference was discovered on BOP piping, the plant was placed in a hold condition, and rework was performed to clear the interference. Following completion of rework, heatup continued. At 450*F, ERIS data indicated improper pipe movement on the RCIC Steam Supply Line. Again, plant 110
STAR'IUP TEST RESULTS - (CONT. )
Piping Tests (Cont.)
conditions were stabilized, a drywell entry was made, and it was determined that the E51 bypass line valve F076 was bound in a support.
Design changes were issued, rework was performed, and heatup continued.
Several other sensors showed movements outside the criteria limits during the heatup, however, engineering analysis allowed continued heatup to operating temperature.
I At operating temperature following the first nuclear heatup, all data was collected and reviewed and exceptions were noted as follows: Based on visual observation and measurement, several supports on NSSS piping were noted as being outside of the hot design setting and piping was found in contact with several whip restraints, causing slight compression of the insulation. Several remote sensors recorded displacements on NSSS pipe which were outside the Level 2 displacement criteria. General Electric
'I (GE) evaluated all violations and determined the conditions as acceptable.
Remote sensors on BOP piping recorded six violations of Level 1 expansion criteria and ten violations of Level 2 criteria. RCIC thermal expansion testing of BOP pipe performed during the first heatup identified one Level 1 and seven Level 2 violations on remote sensors, several of which duplicated violations recorded in the reactor heatup. Overall, eight of these Level 1/2 sensor violations were designated as requiring retest during the next thermal cycle following appropriate rework or reanalysis.
The remainder were determined to be acceptable. Visual inspection of BOP piping Lnd supports during the first heatup identified a single support outside the Level 2 limits, but was determined to be acceptable. BOP thermal expansion testing during the SRV functional test yielded satis-factory results, 111
STARTUP TEST RESULTS - (CONT.)
Piping Tests (Cont.)
Testing of NSSS pipe during the second thermal cycle yielded results similar to the first cycle. Seventeen sensors recorded Level 2 expansion violations because of inadequate expansion (less than design). At this l point it is important to note that the violations were among the same sensors identified in Cycle #1, and also that the ERIS EXPAND program does not adjust Level 2 criteria to the actual temperature as is done for Level 1. Therefore, it is not unreasonable to expect violations of this nature if the piping does not reach its design temperature. As before, and based on the above reasons, GE determined the violations as accept-able. Also during the second cycle, a visual inspection of a recircu-lation loop snubber was made at the request of GE to verify acceptability.
! The inspection showed satisfactory performance. The BOP retest of the eight sensors which failed criteria resulted in a plant hold at the inter-mediate temperature interval, with temporary operational constraints due to an expansion violation on the B21 head vent line. Final test results showed successful correction and testing of four sensors and designated retest during thermal cycle #3 for the remaining four sensors, including l
the head vent line. Appropriate rework and reanalysis was designated for
! completion prior to retest.
Test results for NSSS piping during thermal Cycle #3 were predictable, with similar Level 2 violations. GE evaluation again determined the situation acceptable. Also, as required following the final cycle, a comparison was made on data from all three cycles. Results were satis-factory and indicated consistent and repeatable piping respanse. This 112
I STARIUP TEST RESULTS - (CCNr. )
Piping Tests (Cont.)
completed thermal expansion testing of NSSS piping. Retesting of the four l
BOP sensors again placed the plant on hold at the intermediate temperature l interval due to a Level 1 violation of the head vent criteria due to insufficient growth, identical to previous tests. Following evaluation by Engineering, the heatup resumed. At rated temperature, two of the four sensors indicated Level 2 violations. These were evaluated by engineering and determined to be acceptable. The remaining two sensors indicated Level 1 violations, one on the head vent piping and one on a main steam drain line in the steam tunnel. At this time, the violations remain with the Engineering Department for further analysis. There is no restriction to plant operations, however, NED has set a limit on the total number of complete thermal cycles allowed, that is, ambient to rated to ambient.
This limit is due to the condition of the head vent line. The present outlook regarding the violation calls for hardware changes and retest of I the steam line drain line and formal reanalysis and revision of acceptance criteria for the head vent line. As a footnote, a violation which resulted in revised criteria during the second thermal cycle subsequently re wlted in new support loads based on the reanalysis. The new loads called for the addition of neveral gusset plates on E32 supports, but did I, not affect the expansion test results.
Overall, thermal expansion testing was very satisfactory. It has been demonstrated that the piping response to expansion is in agreement with the analytical model and that the piping suspension system is performing its intended function per design. It has been demonstrated that the piping is free to expand without obstruction from other plant components.
I This completes the required testing on NSSS piping, and satisfactorily demonstrates piping response on much of the BOP piping. Thermal testing
. during various plant conditions remains to be completed on BOP piping, particularly the RHR system.
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STARIUP TEST RESULTS - (CONr.)
Piping Tests (Cont.)
vibration Testing The purpose of vibration testing is to demonstrate, by means of visual inspection and remote sensor signals, that steady state and transient induced piping vibration are within acceptable limits for designated portions, and during specified conditions, of NSSS and BOP piping systems.
Remote sensor data was collected using lanyard potentiometers (NSSS, BOP),
accelerometers (BOP) and strain gages (NSSS). These lanyard potentiometers are, in most cases, the same instruments used for thermal expansion. The signal is split electronically to provide two circuits, one DC circuit (thermal) and one AC circuit (vibration) and processed through the ERIS computer. The NSSS strain gages are located on piping and selected safety relief valve nozzles, also wired directly into ERIS.
The BOP piping is also monitored using accelerometers, which are wired directly to a pulse code modulation (PCM) recorder. Because of the extremely high signal throughput, the accelerometer signal is recorded directly by the PCM, and then played back to the ERIS system at 1/10 the
[
i speed for archiving and data analysis. Data manipulation typically utilizes the time history plot and statistical analysis functions to determine peak displacement for comparison to acceptance criteria. Accel-erometer criteria was provided in "g's" acceleration and also " inches" displacement. The STI provides for initial comparison of acceleration to criteria as the most conservative approach, and then conversion to dis-placement in the event that acceleration crit.eria is not met. Visual inspection was performed by the QPI and QPE. Visual vibration for BOP pipe is evaluated based on visual acuity of 2-3 mils observable vibration l per foot of distance from the pipe. Each section of pipe was analyzed for i allowable vibration level and the viewing distance was determined based on this acuity factor. During the test, if the QPI was unable to detect 1
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STARIUP TEST RESULTS - (CONT.)
Piping Tests (Cont.)
I motion of the pipe from the prescribed viewing distance, then the vibra-tion level was acceptable. For NSSS piping, vibration observed from any distance was unacceptable, and required additional instrumented inspec-tion.
Vibration Testing Results During the non-nuclear heatup performed during the heatup test plateau, two remote vibration tests were run on NSSS piping. Steady state vibra-I tion was recorded with the recirculation pumps on low speed at minimum FCV position and transient vibration was recorded during the recirculation pump cold start (shift from low to high speed at 125'-175'F). Analysis of results from both tests were satisfactory for both Level 1 and 2 criteria limits, demonstrating that vibration displacement and, therefore, pipe stresses were within the code limits for the condition specified.
During the first nuclear heatup to operating conditions, and concurrent with the visual inspection for thermal expansion of NSSS pipe, visual inspection for vibration was also performed. No vibration was observed and all results were satisfactory. Also during the first nuclear heatup, remote BOP steady state testing was performed on the feedwater supply piping to the vessel. Testing was performed at relatively low flow rates compared to operating conditions, however, this presented the first opportunity to evaluate performance of the system. Test results were satisfactory, with very low vibration levels.
Transient vibration for both NSSS and BOP piping was evaluated during the SRV functional test. Test exceptions were taken on NSSS testing due to ERIS data retrieval problems. One SRV strain gage was declared inoperable and data for one SRV was lost in the archiving process due to a problem in the ERIS software. GE analyzed the effect of the missing data, and l
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STARIUP TEST RESULTS - (CONT. )
I Piping Tests (Cont.)
determined that based on the available data and required performance of this test again at rated pressure, satisfactory performance was demon-strated. BOP piping results were also impacted by data acquisition problems due to the failure of two accelerometers. Engineering analysis demonstrated satisfactory performance based on available data. A single Level 2 violation of acceptance criteria was analyzed as acceptable. SZT testing demonstrated satisfactory performance during vibration testing for both NSSS and BOP piping.
Steady state and transient vibration data for NSSS and BOP piping was evaluated during the RCIC Hot Quick Start and subsequent two hour run.
l During the quick start BOP transient vibration was monitored on selected portions of piping. Exception was again taken on two inoperative accelerometers discovered during SRV tests, however, sufficient data was available to demonstrate satisfactory performance. During the two hour run steady state vibration was remotely monitored on NSSS pipe and both remote and visual evaluation was performed on selected portions of BOP piping. All results were satisfactory.
Overall, vibration testing to date has demonstrated satisfactory response l to steady state and transient operating conditions. In fact, only a single criteria violation was identified for failure to meet Level 2 acceptance criterion, which was analyzed by Engineering to be acceptable.
All other exceptions noted during vibration testing involved ERIS and the l data acquisition system and were not related to system vibration.
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STARTUP TEST RESULTS - (CCNr.)
Piping Tests (Cont.)
Chronological Outline Piping Vibration and Thermal Expansion Testing I
July 26, 1986 During the non-nuclear heatup, vibration testing was per-formed on NSSS piping (main steam and recirculation piping, GE scope): transient testing during the reactor recircula-tion pump cold start (shift from low to high speed),
reference TSN 028; steady state testing with reactor re-circulation pump on LmG at minimum flow control position, reference TSN 027. Both tests consisted of ERIS monitoring of remote vibration and strain instrumentation and both tests showed vibration and strain levels to be satisfac-tory, within the acceptable limits.
August 26, 1986 During the first nuclear heatup, a significant effort was put forth for thermal expansion testing of NSSS and selected BOP piping systems. Prior to this time, a series of walkdowns were performed by the Qualified Piping Inspector / Engineers (QPI/QPE) to ensure that adequate clearance was available to allow unobstructed expansion of the scoped piping. The QPI,tPE's were NED Engineering and design personnel trained and tested to program require-ments. During this prerequisite walkdown, all actual and potential interferences were identified for resolution, and cold readings were recorded for snubbers and spring cans to serve as a baseline for pipe growth. Additionally, the ERIS computer was initilized, a process which records the ambient temperature of all thermal expansion sensors to serve as a baseline during heatup. Remote sensors were monitored at 50'r intervals during the heatup from ambient to rated temperature using the " EXPAND" function of ERIS.
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STARTUP TEST RESULTS - (CONT.)
Piping Tests (Cont.)
Chronological Outline (Cont.)
August 26, 1986 (Cont.) At each interval the program was executed and a printout was provided for each sensor showing pipe position and temperature. W e program adjusted the Level 1 acceptance criteria for intermediate temperatures to verify proper pipe expansion. At the 250*-350'r interval, a visual inspection of the piping was made by the QPI to further ensure proper pipe response. At rated temperature, an EXPAND run as execJted to record final pipe position for comparison to criteria, and the visual inspection was repeated to record hot hanger positions and inspect for adequate clearance at operating position. Expansion I testing for NSSS pipe, TSN 033, was satisfactory, with expansion verified as acceptable based on Level 1 analysis and visual inspection results. Expansion testing for BOP pipe, TSN 035, identified several potential violations during the heatup. The plar.t was placed in a hold con-dition while the interferences were resolved. Several Level 1 violations were identified at rated conditions.
tese violations were analyzed and reworked as required, and identified as requiring retest during the next heatup.
A third test was performed as part of the rated temperature walkdown. At this time, a visual vibration inspection of l
the NSSS piping was performed, reference TSN 034. Results of this test were satisfactory. Several other tests were also performed during the course of heatup. These descriptions follow as separate entries.
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I STAR'IUP TEST RESULTS - (COtfr. )
Piping Tests (Cont.)
Chronological outline (Cont.)
I Sept. 30, 1986 Transient vibration / strain testing of NSSS piping during SRV functional test was performed, reference TSN 036. Test results were satisfactory, however, several problems with ERIS hardware and data acquisition were noted. Related transient vibration and thermal expansion tests were run on BOP pipe, reference TSN 037. Again, test results were i
satisfactory with the exception of ERIS problems.
October 8, 1986 Three tests were run in conjunction with the RCIC two-hour run. TSN 054 monitored transient vibration of selected BOP pipe during the RCIC pump quick start with satisfactory results. During stable operation, the same BOP piping was I monitored for steady state vibration and thermal expansion remotely by ERIS, as well as visually by the QPI, reference l TSN 055. Several thermal expansion problems were noted, corrected and designated for retest in the next thermal cycle. Vibration results were satisfactory. TSN 056 monitored steady state vibration of NSSS piping at the same time, also with satisfactory results.
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November 23, 1986 Two thermal expansion tests were performed during the 1
t second nuclear heatup. The first, a scheduled test -
TSN 083, was the second thermal expansion test of NSSS piping. The piping expanded per design as indicated by successful Level 1 analysis. The second test, TSN 084, was
- the BOP retest resulting from the Level 1 violations which occurred in TSN 035. The retests resulted in two Level 1 violations which were designated for retest.
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I STAR'IUP TEST RESULTS - (CCNT. )
I Piping Tests (Cont.)
Chronological Outline (Cont.)
I December 17, 1986 'No thermal expansion tests were again performed at this time during the third nuclear heatup. The first, a scheduled test - TSN 096, was the third thermal expansion test of NSSS piping. The successful completion of this cycle concluded the NSSS thermal test. TSN 097 was the DOP retest resulting from the Level 1 violations which were noted during the second nuclear heatup. Two violations I were noted during this retest and are presently with Engi-neering for completion of analysis. There is a restriction on the plant which limits the total number of thermal cycles to twenty pending final analysis.
Synopsis of Testing to Dates Vibration and thermal expansion testing performed to date has been, for the most part, successful. All vibration tests for NSSS and BOP piping have demonstrated that vibration levels are within the predicted Level 1 limits.
In fact, most of the tests have recorded extremely low vibration levels, approaching the resolution limits of the vibration sensor. This is expected at this point in the power ascension program, considering the low power levels at which testing was performed. The primary problems encountered during vibration
> testing have involved sensor operability and the ERIS data acquisition system.
l These problems were mostly " shakedown" problems which are not expected in future testing. Thermal expansion testing for the NSSS scope of pipe success-fully demonstrated the lack of obstruction and conformation to design limits of the subject piping. Comparison of the results of the three cycles also demon-strated good repeatability of pipe response to expansion. Thermal expansion testing of DOP piping did identify several problems with the plant piping and l
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I STARIUP TEST RESULTS - (CCNr. )
I Piping Tests (Cont.)
Chronological Outline (Cont.)
Synopsis (Cont.)
suspension system. On several occasions, plant heatup was actually suspended while engineering resolved violations. .resently
- two violations are pending final engineering resolution. These are due to insufficient movement of a main steam drain line in the steam tunnel and the head vent line in the drywell.
Based on engineering analysis the violations pose no restriction to plant operation until Test Condition 2.
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I STARTUP TEST RESULTS - (CONT.)
I 4.15 STI-N27-023A Feedwater Control System (TSN -111, 114, 116)
The purpose of this test was to demonstrate the proper operation of the Feedwater Startup Level Controller to control feedwater flow using either Turbine Driven Feed Pump A ('IDFP"A") 'IDFP"B", or the Motor Feed Pump I (MFP).
Acceptance Criteria Level 1
- 1. The transient response of any level control system-related variable to any test input must not diverge.
Level 1
- 1. Level control system-related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25. .
Level 3
- 1. 'Ihe dynamic response of each individual level or ficn sensor shall be as fast as possible. Band width must be at least 2.0 radians /second (faster than 0.5 second equivalent time constant), except for the steam flow sensors which must have bandwidth of at least 1.0 radian /
second (" aster than 1.0 second equivalent time constant).
l 2. Vessel level, feedwater flow, and steam flow sensors must be instal-led with sufficiently short lines and proper damping adjustment so that no resonances exist.
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STAIC'P WST RESULTS - (CONT. )
STI-N27-023A (CONr. )
Level 3 (cont.)
- 3. All auxiliary controls which have direct impact on reactor level and feedwater control (e.g., feedpump minimum recirculation flow valve control) should be functional, responsive, ond stable.
Section 8.3 of STI-N27-023A, Feedwater Control System was performed using each of the feedwater pumps. The testing involved initiating increasing /
decreasing 5% flow demand step changes (controller in MANUAL) followed by I six inch increasing / decreasing steps in reactor water level (controller in AtJIO) . Throughout, each maneuver, the system's responses was monitored and recorded. The data obtained during testing was analyzed to verify system stability and correct controller response.
The only problem identified during this testing was that for each pump, the feedwater recirculatien controller output did not control the feed-water recirc valve. his violated Level 3 cirteria number 3. A Test Exception was written for each pump. This Level 3 acceptance criteria will be verified again as the test will be run as an already scheduled TC-1 test.
I 4.16 STI-B21-025A MSIV Function Test (TSN-070) h is test verified the functional performance and determined the closure times for each Main Steam Isolation Valve (MSIV). A full individual closure of each MSIV was performed immediately following reactor heatup to rated pressure and temperature. Based on the times from the fastest MSIV the Margin to scram for the Simulated Thermal Power, Reactor Pressure, and i Neutron Flux were calculated. The Main Steam Line Flow Hargin to Scram was also calculated.
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STARWP EST RESULTS - (CONT. )
STI-B21-025A (CONT.)
'Ihe reactor pressure and temperature were at normal rated conditions.
Each valve was manually closed and then opened with a waiting period after each movement to enable the plant parameters to stabilize. Only one of the valves at a time went through the open/close cycle while all the remaining valves remained open. After all the valves were individually cycled the data taken by the ERIS-TRA computer was saved and analyzed for the stroke time and the closure time.
Acceptance Criteria Level 1
- 1. The individual stroke times shall be greater than or equal to 2.5 seconds and less than or equal to 5.0 seconda. The actual values I for the inboard MSIV's,1B21-r022A-D, were 4.02, 3.48, 2.93, and 3.29 seconds respectively. The actual values for the outboard MSIV's, 1B21-r028A-D, were 3.78, 3.30, 3.35, and 3.31 seconds respectively.
All MSIV's met the criteria.
- 2. The total effective closure time for each 'MSIV shall be not greater than 5.5 seconds. The total effective MSIV closure time is the total time from when the solenoid is de-energized to when the valve is 100%
closed including any delay time caused by the instrumentation. The l actual values for the inboard MSIV's, 1821-r022A-D, were 4.44, 3.96, 3.41, and 3.74 seconds respectively. The actual values for the g outboard MSIV's, 1821-r028A-D, were 4.39, 3.84, 3.97, and 3.82 seconds respectively. All MSIV's met the criteria.
- 3. The average MSIV stroke time calculated from the fastest stroke times
, for either the inboard or outboard MSIV's for each line shall be not i
I less than 3.0 seconds. The actual value was 3.33 seconds. The MSIV's met the criteria.
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STARTUP TEST RESULTS - (CONT.)
STI-B21-025A (COm'.)
Level 2 I 1. The reactor did not scram or isolate during any individual MSIV closure.
- 2. The Margin to Scram for the Simulated Thermal Power for channels A and B shall be greater than or equal to 5.0%. The actual data was 60.333% for both channels.
- 3. The Reactor Pressure Margin to Scram shall be greater than or equal to 10 psi. The actual value was 110.2 psi.
- 4. The Transient Neutron Flux Margin to Scram shall be greater than or equal to 7.5%. The actual value was 11.3%.
- 5. The Main Steam Line Flow Margin to Isolation shall be greater than or equal to 10%. The actual value was 132%.
All Acceptance Criteria were met, and all components were in conformance with design predictions and as such were considered adequate for their functions.
During the closure of MSIV's IB21-F022B, D and IB21-F0288, D an annunciator (RPS MSIV Closure) activated. According to the applicable Alarm Response Instruction, this annunciator should have activated when at i least 2 MSIV's were less than 90% open. This condition did not exist during the test as each individual MSIV was closed then reopened prior to closing the next MSIV. Further investigation was performed and no problems with the annunciator, relay, or limit switches were found, and no further action is necessary.
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STAR'IUP TEST RESULTS - (CONT. )
4.17 STI-B21-026 SRV Functional Test The purpose of this test was to verify proper operation of each (19) safety relief valve (SRV) at low reactor pressure (250 psig), and to demonstrate pressure control stability during single SRV operation.
Acceptance Criteria With 1-1/2 bypass valves open at test start.
Level 1 Positive steam flow was demonstrated by reviewing plots of bypass valve position vs. time during manual SRV actuation.
Level 2
- 1. All SRV actuations had bypass valve position changes well within 10%
of the average bypass valve position change.
- 2. No oscillations were observed for pressure control related parameters (Bypass valve position, Reactor pressure, Reactor level).
- 3. All SRV's were verified to reseat, by observing that the final relief valve tailpipe temperatures returned to within 10 degrees Fahrenheit l of their initial value.
l One test exception report (TER) was written during the performance of this test. Initially only one bypass valve at 84% was open. This proved to be inadequate because when the first SRV was opened (F051D), bypass valve number 1 went from it's initial position of 84% to full closed. The i
l l
126
STAR'IUP TEST RESULTS - (CONT. )
STI-B21-026 (Carr. )
problem was corrected by having the number 1 bypass valve fully open and the number 2 valve 50% open before SRV lift. This provided enough steam to maintain reactor pressure constant when the SRV was lifted (bypass valves closed but not fully). SRV F051D was then successfully tested.
4.18 STI-B33-029A Recirculation Flow Control - Valve Position Loop (TSN-015)
The purpose of this test was to demonstrate the proper operation of the valve position control locps of the Recirculation Flow Control Sy. stem while in the loop flow manual mode. To acccmplish this, the flow control valves were stroked fully open and then fully closed in both slow and fast speeds and the valve motion is verified to be smooth and that there are no instabilities. Both recirculation flow control valves were stable.
Acceptance Criteria There are no Acceptance Criteria for this test.
i (INTENTICRUu1Y BLANK) l 127
STARTUP TEST RESIETS - (C0tfr. )
STI-B33-29A (CONT.)
Results I Recirculation flow control valves "A" & "B" were cycled in both slow and fast speed. Control system operation and flow control valve motion were stable and no observable instabilities existed.
Discussion Initial stroking of recirculation flow control valve "A" resulted in a failure of the valve to stroke past 26% in slow spied, and a failure of the HPU when FCV "A" was stroked to 0% open when the system was restored.
This problem was documented in Test Exception 015-1. In trouble-shooting the problem, it was discovered that the circulating water pump (< 3 cire.
pumps)/ low vacuum runback seals in. If not reset, opening of the valve is limited. Stoi 029A-1-3 was written to assure that the interlock is reset.
An HPU "A" trip was due to high servo error encountered when rCV "A" reached the 0% open position during the close cycle. This problem was corrected via a control system tuneup. Flow control Valves "A" and "B" were successfully stroked on June 4, 1986.
Other problems discovered during this test were:
- 1. Missing wires (NRI PPDN 0371 and NRt PPDN 0370).
I 2. Failed relay (repaired via Work order).
I 128
STARRIP TEST RESULTS - (CCNr. )
STI-833-29A (CNr. )
It was noted that some electrical noise existed in the control system. At this time, the noise does not affect the test results or appear to prevent proper operation of the system. I&C was informed and examined the system in an attempt to reduce the noise. More testing is planned later in the power ascension test program and the effect of this noise on operation of the system will be closely monitored during this testing.
4.19 STI-r41-034 Reactor Internals Vibration (TSN-030)
The Pre-Nuclear lleatup Flow Test was conducted to verify the adequacy of the reactor internals design with respect to flow induced vibration.
Preliminary results indicate that all acceptance criteria were satisfied at this test condition. General Electric will provide Perry the final results in a later report.
Acceptance Criteria Level 1 The peak stress intensity may exceed 10,000 psi (single amplitude) when the component is deformed in a nunner corresponding to one of its normal or natural modes, but the fatigue usage factor must not exceed 1.0.
I Level 2 The peak stress intensity shall not exceed 10,000 psi (single amplitude) when the component is deformed in a manner corresponding to one of its normal or natural modes. This is the low stress limit Aich is suitable for the design life of the reactor ccmponents.
i l
129 l
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I STARRIP HST RESULTS - (C W r.)
I STI-r41-034 (cmr.)
A nitrogen overpressure of greater then 100 psi was maintained in the reactor vessel to provide adequate suction head for running the reactor recirculation pumps in fast speed during testing. The recirculation pumps were then run at high speed to achieve the required temperature range for testing (480-520*r). Vibration amplitudes and frequencies were monitored at various plant configurations (transient and steady-state) using special sensors mounted on the reactor internals. Sensors were located at the following locations.
I 1. Top of the shroud head, lateral acceleration (displacement).
- 2. Top of shroud, lateral displacement.
- 3. Jet pump riser braces, bending and extension strains.
- 4. Jet pump diffuser, bending strain.
- 5. Control rod drive guide tubes, bending strain.
- 6. Incore housings, bending strain.
- 7. Core spray sparger piping, bending strain.
No vibration data was obtained on the control rod guide tubes due to sensor loss. An analysis of control tod guide tube vibration from other DWR-6 plants and the Perry riow Induced vibration preoperational test I performed gave reasonable assurance that no significant vibrational modes would exist on the guide tubes.
130
I STARWP TEST RESULTS - (COffr. )
STI-r41-034 (cottr. )
Reactor cooldown was implemented twice during non-nuclear heatup. He first cooldown was performed due to drywell airlock problems. After a successful cooldown, the airlock was repaired and heatup comenced. The second cooldown was performed due to recirculation system problems. After the cooldown and subsequent troubleshooting and repairs, heatup comenced and testing was completed.
During testing a discrepancy in core flow readings between the double tap jet pump instruments, single tap jet pump instruments, and ERIS read-out from the double top jet pumps was discovered. This was found to be due to a re-spanning performed on the double tap transmitters with no subsequent span change made to the single tap nor calibration made to the ERIS double tap indications. The problem has been corrected by re-spanning the single tap transmitters and calibrating ERIS.
With the recirculation pumps in fast speed and at high core flows, noise was detected on IM channels A and E. The noise on I M A was absent upon its withdrawal, verifying the noise was vibration induced. This is not considered a problem since it is a unique ccndition established for this test, i.e., high core flows with the IM's fully inserted. Im E was not withdrawn due to interference under the vessel.
During this test the Recirculation System preoperational test and surveillance test for measuring the are suppression time for the recir-culation pump RPT breakers were also completed satisfactorily.
I ll 131
STARTUP TEST RESULTS - (CCNr. )
STI-F41-034 (C N r.)
Summary of GE's Preliminary Analysis (Attachment 1)
From preliminary data analysis vibrational modes (frequency and amplitude)
I for all flow and transient conditions encountered were well within acceptance critoria limits. The following table lists the maximum stress amplitude for each class of component where data was obtained.
Recirc Pump Flow Peak-to-Peak %
Component A B Amplitude Frequency Criteria Jet Pump Max tm 47.6 c 44.0 !!z 22.5%
Jet Pump Diffuser Max tE 12.0yc 29.5 !!z 14.3%
In Core
!!ousing Max Min 9.5 c 13.0 11: 59.4%
core Spray (6pc for all flows)
I Upper Shroud Max 50% 1.26 mil 15.8 Hz 66.0%
I I
132
I l
STARTUP TEST RESULTS - (CONT.)
STI-F41-034 (CCNr. )
The lower shroud wall and upper guide ring signals contained significant noise which precluded accurate preliminary data analysis. The control rod guide tube sensors were failed, and no signal was obtained. An analysis of CRD guide tube vibration data obtained from Grand Gulf, Kuosheng and Leibstadt, when compared with the CRD guide tube data obtained at Perry during preoperational testing, indicated that no significant vibrational modes would be expected on the guide tubes and that the loss of CRD guide tube data is acceptable.
Final test results will be presented to Perry within 120 days of the completion of the startup test program and will be based on a detailed analysis of vibration data taken during preoperational and startup testing.
4.20 STI-G33-070 Reactor Water Cleanup System (TSN -060, 063, 092)
The Reactor Water Cleanup System was operated in the Normal, Hot Standby, and BloWown modes to demonstrate specific aspects of system mechanical performance. Actual plant data obtained was compared to values from the I General Electric process diagrams. The following Acceptance Criteria were verified during the various sections of this test.
Acceptance Criteria Level 1 None 133
I I STArnJP TEST IU2TJLT3 - (CCtTr. )
Y STI-G33-070 (Carr.),
Level 2
- 1. With rated process flow in the normal mode, the non-regenetative hear' exchange: (ta!!X) tube side outlet temperature shall not ev.Nad !.20*r.
- 2. With rated process flow in the blowdown modo, the tcutx h+a side outlet temperature shall not exceed 130'r.
- 3. In the hot shutd:un mode, the available tiet Positive Sueticn !!eaii I (NPSil) shall nat be less than 13 feet.
l 4. The cooling water (Nuclear Close Cooling Mater or t1CC) supplied to the imIIX's shall not exceed 106% of cated flow. Rated cooling water flow corresponda to the flow rate necessary to operate the tm!!X at I capacity, assigning a maximum cooling water outlet temperature of 150*F. The tirJiX capacity is defined by the process flow diagram.
- 5. Recalibrate botters head flow indicator against IWCU flow indicator if the deviction is greater than 25 gpm.
- 6. Pump vibratico shall not exceut 2 mils peak-to-peak men:.ured in any I direction. l I
I 134
l I
l STARTUP TEST BESULTS - (CCNr.)
I l STI-G33-070 (CCNr.)
Section 8.1 - RWCU NORMAL MODE I
This section was performed at rated reactor pressure and temperature. l Rated process flow through the A and B filter demineralizers was established and NCC temperature at the inlet to the NRHX was recorded.
Rated NCC flow corresponding to the heat exchanger capacity for this inlet temperature was then calculated from vendor supplied curves., 106% of this value was calculated. After system parameters stabilized the following data was taken. j 1
RNCU System Performance Data for the Normal Mode I Parameter Value Rx Pressure (psig) 952 Reactor Level (inches) 199 (on narrow range)
Blowdown Flow (gpm) 8 (F041, F046, F035 closed)
RWCU Flow In (gpm) 440 RWCU F/D A Flow (gpm) 158 (from local instrumentation)
RWCU F/D B Flow (gpm) 158 (from local instrumentation)
Reactor Drain Flow (gpm) 46 RWCU Return Temp (*F) 425 Hx Tube In Temp ('F) S10 NHX Tube In Temp (*F) 200 NHX Tube Out Temp ('F) 105 1
I NCC Fm NHX Flow (gpm)
NCC Fm NHX Temp ('F) 490 145 NCC Ex Out Temp ('F) 80*
- Taken by digital thermometer locally i.
135
I STARWP TEST RESULTS - (CNr. )
I '
t STI-o33-070 (Ctwr.)
1G33-C001A 1G33-C001B l Point l Peak to Peak Vibration (Mils)l l Point l Peak to Peak Vibration (Mils)l
- i 1 l 0.15 mils ll 1 1 0.13 mils I I 2 l 0.1 mils ll 2 1 0.is mils l
, l 3 l 0.2 mils ll 2 l 0.23 mils l l 4 l 0.2 mils ll 2 ! 0.27 mils l The results of this data showed Section 8.1 successfully satisfied acceptance criteria one, four, and six.
Section 8.2 - RWCU Bottom Head Drain Flow Indication This section was performed to verify proper operation of bottom head drain flow indicator calibration. The filter demineralizers were taken out' of service and system suction flow was established through the bottom head drain path exclusively. Data from system flow indicator G33-R609 and bottom head drain flow indicator G33-R610 was then recorded at various system flowrates. The data showed good agreement with a maximum
! . difference of 13 gpm above the 130 gpm range. Data frem the test at the various measured flow points is as follows:
- I
'I (See following page)
I I
13.
g
'I STNtIUP TEST RESULTS - (CCNr. )
STI-G33-070 (CONT.)
Flow Indicator Data Points alue l
! arameter P Instrument l l l0gpm ! 80-90gpn!110-120gpm!130-140gpm!155+0gpm!
l l 1 -5 l I l l l lmKUInletFlow 1G33-R609l 0 ! 89 ! 117 I 135 I 153 !
l(gpm) 1H13-P680 l l l l l l lDrainFlow 1G33-R610l 12 l 91 l 114 l 128 l 140 l l< cpm)Im3-P680 l l l l l l I lBottomDrain lLineTemp('F) 1B21-R643l pt. 3 l 531 l
l 531 l
l 531 l
l 531 l
l 531 l
l l
l l1H13-P614 l l l l I lDrywellTemp.A 1D23-R210Al l car)in3-P601 l 145 l l
145 !
l 145 I I
145 !
I 145 I I
lSquareRootCon- 1s31-x602Al l l l g l l lverterInput l 1,og l 1,13 l 1.21 l 1.28 l 1.33 I lJacksElandSC l l l l l l I i(volts) im 3-P642 lSquareRootCon- 1r31-x602Bl l l l
l l
l l
l l
l l
l lverterInput lJacksElandSC 1
l 0.93 l l
1.06 l l
1.14 1 l
1.21 l l
1.27 I l
I l l l l(volts) Im3-P642 l l l lFlowIndicator 1G33-R610l l l l l l lInputVoltage l l l l l l l lSRu-1, terminals l13and15(Volts) l 1.26 l 3.21 l l l
l 3.73 l l
4.11 l 4.40 l
l l
l l l l l l llH13-P613 I
I I
137 I ;
I STARTUP TEST RESULTS - (CONT.)
STI-G33-070 (CONT.)
Section 8.3 - Blowdown Mode I This test was performed to demonstrate the adequacy of the RWCU NRHX during rated blowdown flow. The filter demineralizers were removed from I service, the regenerative heat exchanger was bypassed, and all flow was directed through the NRHX to the condenser. Data was then taken on heat exchanger performance. NCC temperature at the outlet of the NRHX exceeded design limits during the test (acceptance criteria 4) and a test exception report was generated to document this fact. Post test investigation of the exception revealed that the vendor curves that were being used to obtain rated NCC flow assumed the process (RWCU) flow to be at a lower temperature than the actual conditions at the time of the test. The curves were revised to envelope the expected range of RWCU temperatures that could be encountered during the test and the test will be reperformed. When successfully completed this test section will satisfy acceptance criteria two and four.
. 4.21 STI-N64-074 Off-Gas System (TSN -052, 079, 106)
The purpose of the off-Gas System test is to demonstrate proper system operation by taking data from process instrumentation and analyzing gas samples taken at various points in the system.
Acceptance criteria Level 1
- 1. The release of radioactive gaseous and particulate effluents must not exceed the limits specified in the Perry Technical Specifications.
(8.1) 138
STAR 7UP TEST RESULTS - (CONT.)
l l
l l
STI-N64-074 (CONT.)
I 2. Flow of dilution steam to the noncondensing stage must not fall below 92% of the specified normal value when the steam jet air ejectors are operating. (8.1)
Level 2
! 1. The system flow, pressure, temperature, and dew point shall comply with the Process Data Sheets. (8.1) 1
- 2. The catalytic recombiner, the hydrogen analyzer, the desiccant dryers, the activated carbon beds, and the filters shall be working properly during operation, i.e., there shall be no gross malfunc-tioning of these components. (8.1) l Startup During Heatup TSN 052 was perfocned during TC-HU with the "A" train in service. Due to the lower power level at which this test was performed (~3.5%), the charcoal adsorbers were not in service. Off-Gas System flow was 60 SCm.
An effluent sample taken from the off-Gas vent pipe verified that plant releases were within the Technical Specification limits for noble gases,
! particulates and iodine. This satisfied a level 1 acceptance criteria.
Data recorded from process flow meters in the steam flow line to steam jet air ejector "A" verified that adequate dilution flow was being supplied to the air ejector. This satisfied the other level 1 acceptance criteria.
I 139 l
1
I STARTUP TEST RESULTS - (CONT.)
STI-N64-074 (CONT.)
Ten process parameters were found to be outside of their level 2 acceptance criteria band. General Electric evaluated the out of range parameters and stated that the difference between the expected and measured values is due to minor differences between the standard GE design I and the "As-Built" configuration at Perry. GE also stated that the system operation is acceptable as is. Additional data will be analyzed when this test is reperformed at higher power levels.
Normal Operations During Heatup TSN 079 was performed during TC-HU with the "B" train in service. Due to the low power level at which this test was performed (-4%), the charcoal adsorbers were not in service. Off-Gas Syst m flow was 62 SCFM.
I An effluent sample taken from the Off-Gas vent pipe verified that plant releases were within the Technical Specification limits for noble gases, I particulates and iodine. This satisfied a level 1 acceptance criteria.
Data recorded from process flow meters in the steam flow line to steam jet air ejector "B" verified that adequate dilution flow was being supplied to the air ejector. This satisfied the other level 1 acceptance criteria.
l Eleven process parameters were found to be outside of their level 2 acceptance criteria band. General Electric evaluated the out of range parameters and stated that the difference between the expected and measured values is due to minor differences between the standard GE design and the "As-Built" configuration at Perry. GE also stated that the system operation is acceptable as is. Additional data will be analyzed when this test is reperformed at higher power levels.
140
STARTUP TEST RESULTS - (CONT.)
STI-N64-074 (COm'. )
TSN 106 was performed during TC-1 with the "A" train in service. The charcoal adsorbers were in service with reactor power approximately 6%.
Off-Gas System flow was 55 SCFM.
An effluent sample taken from the Off-Gas vent pipe verified that plant releases were within the Technical Specification limits for noble gases, particulates and iodine. This satisfied a level 1 acceptance criteria.
(
l Data recorded from process flow meters in the steam flow line to steam jet air ejector "A" verified that adequate dilution flow was being supplied to the air ejector. This satisfied the other level 1 acceptance criteria.
Ten process parameters were found to be outside of their level 2 acceptance criteria band. General Electric evaluated the out of range parameters and stated that the difference between the expected and j measured values is due to minor differences between the standard GE design l
and the "As-Built" configuration at Perry. GE also stated that the system operation is acceptable as is. Additional data will be analyzed when this test is reperformed at higher power levels.
Noble gas samples were taken immediately upstream and downstream of the charcoal adsorbers in order to obtain data for calculating the residence (holdup) time of the adsorbers. However, noble gas activity was too low at this power level to obtain the necessary data. Residence time calcula-l tions will be performed when reactor power is high enough to obtain adequate data.
l I
I 141 I
I I STARTUP TEST RESULTS - (CONr.)
4.22 STI-C95-099 ERIS (Emergency Response and Information System)
The ERIS testing conducted in the startup test program thus far has consisted of single rod scram timing verification and ERIS SPDS screen verification. The scram timing feature of ERIS is run on the transient recording analysis computer (TRA) and the SPDS screens are run on the real time analysis for display computer (RTAD).
I This test was performed in Test Condition Open vessel and Test condition Heatup. There are no level 1 Acceptance Criteria associ-ated with this test and the applicabla level 2 Acceptance criterion reads as follows:
I The control rod scram timing function shall indicate scram times of selected rods to the appropriate notch positions to within 0.01 second of an independent measurement.
The ERIS scram timing program measures scram times to the Technical Specification notch positions 43, 29, and 13. The ERIS times were compared to a time events analyzer recorder for the validation I process. The time events analyzer was connected to 1H13-P610 to monitor the scram initiation signal and the position of the selected rod being tested, and also served as the timing analysis for Section 8.3 of STI-C11-005 (Single Rod Scram Testing).
I I
I 142 I
STARTUP TEST RESULTS - (CONT.)
1
. STI-C95-099 (CCNT. )
1I 1.
l (Cont.)
All control rods were individually scram tested in Test Condition open vessel and all scram times satisfied the level 2 criterion. All I control rods were again tested at approximately 1050 psig and again satisfied the level 2 criterion.
I ERIS can now be used as a measuring tool to time individual control rod scrams to satisfy Technical Specification requirements.
- 2. Real Time Analysis for Display (RTAD) SPDS Screen verification.
- a. Plant Specific Constant Recalculation
'I Reactor Vessel Temperature Constants at TC-1 was performed only l
to record selected plant parameters for use in future ERIS tests. The fraction of rated core flow, bottom drain line l temperature, and average recirculation loop temperature data were recorded. No criterion was associated with this test.
- b. Event Target Verification Sections of this test are performed to verify that selected ERIS event markers correctly reflect actual plant conditions.
I I
143 I
I I STARTUP TEST RESULTS - (CCNr.)
I STI-C95-099 (CCNr.)
I 2. (Cont.)
I b. (Cont.)
l Safety Relief Valve Test was performed in TC-HU in conjunction with STI-B21-026 Section 8.1, SRV Functional Test. As selected SRV's were cycled, the BASIC ERIS displays which contained SRV event markers were monitored to determine that each display I reflected the correct SRV status. There are no level 1 accept-ance criteria associated with this test. 2e applicable level 2 criterion reads as follows:
Selected BASIC ERIS event targets (safety relief valve, MSIV, scram) shall agree with actual plant status.
Fourteen ERIS screens were monitored per this instruction during the SRV cycles. W e SRV event marker on each screen responded correctly as an SRV was cycled and returned to indicate "O SRV's
- OPEN" after the SRV was closed. All acceptance criteria was satisfied.
I 3. Validated Parameter Verification his portion of ERIS testing compares the calculated ERIS validated plant parameters with measured plant data. This comparison verified that the processors algorithms, plant specific constants, composed and measured point data base have been correctly setup / installed.
Sere are n level criteria ass ciated with this testing. The
'E applicable level 2 criteria read as follows:
E I
I 144 I
I STAR'IUP TEST RESULTS - (CONT. )
STI-C95-099 (CONT.)
- 3. (Cont.)
All ERIS validated data will agree with actual plant data within ! 3% (of rated).
- All ERIS validated data on the various BASIC ERIS displays (taken as near sintitaneously as possible) will agree with each other within a 2 sigma deviation.
Twelve ERIS screen verification tests were performed in test condition open vessel, heatup and one. The following tests were performed:
Drywell Temperature - Cold
- Reactor Pressure - Low
- RPV Press /2D Plot / Limit Tag Reactor Pressure Rated Reactor Power - Low Containment Temperature - Normal Drywell Pressure j Containment Pressure Drywell Temperature Normal Reactor Level - Limit Tags Suppression Pool Temperature Suppression Pool Level i
- Performed more than once.
I I
145
STAR'IUP TEST RESULTS - (CONT. )
l l
STI-C95-099 (CONT.)
i
- 3. (Cont.)
l The above tests verified that the ERIS displays were reading plant parameters that were in agreement with non-ERIS measured plant data.
l l
'Ihe testing also verified that certain ERIS validated parameters were being displayed correctly and censistent with the specific screens that called for them. A number of static and dynamic limit tags were l
verified to read correctly based on the plant parameters at the time l of the testing. Upon review of all the screen testing thus far, all acceptance criteria was satisfied.
1 4.23 STI-P44-116 Turbine Building Closed Cooling System l
(TSN -031, 121)
The purpose of this test is to demonstrate that the Turbine Building Closed Cooling System (TBCC) can provide a sufficient amount of cooling water to the heat loads it supplies to maintain design temperatures.
i I The plant will operate at various power levels with various heat loads on TBCC. During Heatup (HU), Test Condition 1 (TC-1) and Test Condition 6 (TC-6), data on components serviced by TBCCN will be gathered.
Acceptance Criteria Level 1 None l
I 146
l l
l I
STARIUP TEST RESULTS - (CONT.)
l I STI-P44-116 (CONT.)
Level 2 I The Turbine Building Closed Cooling System is capable of providing cooling water to the following heat exchangers and coolers to maintain system parameters within the temperature limits given below:
System Parameter Temperature Limits
- 1. Off Gas Glycol Coolers, Glycol exit solution 35 2*F temperature j 2. Reactor Feedwater Pump Turbine Oil Coolers, f 130*F l lube oil temperature
- 4. Turbine Control System Electro-Hydraulic Fluid 110 - 125"F Cooler, hydraulic fluid reservoir temperature
- 5. Steam Bypass System Hydraulic Power Unit Coolers, 90 - 130*F hydraulic fluid reservoir temperature
- 6. Mechanical Vacuum Pump Heat Exchangers Seal Water f 110*F l temperature for Mechanical Vacuum Pumps
- 7. Generator Station Coolers, Generator Stator Winding i 113*F Cooling Water System inlet temperature 147
STARIUP TEST RESULTS - (CONT.)
STI-P44-116 (CONT.)
Level 2 (Cont.)
System Parameter Temperature Limits
- 8. Exciter Duplex Cooler, Alternator Air Cooler 86 - 115'r discharge temperature l
l 9. Generator Hydrogen Cooler ventilating hydrogen 86 - 115'r gas temperature
- 10. TBCC temperature at Turbine Building Closed Cooling 90 - 100*r Heat Exchangers Outlet
- 11. Turbine Building Closed Cooling Heat Exchangers A, f 15'r B shell side differential temperature The Turbine Building Closed Cooling System STI has been broken into three
- sections.
l l 1. TBCC System - Hot Test (TC-6)
- 2. 1BCC System - Mechanical Vacuum Pump Data (TC-HU)
- 3. TBCC System - Motor Driven reed Pump Data (TC-1)
TBCC System - Hot Test has not yet been performed due .to requiring l
95 - 100% power.
I I
148
I 1
STARTUP TEST RESULTS - (CONT.)
STI-P44-116 (CCNT. )
TBCC System - Mechanical vacuum Pump Data was performed on 7/31/86 during Test Condition Heatup. The purpose of this test was to verify that operating temperatures of the seal water for the mechanical vacuum pumps 1N62-C001A and 1N62-C001B were below the required temperature limit. With the mechanical vacuum pumps running, seal water was verified to be within acceptance criteria limits.
TBCC System - Motor Driven Feed Pump Data was performed in TC-1 on l 1-24-87. The purpose of this test was to verify that the TBCC System maintained the motor driven feed pump lube oil temperature between 110'F and 130'F with either lube oil cooler in service.
The TBCC system - Motor Driven Feed Pump test was performed and failed level two criteria for two reasons. The System operating Instruction (SOI) initially set up the system properly but did not address throttling cooling water at the lube oil cooler when changing coolers. The second l
problen was the rounds sheet listed lube oil to be > 70'F which also was not within the 110 - 130*F Acceptance Criteria.
! The reselution to the above problem was to revise the SOI and the rounds sheet and re. peat the test after those corrections have been completed.
l l
149
I I STARIUP TEST RESULTS - (CONT. )
4.24 STI-G42-119 SPCU Performance Test (TSN-013) he suppression Pool Cleanup (SPCU) System demonstrated its capability to maintain Suppression Pool water chtmistry within specified limits.
Acceptance criteria Level 1 None Level 2 I 1. The SPCU demineralizer influent and effluent were within the following limits.
Influent Effluent
- a. Conductivity at 25'c f 10 pmho/cm f 3.0 mho/cm
- b. Chlorides 5 0.5 ppm f 0.1 ppm
- c. pH at 25'c 5.3 to 8.6 6.5 to 7.5
- d. Suspended Solids f 5.0 ppm f 1.0 ppm
- 2. Radiation levels on the 599'9" elevation of the reactor building were lI l
less then 2.5 mR/hr.
The test was performed on 5/9/86 by placing the SPCU system in service and drawing samples from both the demineralizer influent and effluent. The
- first run failed the Level 2 Acceptance Criteria for effluent pH (5.9).
l
< The demineralizer resins were replaced and the test was reperformed on 5/17/86 satisfying all water chemistry Acceptance criteria.
The Health Physics surveys were performed during the initial performance of the test on 5/9/86. All points were less then 0.2 mR/hr.
150
I STAlYIUP TEST RESULTS - (CONT.)
I 4.25 STI-T23-123 I
Concrete Temperature Survey at Low Power (TSN-062)
The purpose of this test was to demonstrate the ability of natural heat transfer to cool the concrete surrounding selected pipe penetrations.
Acceptance Criteria Level 1
- 1. During the initial nuclear heatup exterior Drywell and Shield Building concrete temperatures at selected penetrations shall not exceed a temperature of 200*F minus loop accuracy.
- 2. With one of the Auxiliary Boilers operating at rated temperature and pressure and supplying steam to an operating Radwaste Evaporator, the concrete temperature surrounding the Auxiliary Steam line penetration through the Steam Tunnel roof shall not exceed a temperature of 200F minus loop accuracy for a minimum six hour period.
Due to the high radiation fields present in the Steam Tunnel, remote monitoring was performed by attaching thermocouples to the concrete and routing the thermocouple wire to low radiation areas.
Concrete temperature data was collected over a six hour period and included concrete temperatures near main steam, feedwater, reactor water cleanup, and auxiliary steam penetrations. The maximum allowable concrete temperature was determined to be 195'F based on a 5'F maximum thermocouple ,
loop-digital thermometer error. The following table lists the maximum
~
temperatures recorded:
I 151
I
- I STAR'IUP TEST RESULTS - (CONr. )
lI l
l STI-T23-123 (CONr. )
l Maximum Recorded Temperature ("F)
Location Drywell Shield Bldg I Main Steam 118 146
( Feedwater 108 107 Reactor Water Cleanup 111 121 Auxiliary Steam N/A 124 l
The maximum recorded temperature was 49'F below the Level 1 analysis limit, thus satisfying all criteria.
4.26 STI-N33-129 Steam Seal (TSN -029, 047)
During the Heatup Plateau, the Steam Seal System flow path of non-nuclear steam to the main turbine stop, control and combined intercept valves was verified by measuring piping temperature at the respective gland seals.
Additionally, operating performance data for the Steam Seal System was l obtained utilizing main steam at 600 psig as its source of heat addition.
Finally, the operability of the following regulating valves was successfully demonstrated by cycling a remote, manual valve in parallel with each regulating valve and observing the regulating valve compensate accordingly:
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v I STARTUP TEST RESULTS - (CCNT. )
I STI-N33-129 (CCNT. )
IN33-PCV-F025 Main Steam feed to Steam Seal Evaporator Valve (Steam Seal evaporator pressure regulation)
IN33-PCV-Fil5 Steam Seal Evaporator Water feed valve (shell side Level Control)
IN33-PCV-F070 Steam Seal feed regulating valve (Steam Seal supply header pressure Regulation)
Acceptance Criteria Level 1 1 .
None ;
Level 2 l
None I- Level 3 I 1. Steam seal system supply lines to each gland seal were verified as being open as demonstrated by visual observation of steam from the glands or by verifying the supply line is a minimum of 50'F above ambient ten:perature.
I 2. The Steam Seal Evaporator level is maintained at 653'-3/8" 1" when supplied heating steam from the Main Steam System and the Extraction Steam System. (As measured by ERIS point N33EA001 this translates to 1 inch)
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I STARTUP TEST RESULTS - (CONT.)
STI-N33-129 (CCNT. )
There was no visual steam emitted from any of the gland seals. Conse-quently, verification of steam flow path required measuring the surface l temperatures of all supply lines to the gland seals. Table 1 below shows l that the recorded temperatures for all gland seals satisfy the Level 3 Acceptance Criteria by being greater than 50'F above ambient.
Table 1 - Main Turbine Valve Gland Seal Temperatures Ambient Steam Supply Temperature Line Temperature Description 'F 'F l
Main Turbine Stop Valves 1N11-F200A 82 > 180 (1 seal per valve) 1N11-F200B 82 > 180 1N11-F200C 82 > 180 IN11-F200D 82 > 180 Main Turbine Control valves 1N11-F250A 86 > 180 (1 seal per valve) 1N11-F250B 84 > 180 1N11-F250C 77 > 172 l 1N11-F250D 82 > 180 Main Turbine Combined 1N11-F300/350A 85 > 140*
Intercept Valves 1N11-F300/350B 85 > 140*
(2 seals per valve) 1N11-F300/350C 85 > 140*
1N11-F300/350D 85 > 140*
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- Temperature at both 1N11-F300/350E 85 > 150*
seals 1N11-F300/350F 85 > 150*
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STARTUP TEST RESULTS - (CONT.)
I STI-N33-129 (CONr.)
Section 8.1 & 8.4 System Performance data at 600 psig Reactor pressure were collected with the Steam Seal Evaporators using steam from the Main Steam lines. This data is presented in Table 2 below. The Level 3 Acceptance Criterion for evaporator level was easily satisfied.
The only deficiency identified during the test was related to the sluggish response of the Evaporator level control valve, IN33-PLV-Fil5. TER-047-01 was initiated to document the condition. The proportional band for its controller was readjusted and the affected steps were reperformed successfully resulting in adequate level control.
Table 2 - Steam Seal Evaporator (SSE) Performance Data (Main Steam Supply)
Parameter Instrument Data I
l SSE Level N33EA001 -0.19 inches SSE Heating Steam Pressure IN33-R052 28 psig l SSE Shell Pressure IN33-R057 13 psig Steam Seal Supply Header Pressure IN33-R083 3.9 psig SSE Heating Steam Temperature IN33-R195 300'F
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E STAR 1UP TEST RESULTS - (CChT.)
Condenser Air Removal System I 4.27 STI-N62-130 (TSN-051)
This test demonstrated the proper operation of the steam jet air ejectors (SJAE's) with respect to maintenance of a sufficient vacuum in the main auxiliary condensers at 100% load, continuity of flow path from the SJAE to the off gas system, and correct operation of the low flow interlock.
The purpose of the low flow interlock is to prevent sending a cembustible mixture of hydrogen to the off gas system by shutting the suction valves on the main and auxiliary condensers when low SJAE steam flow is sensed.
I Acceptance Criteria Level 1 None Level 2
, 1. During the 100% (+0, -5) power phase of PNPP Unit 1 testing each SJAE, individually, must be able to maintain the Main and Auxiliary f condensers at an absolute pressure equal to or less than the vendor generated curves of expected condenser backpressure as a function of l heat load and condenser inlet temperature.
l 2. During the post fuel load testing, SJAE IN62-C002A or 1N62-C002B air suction valves 1N62-F140A (F140B) and 1N62-F170A (F1708) close and the alarm, SJAE A (or B) FI4W LCH CONDR ISOL, annunciates when the steam flow to the a stage ejector decreases to 9,000 lb/hr 180 lb/hr.
- 3. Level control valves of the SJAE must maintain the level of water in each Intercondenser Loop Seal between a low of 589'-11 3/8" and a high of 596'-3".
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I STARTUP TEST RESULTS - (COtTP. )
STI-N62-130 (CONT.)
- 4. During the post fuel load testing, the continuity of the flow path for the noncondensible gases leaving the SJAE N62-C002A or N62-C002B going to the off-Gas (N64) system is verified.
Section 8.1 - SJAE Low Flow Test With the reactor at rated conditions and the main and reactor feed pump turbines on their respective turning gears, the SJAE were started up per normal plant procedures. Pre and post startup off gas pre-heater pressure data was taken. Main steam flow to the SJAE was then decreased by throttling the main steam to SJAE supply valve until the low flow annunciator alarmed and the SJAE suction valves closed. The steam flow at which these actions occurred was recorded. The main steam valve was then opened and the suction valves were verified to open again. The test was then repeated for the other set of SJAE's. The mechanical vacuum pumps
, were run during the test to ensure condenser vacuum was maintained. On
! the original run of this test for SJAE train 'A' the setpoint for annunciation / isolation was low. A Test Exception Reoort was generated and after recalibration the test was rerun satisfactorily. Section 8.1 successfully satisfied acceptance criteria two and four.
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l l 4.28 STI-R63-133 Loose Parts Monitoring System (LPMS) Baseline Data (TSN -081, 113)
The Loose Parts Monitoring system (LPMS) was successfully used to collect steady stae baseline data during the Heatup Plateau (HU) and Test Condition one (TC 1).
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I STAR'IUP TEST RESULTS - (CONT. )
STI-R63-133 (CCNT. )
Acceptance Criteria Level 1 None Level 2 Initial Baseline data was obtained for steady state conditions established for the appropriate test conditions (HU or 'IC 2.).
For the Heatup Plateau, the reactor power was at 3% of rated with both recirculation pumps on slow speed and maximum flow control valve position.
I For 'IC 1, the reactor power was at 15% of rated with both recirculation pumps on slow speed with maximum flow control valve position. The generator was not synchronized to the grid.
l All twelve vibration and loose parts channels for the reactor were monitored for this test. 'Ihe baseline data was collected on three 90 minute cassette tapes lasting approximately 11 minutes each. All baseline data was collected satisfactorily with the noise plots showing the l expected characteristic waveform distribution. .
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