IR 05000354/1986038

Insp Rept 50-354/86-38 on 860811-22.No Violations Noted. Major Areas Inspected:Previous Insp Findings & Overall Power Ascension Test Program,Including Procedure Reviews & Qa/Qc Interfaces

ML20203N262
Person / Time
Site:	Hope Creek
Issue date:	09/09/1986
From:	Briggs L, Florek D, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20203N261	List: IR 05000354/1986038 ML20203N259
References
50-354-86-38, NUDOCS 8609230258
Download: ML20203N262 (15)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-354/86-38 Docket N License No. NPF-57 Licensee: Public Service Electric and Gas Company 80 Park Plaza Newark, New Jersey 07101 Facility-Name: Hope Creek Generating Station, Unit 1 Inspection At: .Hancocks Bridge, New Jersey InspectionConducyed: August 11-22, 1986 Inspectors: .

D. Flor'6k L ad R ctor Engineer

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L. Wink, Reactor Engineer W9lec

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L. Briggs,g(pf,TestProgramsSection, ~ dat'e OB, DRS Inspection Summary: Inspection of August 11-22, 1986 (Inspection Report No. 50-354/86-38)

Areas Inspected: Routine, unannounced inspection of previous inspection findings, overall power ascension test program including procedure reviews, test results evaluation and test witnessing, independent measurements and verifications, QA/QC interfaces and tours of the facilit Results: No violations were identifie NOTE: For acronyms not defined, refer to NUREG-0544 " Handbook of Acronyms and Initialisms."

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DETAILS 1.0 Persons Contacted Public Service Electric and Gas Company (PSE&G)

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-J. Adams, Power Ascension Technical Coordinator L. Alversa, Senior Nuclear Shift Supervisor G. Chew, Power Ascension Results Coordinator G. Connor, Operations Manager

• G. Daves, Senior Engineer, Operations e P.'Dempsey, Shift Test Coordinator

, *M. Farshon, Power Ascension Manager B. Forward, Power Ascension Administrative Coordinator

• A. Giardino,-Manager-Station Quality Assurance (QA)

• Griffith, Principal QA Engineer

• J. Hagan, Maintenance Manager D. Hosmer, Lead Shift Test Coordinator i

• P. Krishna, Assistant to the General Manager R. Salvesen, General Manager, Hope Creek Operations W. Schell, Power Ascension Technical Director E. Skeehan, GE Operations Manager W. Thomas, Shift Test Coordinator C. Vondra, Operating Engineer U.S. Nuclear Regulatory Commission D. Allsopp, Resident Inspector R. Borchardt, Senior Resident Inspector i * Denotes those present at the exit meeting on August 22, 1986.

The inspector also contacted other members'of the licensee's staff in-cluding senior nuclear shift supervisor, reactor operators, test engineers

and members of the technical-staff.

i 2.0 Licensee Action on Previous Inspection Findings

(0 pen) Unresolved Item (354/86-35-03).

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Temperature distribution in the reactor under-vessel area. The inspector discussed the status of this item with the Power Ascension Manager. The inspector was informed that, based on preliminary General Electric analy-sis, the under vessel area temperatures were considered to be acceptable i but that the temperature distribution of the vessel skirt was still under investigation. Additional data will be obtained following the next scheduled reactor shutdown. This item will remain open pending completion of the licensee's investigation and NRC review of the results.

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3.0 Power Ascension Test Program (PATP)

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3.1 References s

Regulat'ory Guide 1.68, Revisionc2, August 1978, " Initial Test Programs for Water-Cooled Nuclear Power Plants"

ANSI N18.7-1976, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants"

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Hope Creek Generating Station (HCGS).. Technical Specifications, Revision 0, April 11,1986 ,

HCGS Final Safety Analysis Rep, ort (ESAR), Chapter 14, " Initial Test Program"

HCGS Safety Evaluation Report (SER), Chapter 14, " Initial Test Program" Station Administrative Procedure, SA-AP.ZZ-036, Revision 3,

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" Phase III Startup Test Program"

" Specification NEB 0 23A4137, Revision 0, " Hope Creek Startup Test s Specification" '\ ,

HCGS Power Ascension Test Matrix, Revision 7 3.2 Overall Power Ascension Test Program Discussion The4 inspector met with the'Fower Ascension Manager, Power Ascension Technical Director, Power Ascension Administrative Coordinator and other members of the Power, Ascension staff to determine the overall l

status of the test program. The licensee completed low power testing

(Test Condition Heatup) on August 5,1986 (see paragraph 3.5 for a discussion of the plateau review) and began Test Condition 1 (10%-20%

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l Power) on August 11, 1986. On August 22, 1986 the licensee conducted i TE-SU.SV-281, Shutdown from Outside the Control Room, (see paragraph 3.4 for a discussion of this test) and entered a planned outage.

l Following completion of the outage and the plateau / review of test i results from Test Condition 1, the testing program will resume with l testing in Test Condition 2 (25%-50% Power).

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During the course of the ' inspection the inspector observed various aspects of the administrative end technical staff support of the

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Power Ascension Program. This included reduction of test data, l analysis of test results, preparations for test performance, and l

planning and scheduling of test activitie Findings '\

No unacceptable conditions were identifie ~

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. 3 3.3 Power Ascension Test Procedure Review Scope The Power Ascension Test Procedures listed in Attachment A were re-viewed for their conformance to the requirements and guidelines of the references in paragraph 3.1 and for the attributes previously defined in Inspection Report 50-354/86-0 Discussion TE-SU.ZZ-004. The Test Plateau Matrix for Test Condition One was reviewed to verify that, 'in addition to the testing previously identified for this test condition, tests deferred from Test Condi-tion Heatup were include TE-SU.SV-28 In addition to the reference listed in pa'ragraph this test was also reviewed for conformance to Regulatory Guide 1.68.2, Revision 1, July 1978, " Initial Startup Test Pronram to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants."

Findings No unacceptable conditions were identifie .4 Power Ascension Test Witnessing Scope

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The inspector witnessed the performance of the power ascension tests listed in Attachment B and discussed below. The performance of these tests were witnessed to verify the attributes previously defined in Inspection Report 50-354/86-3 Discussion TE-SU.BD-144. This was the second of two required demonstrations of the ability of RCIC to automatically start and inject into the re-actor pressure vessel from a cold standby condition. At the time of the test the reactor was at approximately 16% of rated power and RCIC had been in a standby condition for more than the minimum required time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The automatic start sequence was begun by arming and depressing the initiation pushbutton. The inspector observed ,

overall system performance to be excellent and independently con-firmed that RCIC achieved rated flow to the reactor pressure vessel within the required 30 seconds. During the course of the test the inspector also observed operations personnel monitoring suppression pool temperature as required by technical specification '

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TE-SU.BJ-151. This was a reperformance of a test originally con-ducted during Test Condition Heatup. Following review of the ori-ginal test data the licensee elected to repeat this test. While the original data was sufficient to evaluate the acceptance criteria (see discussion of test results in paragraph 3.5), the licensee wished to improve on the quality of the data available for detailed analysis of system performance. The inspector observed a controlled start of the HPCI system at rated reactor pressure with a flow path established to the condensate storage tank (CST), followed by flow steps with the controller in both manual and automatic. Following these stability demonstrations the system was shutdown and an automatic quick start was performed with a flow path to the CST. The inspector observed adequate system performance and independently verified that rated

flow was achieved within the required 27 seconds. Operations per-sonnel were observed to be complying with technical specification requirements for monitoring suppression pool temperature during the tes '

TE-SU.AE-251. This was the second of two functional demonstrations of the Main Steam Isolation Valves (MSIVs). The test was performed at approximately 16% of rated power sith a reactor pressure of 930 psig. The test involved individual fast. closures of each of the eight MSIVs and verification that their closure times are less than

. 5 seconds and greater than 3 secondi. Following the initial testing

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of each MSIV it was discovered, during the on-the-spot data eval-

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'uation, that the correct. signals for the "D" MSIVs had not been

.i loaded into the transient recorder (GETARS). Immediate actions were initiated to load the correct signals and the "D" MSIVs were 3 re-stroked to obtain the required data. The inspector reviewed the y GETARS data and independently verif,ied that the longest closure time

was 3.97 seconds ("C" Outboard MSIV) and the shortest stroke time was 5,

3.01 seconds ("C"_ Inboard MSIV) thus satisfying the acceptance criteria. During_later discussions with the Power Ascension Tech-

, nical Director the inspector was informed that a recommendation was

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being made to adjust the stroke times of the two fastest MSIVs.to provide additional margin to the acceptance criterion limit. The licensee plans to make these adjustments during the outage following the completion of Test Condition '

TE-SU.SE-261. This functional test oft the main steam Relief Valves was performed at approximately 925 psig reactor pressure with five

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main steam bypass valves open and the main turbine off-line. During

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this test, data was also gathered on the response of-the main steam-

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lines, relief valve discharge piping, and acoustic nonf tor During the performance of the test the inspector monitored diverse' plant

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parameters to confirm expected response and verify positive indi-cation of steam flow throtgh each Relief Valve. All relief valves 3 performed as expected. However, three acoustic monitors (Fce ' Relief

% Valves "P", "L" and "R") failed to alarm as required. The senior 4;  ; nuclear shift supervisor en'tered the required technical specification

'a action statement (3.4.2.1.C) for inoperable acoustic monitors, which

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allows operations to continue for up to 7 day .

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.- 5 Subsequent evaluation of the GETARS data indicated that two of the acoustic monitors showed some response below the alarm level while the third did not show any indication. The inspector _will follow the resolution of this problem during.a subsequent inspection'. The inspector observed good overall coordination and performance by both operations and test personnel during these test TE-SU.SV-281. The performance of the hot shutdown portion of this test was witnessed by two regional inspectors and the resident in-spector. The purpose of the hot shutdown portion of this test was to demonstrate-the ability to shutdown the plant from outside the Main Control Room and maintain it in a stable hot standby conditio The initial conditions for the test were approximately 20% rated power with all systems in.their normal alignments. During the test, a backup crew was maintained in'the control room to monitor plant ;

status, protect balance of plant equipment and assume control in the j event of an emergency. The actions of this backup crew were moni-tored by an inspector to ensure the' validity of the tes The test was begun at 1622 with the announcement of a simulated evacuation of the Main Control Room. The test crew (minimum manning required by technical specifications) left the control room and pro-ceeded to take actions in accordance with operations procedure OP-IO.ZZ-008, Remote Shutdown. At 1624 power to the RPS busses.was i removed from the RPS distribution panels resulting in a reactor scram,L main steamline isolation and trip of the reactor recirculation pumps. -Plant response was as. expected and-the reactor was shutdown and isolated at a pressure _of approximately 935 psig and level slightly above the wide range indication (+60 inches wh'ich equals 221 ;

inches above the top of the active fuel). At 1626 the transfer

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switches were activated at the remote' shutdown panel and control of

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safe shutdown equipment was transferred. Between 1643.and 1708 positive control of reactor pressure was demonstrated at the remote i shutdown panel by lifting each of the three remotely controlled re-

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lief valves. At approximately 1730 the "A" Loop of RHR was placed l

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in suppression pool cooling (the licensee elected to demonstrate the L "A" Loop because it is the more difficult to align, requiring local l

! manual operation, and reserved the "B" Loop, which can be aligned from the remote shutdown panel, for use in shutdown cooling). .

Following the third relief valve lift the wide range level indicator '

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recovered on scale indication and RCIC was established to make up to the vessel at 150 GPM at 1748. After demonstrating positive control of reactor level the RCIC system was re-aligned in the CST to CST mode to provide pressure control. The hot shutdown demonstration was concluded at 1845 when control was returned to the Main control Roo During the performance of the test the inspectors and licensee per-

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sonnel assessed the adequacy of the remote shutdown procedure in order to identify difficulties and areas for improvement. One

difficulty arose when operators needed a key from the remote shutdown

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. 6 panel key cabinet to establish suppression pool cooling. The master key for the key cabinet had been left in the shift supervisors office in the Main Control Room. This required the test crew to force entry into the remote shutdown key cabinet. The licensee plans to address this problem. Another area of concern is that communication between personnel at the remote shutdown panel and field operators is via portable radios. Due to the possibility of RFI, personnel at the re-mote shutdown panel must leave the room to communicate with the field. During the test, operators retained visual contact with the remote shutdown panel by maintaining the security doors open (with security personnel on hand to control access). The licensee's com-plete evaluation of the test and plans for improvements will be re-viewed by the inspector during a subsequent inspectio Overall test performance was judged to be very good. A complete path to the ultimate heat sink was demonstrated as well as the ability to control reactor water level. The performance of the operations crew and test personnel was judged excellen Findings No unacceptable conditions were identifie .5 Power Ascension Test Results Evaluation Scope The power ascension test results listed in Attachment C were eval-uated for the attributes identified in Inspection Report 50'354/86-24. A summary of significant test results and identified test result deficiencies is provided in the discussion belo Discussion TE-SU.ZZ-01 Reactor water chemistry was within technical specification limits and is summarized below:

Parameter Measured Limit Conductivity (pMH0/cm) .18 5 Chlorides (ppb) <5 s100 pH .6- Four results deficiencies (RDFs) were identified for Condensate /Feedwater chemistry and were evaluated and accepted

"As-is" based on reactor water quality. A recommendation was made to clean the condenser hotwell at a convenient tim . . _ . _

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. 7 TE-SU.ZZ-021. All radiation measurements were well within acceptance criteria limits and consistent with current plant radiological zone TE-SU.ZZ-052. This test was performed twice (at 562 psig and at 806 psig reactor pressure) to measure the scram times of four selected control rods (Coordinates 26-07, 34-07, 34-23 and 34-47). Measured scram times were well within the acceptance criterion and technical specification limit of 7.0 seconds and_no unusual trends were ob-serve TE-SU.BF-055. At 955 psig reactor pressure scram time measurements were made for all control rods and friction measurements were made for the four selected control rods identified above. All acceptance criteria were satisfied. The average scram times for all rods were:

Notch Average Time (sec) Limit (sec)

45 .266 .43 39 .545 .86 25 1.238 1.93 05 2.267 3.49 TE-SU.BF-056. Following the scram time testing of all control rods discussed above, four new control rods were selected for monitoring during future planned scrams in the power ascension test progra These four control rods (Coordinates 26-23, 42-15, 50-47 and 54-19)

were then functi;nally checked and scram time tested in this test to provide baseline data for future comparisons. All acceptance criteria were satisfie TE-SU.SE-121. The APRMs were calibrated conservatively by means of the constant heatup rate method. The technical specification set-points for scrams and rod blocks were verified. All acceptance criteria were satisfie TE-SU.BD-141. This test was performed twice (at 156 psig and at 962 psig reactor pressure) to span the design operating range for RCI Time to rated flow at both conditions was well within the acceptance criterion limit of 30 seconds (18.2 seconds at low pressure and 1 seconds at high pressure). The RCIC turbine did not trip or isolate during these tests and the gland seal condenser system was demons-trated capable of preventing steam leakage. A level 2 results de-ficiency was identified during the high pressure test when the second turbine speed peak exceeded the limit of 4725 RPM (4800 RPM). The deficiency was evaluated and accepted "As-is" based on system per-formance during actual vessel injection.

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. 8 TE-SU.BD-142. At a reactor pressure of 963 psig, RCIC was tested for operability during actual vessel injection. All acceptance criteria were satisfied including a time to rated flow of 20.0 seconds (Limit 30 seconds) during a hot quick star TE-SU.BD-144. At a reactor pressure of 931 psig, the first of two demonstrations of RCIC performance from a cold standby condition was performed. All acceptance criteria were satisfied including a time to rated flow of 20.0 seconds (Limit 30 seconds).

TE-SU.BD-14 This test, at 160 psig reactor pressure, gathered baseline data for comparison during future routine surveillances of RCIC. All acceptance criteria were satisfie TE-SU.BJ-151. This initial test of the HPCI system at high reactor pressure (930 psig) was to demonstrate system operability and stab-flity. A results deficiency was identified when the decay ratios for various controlled parameters (control valve position, turbine speed and EGM output) failed to satisfy the quarter wave damping criterio In addition, while the time to rated flow (26.6 seconds) met the acceptance criterion of 27 seconds, General Electric recommended that the ramp generator signal convertor ramp time be decreased from 12 seconds to 10.5 seconds to add additional margin to the limit. The data recorded by the GETARS system also proved less than adequate for diagnosing the control system problem and the licensee decided to reperform the test in Test Condition 1. The retest was witnessed by the inspector (see discussion in paragraph 3.4).

TE-SU.ZZ-162. The assumed environmental conditions for the water level measurement calibration were verified to be acceptable. All acceptance criteria were satisfie TE-SU.ZZ-172. A walkdown of the nuclear steam supply system piping was performed twice, during heatup (moderator temperature of 251 F)

and at rated conditions (moderator temperature of 533 F) to verify that the main steam and reactor recirculation system piping was capable of free and unrestrained movement due to thermal expansio Two results deficiencies were identified at rated conditions. The

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first involved an inaccessible snubber (AB-030-H5), this was eval-uated and accepted "As-is" based on previous results, remote sensors-readings and acceptability of nearby points. The second involved a snubber on main steam line "B" that was found 1/4 inch outside of its normal operating range. This was evaluated and accepted "As-is" based on actual thermal growth and the remaining 3/4 inch for snubber movemen .

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TE-SU.ZZ-173. Remote sensor readings were made of the thermal ex-pansion of the main steam and recirculation piping during the first heatup to rated conditions. A Level 1 results deficiency was ident-ified for the "B" reactor recirculation discharge : loop. Since data was trended during the course of the heatup, the possible failure of this point was identified early to the General ~ Electric piping engineers and FDDR KT1-619 was issued to modify the thermal expansion

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limits for this point and resolve the deficienc In addition, 17 points were identified as Level 2 acceptance criteria violation These were accepted "As-is" based upon the walkdown of the piping discussed abov TE-SU.AE-236. This test was performed to demonstrate the stability of the feedwater startup level controller at the low pressure end of its operating range (tested at 187 psig reactor pressure). The acceptance criterion for non-divergent response to level step changes was satisfactorily demonstrate It was noted, however, that in the automatic control mode, limit cycle behavior of 2 to 3 inches occur-re In addition, although not an acceptance criteria at low pres-sure, the decay ratios of control system related variables exceeded the desired quarter wave damping criterion. The inspector plans to follow the resolution of these problems during future routine in-spection TE-SU.BB-332. All acceptance criteria for steady state vibration of the reactor recirculation system piping were satisfie TE-SU.BG-701. All acceptance criteria for the RWCU system in its normal performance mode were satisfie TE-SU.ZZ-003. The inspector reviewed the Test Matrix results package

for Test Condition Heatup presented by the Power Ascension Manager to the Station Operations Review Committee on August 5, 1986. The purpose of the presentation was to obtain approval of the General Man-ager - Hope Creek Operations to officially close testing in Test Condition Heatup. This plateau review included a presentation of all testing performed in the test condition and the results achieved, a summary of all incomplete testing.with justification supporting deferral to the next test condition, a review and evaluation of all open results deficiencies and unapproved on-the-spot procedure changes and a review and evaluation of open oc~urrence c list item Conformance with FSAR commitments and regulatory guidance was addressed and the Power Ascension Manager certified that the testing had demonstrated adequate systems performance to safely raise power and proceed with the next test plateau.

, Findings No unacceptable conditions were identified.

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4.0 Independent Measurements and Verifications The inspector performed multiple independent measurements and verifica-tions during the witnessing of power ascension testing (paragraph 3.4) and during the evaluation of power ascension test results (paragraph 3.5). In all cases the inspector's measurments and verifications agreed with those

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of the license No unacceptable conditions.were note ! 5.0 QA/QC Interface with the Power-Ascension Test Program During the course of witnessing power ascension testing the inspector observed QA engineers performing surveillances of selected power ascension tests. The inspector also noted during evaluation of power ascension test results that the test results packages had been reviewed by QA engineer No unacceptable conditions were note .0 Tours of the Facility In the course of witnessing power ascension test activities the inspector made tours of various areas of the facility to observe operations and test activities, housekeeping and cleanliness control No unacceptable conditions were identifie .0 Exit Interview At the conclusion of the site inspection on August 22, 1986, an exit meeting was conducted with the licensee's senior site representative (denoted in paragraph 1.0).

At no time during the inspection was written material provided to the licensee by the inspector. Based on the NRC Region I review of this re-port and discussions held with licensee representatives during the in-spection, it was. determined that this report does not contain information subject to 10 CFR 2.790 restrictions.

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Attachment A Power Asension Test Procedures Reviewed TE-SU.ZZ-004 Test Plateau Matrix Test Procedure for Test Condition One, Revisidn 1, Approved August 6, 1986 TE-SU.ZZ-021 Radiation Measurements, Revision 2, Approved June 24, 1986 TE-SU.ZZ-133 Prccess Computer 00-1 Verification, Revision 1, Approved August 8, 1986 TE-SU.SV-281 Shutdown from Outside the Control Room, Revision 1, Approved August 20, 1986-TE-SU.GT-722 Drywell Cooling System Normal Shutdown Performance Test, Revision 1, Approved June 16, 1986 TE-SU.ZZ-762 Confirmatory Test of Safety Relief Valve Discharge, Revision 2, Approved June 24, 1986 TE-SU.AC-774 Main Turbine First Stage Pressure Scram Bypass and Rod Sequence Control System Setpoints, Revision 0, Approved July 30, 1986

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Attachment B Power Ascension Tests Witnessed TE-SU.BD-144 Reactor Core Isolation Cooling System Cold Quick Start to the Reactor Pressure Vessel - Performed August 19, 1986 TE-SU.BJ-151 High Pressure Coolant Injection System Condensate Storage Tank Injection - Performed August 22, 1986 TE-SU.AE-251 Main Steam Isolation Valve Functional Test - Performed August 20, 1986 TE-SU.SE-261 Relief Valve Test - Performed August 21, 1986 TE-SU.SV-281 Shutdown from Outside the Control Room - Performed August 22, 1986 i

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Attachment C Power Ascension Test Results Evaluated TE-SU.ZZ-012 Chemical and Radiochemical Heatup Test, Revision 4, Completed July 24, 1986, Results accepted July 31, 1986 TE-SU.ZZ-021 Radiation Measurements, Revision 2, Completed July 29, 1986, Results accepted August 2, 1986 TE-SU.BF-052 Control Rod Drive System Scram Testing of Selected Control Rod Drives, Revision 2, Completed July 16, 1986, Results accepted July 27, 1986 TE-SU.BF-052 Control Rod Drive System Scram Testing of Selected Control Rod Drives, Revision 2, Completed July 16, 1986, Results accepted July 27, 1986 TE-SU.BF-055 Control Rod Drive System Control Rod Drive Friction and Scram Testing at Rated Pressure, Revision 3, Completed July 22, 1986, Results accepted July 31, 1986 TE-SU.BF-056 Control Rod Drive System Rated Reactor Pressure Control Rod Drive Functional Checks and Scram Testing of Selected Rods, Revision 2, Completed July 24, 1986, Results accepted July 31, 1986 TE-SU.SE-121 APRM Calibration During Heatup, Revision 13, Completed July 7, 1986, Results accepted August 11, 1986 TE-SU.BD-141 Reactor Core Isolation Cooling-System Condensate Storage Tank Injection, Revision 4, Completed July 9,1986, Results accepted July 29, 1986 TE-SU.BD-141 Reactor Core Isolation Cooling System Condensate Storage Tank Injection, Revision 8, Completed July 23, 1986, Results accepted August 2, 1986 TE-SU.BD-142 Reactor Core Isolation Cooling System Reactor Pressure Vessel Injection, Revision 5, Completed July 23, 1986 Results accepted August 4, 1986 TE-SU.BD-144 Reactor Core Isolation Cooling System Cold Quick Start to the Reactor Pressure Vessel, Revision 2, Completed July 30, 1986 Results not yet accepted TE-SU.BD-145 Reactor Core Isolation Cooling System Surveillance Test Demonstration, Revision 2, Completed July 31, 1986, Results not yet accepted

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o- Appendix C- 2 TE-SU.BJ-151 High Pressure Coolant Injection System Condensate Storage Tank Injection, Revision 5, Completed July . 29, 1986, Results accepted August 21, 1986

.TE-SU.ZZ-162 Water Level Measurement Test, Revision 3, Completed July 24, 1986 Results accepted July 31, 1986 TE-SU.ZZ-172 NSS Piping Expansion Walkdown, Revision 2, Completed July 3, 1986, Results accepted July 27, 1986 TE-SU.ZZ-172 NSS Piping Expansion Walkdown, Revision 3, Completed July 18, 1986, Results accepted July 31, 1986 TE-SU.ZZ-173 _NSS Piping Thermal Expansion Sensor Data, Revision 2, Completed July 25, 1986, Results accepted August 4, 1986 TE-SU.AE-236 Feedwater Startup Controller Test, Revision 3, Completed July 10, 1986, Results Accepted July 27, 1986 TE-SU.BB-332 Recir ulation System Piping Steady State Vibration, Revision 2, Completed July 17, 1986, Results accepted July 29, 1986 TE-SU.BG-701 Reactor Water Cleanup System Normal Mode-Performance Test, Revision 4, Completed July 26, 1986, Results accepted July 31, 1986 TE-SU.ZZ-003 Test Plateau Matrix Test Procedure for Test Condition Heatup, Revision 6, Completed August 5, 1986, Results accepted August 5, 1986

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