ML20212Q178

Insp Rept 50-354/86-35 on 860707-24.Violations Noted: Procedural Deficiency & Failure to Follow Startup Test Procedure & Adequately Review Completed Test

ML20212Q178
Person / Time
Site:	Hope Creek
Issue date:	08/13/1986
From:	Eselgroth P, Florek D, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20212Q155	List: ML20212Q151 ML20212Q160 ML20212Q178
References
50-354-86-35, NUDOCS 8609040039
Download: ML20212Q178 (14)

See also: IR 05000354/1986035

Text

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-354/86-35

Docket No. 50-354

License No. NPF-50

Licensee: Public Service Electric & Gas Company

Post Office Box 236

Hancocks Bridge, New Jersey 08038

Facility Name: Hope Creek Generating Station, Unit 1

Inspection At: Hancocks Bridge, New Jersey

Inspection Conducted: July /-24, 1986

Inspectors: 7M. fa,tw+ .1,., __7!/S!f6

0. Florek, Lead Reacftor Engineer '

d&te

S  % Y 19 Th

L. Wink, R c or En neer ' date

Approved by:

P..Eselgro W, Chief,

f//3 [fd

/ dite

,

Test Prog Nms Section, 08, DRS

Inspection Summary: Inspection on July 7-24, 1986 (Inspection Report Number

50-354/86-35)

Areas Inspected: Routine unannounced inspection of the overall power

ascension test program including procedure reviews, test witnessing and test

results evaluation, surveillance test activities, independent measurements and

evaluations, QA/QC interfaces and tours of the facility.

Results: Two violations were identified. One involved a procedural

deficiency (paragraph 2.4) and the other involved failure to follow a startup <

test procedure and failure to adequately review the completed test (2.5).

Note: For Acronyms not defined, refer to NUREG-0544 " Handbook of Acronyms

and Initialisms."

8609o40039 860s26

{DR ADOCK 05000354

PDR

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ ______ _____ ___________ __ _ _ __ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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DETAILS

1.0 Persons Contacted

,

Public Service Electric and Gas Company (PSE&G)

• o R. Beckwith, Station Licensing Engineer

G. Chew, Power Ascension Results Coordinator

• o G. Connor, Operations Manager

E. Dalton, Nuclear Systems Engineer

G. Daves, Senior Engineer, Operations

P. Dempsey, Shift Test Coordinator

M. Dick, Startup Test Engineer

• o M. Farshon, Power Ascension Manager

B. Forward, Power Ascension Administrative Coordinator

• o A. Giardino, Manager, Station QA

D. Hosmer, Lead Shift Test Coordinator

R. Hovey, Senior Nuclear Shift Supervisor

P. Krishna, Assistant to the General Manager

o S. La8runa, Assistant General Manager

• o M. Metcalf, Principal QA Engineer

L. Newman, Senior Nuclear Shift Supervisor

E. Rush, I&C Supervisor .

• o R. Salvesen, General Manager, Hope Creek Operations

W. Schell, Power Ascension Technical Director

E. Skeehan, GE Operations Manager

W. Thomas, Shift Test Coordinator

L. Zull, Lead Startup Test, Design and Analysis Engineer

U.S. Nuclear Regulatory Commission (NRC)

o D. Allsopp, Resident Inspector

R. Borchardt, Senior Resident Inspector

o S. Collins, Chief, Reactor Projects Branch No. 2

! o L. Norrholm, Chief, Reactor Projects Section, No. 2B

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• o R. Summers, Project Engineer

o Denotes those present at the interim exit meeting on July 17, 1986.

• Denotes those present at the exit meeting on July 24, 1986.

The inspector also contacted other members of the licensee's staff

including Senior Nuclear Shift Supervisors, Reactor Operators, Test

Engineers and other members of the Technical Staff.

2.0 Power Ascension Test Program (PATP)

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l 2.1 References

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Regulatory Guide 1.68, Revision 2, August 1978, " Initial Test

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Programs for Water-Cooled Nuclear Power Plants"

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ANSI N18.7-1976, " Administrative Controls and Quality Assurance

for the Operational Phrase of Nuclear Power Plants"

Hope Creek Generating Station (HCGS) Technical Specifications,

Revision 0, April 11, 1986

HCGS Final Safety Analysis Report (FSAR), Chapter 14 " Initial

Test Program"

HCGS Safety Evaluation Report (SER), Chapter 14, " Initial Test

Program"

Station Administrative Procedure, SA-AP.ZZ-036, Revision 3,

" Phase III Startup Test Program"

Specification NEB 0 23A4137, Revision 0, " Hope Creek Startup

Test Specification"

HCGS Power Ascension Test Matrix, Revision 7.

2.2 Overall Power Ascension Test Program

Scope

The inspector observed various activities associated with the conduct

of low power testing - these included daily management meetings,

operations shift turnovers, planning and scheduling meetings, test

results reviews by the Test Review Board (TRB) and the Station Opera-

tions Review Committee (SORC) and meeting to address specific prob-

lems identified during the performance of power ascension tests. The

inspector also observed the general conduct of operations and the

coordination of activities among the various departments.

Findings

No unacceptable conditions were identified.

2.3 Power Ascension Test Procedure Review

Scope

The Power Ascension Test Procedures listed in Attachment A were

reviewed for their conformance to the requirements and guidelines of

the references in paragraph 2.1 and for the attributes previously

defined in Inspection Report 50-354/86-03.

Findings

,

No unacceptable conditions were identified.

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2.4 Power Ascension Test Witnessino

Scope

The inspector witnessed portions of the power ascension tests

discussed below. The performance of these tests were witnessed to

verify that:

--

The current, approved revision of the test procedures was

available and in use by all participants.

--

The minimum crew requirements, as defined in the approved

procedure and technical specifications, were being met.

--

Required prerequisites and initial conditions were

established.

--

Test equipment was calibrated and operated in accordance with

procedure.

--

The procedure was technically adequate and appropriate to the

circumstances.

--

Crew performance was correct and timely.

--

Coordination of test related activities was adequate.

--

A preliminary analysis of all data collected was expeditiously

performed following completion of the test.

--

The test was performed in accordance with all administrative

requirements.

Discussion

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TE-SU.BF-052, Scram Testing'of Selected Rods. This test was

performed on July 16, 1986 at a reactor pressure of 800 psig.

The inspector observed the scram tests performed on control rods

34-07 and 34-23. The overall test crew performance was

satisfactory and coordination with ongoing surveillance testing

was well controlled by the shift operations personnel. The

inspector observed pre-scram data taking, initiation of testing

for each rod and data reduction following the testing of each

rod. All test results were well within acceptance criteria

Ifmits.

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TE-SU.BD-141, Reactor Core Isolation Cooling System Condensate

Storage Tank Injection. This test was performed on July 9,

1986 at a reactor pressure of 150 psig.

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The inspector witnessed the performance of a hot quick start to

the Condensate Storage Tank. Prior to the performance of this

portion of the test, the inspector observed an on-station test

briefing of the test crew by the test engineer. This test

brief was in addition to the required pre-test brief. This

additional briefing was very thorough and timely and

contributed to the subsequent smooth performance of the test.

During the performance of the quick start, the inspector

independently verified that the RCIC pump achieved a flow rate

greatar than 600 GPM within the required 30 seconds against an

appropriate discharge head. The inspector also observed ap-

propriate overall system performance and response.

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TE-SU.BD-142, RCIC Reactor Vessel Injection. This test was

performed on July 23, 1986 at rated reactor pressure.

The inspector witnessed test preparations, test briefings and

portions of the actual performance of the test. Test prepara-

tions were very thorough with operations personnel taking the

lead to assure that the test was effectively planned and imple-

mented. The test briefings were acceptable, the planned testing

was described and precautions discussed. Test personnel,

augmented by senior operations personnel, discussed several

potential problems that could occur during the test. Several

potential problems discussed were based on inspector questions

during the test preparation phase regarding the possibility of a

RCIC pump. suction swap from the Condensate Storage Tank to the

Torus.

The inspector witnessed the licensee's initial attempt'to

operate RCIC at rated flow into the reactor vessel. Because

reactor power was not quite high enough, reactor vessel level

continued to increase, even with no flow from the feedwater

pumps. The licensee subsequently, raised power to 4.1% of rated

and was then able to perform the remainder of the test.

Having established stable conditions, with full RCIC flow into

the reactor vessel, the licensee prepared to perform step 5.3.1

of the procedure which required RCIC turbine speed to be

reduced to 2150 (+1C0, -0) RPM and the subsequent performance

of speed steps. The inspector inquired as to the appropriate-

ness of this step, since this low speed would result in no flow

to the reactor vessel. After review by the licensee, it was

determined that the step was not appropriate and the procedure

was changed to require a specific pump flow rather than turbine

speed. This procedure sequence could have resulted in equipment

damage and low reactor vessel water level if feedwater pump flow

was not manually increased. 10 CFR 50, Appendix B, Criterion V

states, in part, that " activities affecting quality shall be

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prescribed by documented procedures appropriate to the circum-

stances." Technical Specification 6.8.la through its reference

to Regulatory Guide 1.33 Revision 2, 1978 and ANSI N18.7. 1976

Section 5.3 contain similiar requirements. Contrary to these

requirements, step 5.3.1 of procedure TE-SU.BD-142 was not appro-

priate to the circumstances and is considered a violation

(354/85-35-01). The licensee took prompt action to correct this

specific procedural problem at tFe time it was identified by the

inspector.

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The inspector also witnessed the hot quick start portion of

this test. RCIC was observed to achieve rated flow into the

reactor vessel within twenty seconds.

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TE-SU.AE-236, Feedwater Startup Controller Test. This test

was performed on July 7, 1986 at a reactor pressure of 145

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psig.

The inspector witnessed the test briefing and performance of

level step changes in both manual and automatic modes of con-

trol. The inspector noted that while the controller was not

divergent (Level 1 acceptance criterion) in either the manual

or automatic modes it did exhibit limit cycle behavior in auto-

matic. This limit cycle behavior made it essentially impossible

to achieve the stable level conditions required prior to the

performance of the level step changes in automatic and impos-

<

sible to distinguish the response of the system to these changes

from the limit cycling. The inspector also noted that the Level .

2 acceptance criterion for these step changes was not applicable s

to the current test condition. I

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Following the completion of the test and examination of the

recorded test data, the test engineers decided to abort the test

and attempt to reperform it at a later time. A retest was

performed on July 9,1986.

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TE-SU.GT-721, Drywell and Steam Tunnel Cooling System - Normal

operation performance. This test yas performed on July 16, 1986

l at a reactor pressure of 800 psig.

The inspector observed the post test evaluation of recorded

l data and comparison to acceptance criteria. The licensee deter-

' mined that three Level 2 acceptance criteria were not satisfied

and issued results deficiency form No. 048 to document and track

the resolution of these problems. The problems all involved

local temperatures in excess of the 150* F acceptance limit and

circumferential variations in temperature greater than the 10* F

acceptance limit. The Level 1 acceptance criterton for the

volumetric average drywell temperature of 135* F was satisfied.

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l The inspector observed a licensee meeting, which was held in the

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control room, to discuss the test results, impact on future

l testing and possible resolution to the problem. Present for

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____ __ ________ _ _ ___- ___ _______-__ _ __

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this discussion were the Operations Manager, Power Ascension

Manager, Senior Nuclear Shift Supervisor, Shift Test Coordinator

and Test Engineer. The discussion was comprehensive and inte-

grated all available data, including the results of a special

drywell entry made to assess the situation. Short term plans

were formulated to ameliorate the problem, future tests were

discussed and plans were made for a formal engineering meeting

to assess the data, success of the short term plans and formu-

late any required long term solutions.

Findings

One violation was identified in the above discussion for pro-

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cedure TE-SU.BD-142, which was inappropriate to the

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circumstances.

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2.5 Power Ascension Tests Results Evaluation

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Sc99_e

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The power ascension test results discussed be kw were evaluated for

the attributes identified in Inspection Report 50-354/86-24.

Otscussion

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TE-SU.22-018, Process Radiation Monitoring Baseline Data. Test

completed June 18, 1986, results review completed July 22, 1986,

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pending management acceptance.

This test contains no acceptance criteria. It gathers baseline

j data (prior to initial criticality) which will be used to es-

tablish setpoints for the Main Steam Line Radiation Monitors.

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TE-SU.22-041, Full Core Shutdown Margin Demonstration. Test

completed June 29, 1986, results review completed July 10, 1986

and accepted by management on July 17, 1986.

All acceptance criteria were met. Initial criticality was

achieved on RWM step 84, RSCS Group 3, Rod 46-27 at Notch 8

(2272 total notches) at a moderator temperature of 123' F. The

inspector independently calculated the shutdown margin and veri-

fled the licensee's value of 2.4% delta K/K. The rod notch

limits to satisfy the 1% delta K/K reactivity anomaly criterion

at 123' F were 1428 to 2364 notches. Criticality was achieved

within these limits.

i During a detailed review of the test package the inspector

discovered a calculational error which had not been identified

during the results review. Prerequisite step 2.15 requires the

test engineer to make a " Predicted Critical Calculation" in

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accordance with Attachment 2 of the procedure and to use the

results in Caution Statement 5.2.2. Attachment 2 provides the

equations to be used to perform a calculation of the estimated

critical rod position and calculations of minimum and maximum

allowable rod withdrawal positions based on the 11% reactivity

anomaly limits. These limits on rod withdrawal are then entered

in caution statement 5.2.2 to prevent test personnel from un-

knowingly exceeding these limits. Attachment 2 requires that

the completed calculations be independently reviewed.

The inspector noted that the calculation of the expected

critical position was correctly performed but that the

calculation of the minimum and mayimum Tod withdrawal limits

were not performed in accordance with the procedure. The

procedure indicated that the calculations had been reviewed by

a senior test engineer. The result of the error was that the

minimum allowable rod withdrawal limit was non-conservative.

The potential consequences of this error are of minor safety

sigt'ificance since the reactivity anomaly limits were being

independently monitored by the licensee's reactor engineer

personnel during the performance of this test and these er-

roneous calculations were not used to verify acceptance criteria

due to moderator temperature changes that occurred during the

test. However, the error was made and the licensee's detsiled

review of the test results failed to identify the deficiency.

10 CFR 50, Appendix B, Criterion XI states, in part,

that " testing shall be performed in accordance with written

test procedures," and further states that " test results shall

be evaluated to assure that the test requirements have been

satisfied." Contrary to these requirements the calculation of

minimum and maximum rod withdrawal limits were not performed in

accordance with the written procedure and the review of the test

results failed to identify this deficiency. This is considered ,

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a violation (354/86-35-02).

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TE-SU.SE-062, Source Range Monitor Response to Rod Withdrawal.

Test completed June 28, 1986, results review completed July 10,

1986 and accepted by management on July 14, 1986.

This test contains no acceptance criteria. It records SRM

response to each step of the Rod Pull Sheet for reference during

future startups. It also produces 1/M plots for the purpose of

anticipating criticality.

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TE-SU.SE-064, Source Range Monitor Non-Saturation

Demonstration. Test completed June 29, 1986, results review

completed July 10, 1986 and accepted by management on July 14,

1986.

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This test contains no acceptance criteria. It denonstrates

that the SRMs would not saturate below 150% of their normal

scram setpoint (2 x 10 5cps). All SRMs were demonstrated to be

capable of recording a count rate of at least 5 x 105 cps.

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TE-SU.SE-101, Source Range Monitor / Intermediate Range Monitor

Overlap Verification. Test completed June 29, 1986, results

review completed July 10, 1986 and accepted by management on

July 14, 1986. .

During the initial performance of this test IRM "A" failed to

respond to increasing neutron flux. Results Deficiency Forms

Nos. 38 and 39 were written to document the failure of IRM "A"

to satisfy the Level 1 and Level 2 acceptance criteria and to

allow continued operation and testing in accordance with

technical specifications. IRM "A" was repaired and

successfully retest on July 1,1986 to close RDFs Nos. 38 and

39. All acceptance criteria were satisfied.

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TE-SU.SE-102, IRM Range 6-7 Continuity Verification. Test

completed June 30, 1986, results review complete July 22, 1986,

pending management acceptance.

This test contains no acceptance criteria. It verifies IRM

continuity when switching from the low frequency amplifier

(Ranges 1-6) to the high frequency ampitfier (Ranges 7-10).

All IRM channel required a range correlation adjustment to

achieve the desired continuity. IRM "A" was inoperable during

the initial performance of this test and was tested separately

on July 1, 1986.

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TE-SU.BD-141, Reactor Core Isolation Cooling Condensate Storage

Tank Injection. Test performed July 21, 1986.

This test was aborted due to a RCIC turbine trip. The turbine

tripped on overspeed due to controller hardwrce problems and

lack of a flowpath to the CST. This test also caused an over-

pressure condition to exist in the RCIC discharge piping (pres-

sure extrapolated from available test data to be 1900 psig for

less than five seconds). To prevent a reoccurrence, a procedure

change was made to assure that the CST test return valve and the

RCIC steam admission valve are opened simultaneously rather than

in series. The inspector reviewed the safety evaluation per-

form 1d on the affected piping which indicated that the piping

was not damaged during the event.

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TE-SU.AB-251, Main Steam Isolation Valve Functional Test. Test

completed July 21, 1986.

Licensee review of the test results is in progress. The in-

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spector performed independent calculations of several MSIV

stroke times based on recorded data. For those calculations

performed, the acceptance criteria of 3 to 5 seconds were

satisfied.

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TE-SU.GT-721, Drywell and Steam Tunnel Cooling System - Normal

Operation Performance. Test completed July 17, 1986.

Licensee review of the test results is in progress. The

calculated drywell volumetric average temperature of 107" F

satisfied the Level 1 acceptance criterion. Level 2 acceptance

criteria failures occurred for maximum local air temperature and

maximum temperature differentials and are being processed via

the administrative controls defined. The inspector questioned

the licensee on the minimum temperature limits for the under-

vessel area. The inspector noted that this area was being

cooled to 70* F. The inspector inquired as to whether this

temperature was consistent with General Electric Design Require-

ments for the reactor vessel support skirt. The licensee re-

sponse was not formulated at the conclusion of this inspection.

Pending the response of the licensee, this item is considered

unresolved (354/86-35-03).

In addition to the test results evaluated above, the inspector

reviewed the status of 17 results deficiency forms (RDFs)

generated during low power testing. Six RDFs were resolved but

not formally closed. The remaining RDFs still require

resolutions.

Findings

One violation was identified for failure to perform testing in

accordance with written procedures and for failure to

adequately review. One unresolved item was identified for

minimum temperature limits in the drywell undervessel area.

3.0 Surveillance Test Activities

Scope

The inspector witnessed portions of the operations surveillance tests

discussed below. The tests witnessed are closely related to testing

performed in the power ascension test program and were evaluated for the

attributes defined in paragraph 2.4.

Discussion

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OP-ST.SN-001, ADS and Safety Relief Valve Manual Operability Test -

18 months. Test completed July 8, 1986.

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The inspector witnessed the test preparations, test briefing and

test performance of Section 5.1 of the test procedure to satisfy, in

part, the technical specification requirements for ADS valves

operability. All aspects of the testing were performed in a

thoroughly competent and professional manner. Four of the five ADS

were satisfactorily tested and the inspector independently verified

various diverse indications of actual steam flow through the valves.

One ADS valve failed to respond to the manual actuation. Operations

personnel responded appropriately and in accordance with technical

specification requirements. Immediate plans were made to isolate and

identify the cause of the failure. Subsequent investigation identi-

fled the problem as a crimped Containment Instrument Gas Line to the

valve pilot. The line vas repaired and the valve was successfully

retested.

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OP-ST.BJ-002, HPCI System Functional Test (Low Pressure) - 18

months. Test completed July 10, 1986.

The inspector witnessed the performance of Section 5.2 of the test

procedure to satisfy, in part, the technical specification require-

ments for HPCI response time. During initial attempts to perform the

quick start required for the test difficulties were encounter in

stroking the fell flow test return line to establish a flow path back

to the CST. A procedure change was made to allow positioning of the

valve prior to manually initiating HPCI, Following this change a

quick start was successfully performed but the time to rated flow

exceeded the technical specification requirement of 27 seconds. It

was noted that the discharge head was significantly higher than re-

quired (700 psig versus 300 psig) due to the position of the full

flow test return valve. The valve was throttled open slightly and

the quick start was satisfactorily completed. All aspects of

operations personnel performance were judged to be satisfactory.

During the performance of this test the inspector noted that the

observed correlations between HPCI turbine speed, pump flow and

discharge pressure were not consistent with expected relationships.

The inspector obtained preliminary data from the performance of

Power Ascension Test TE-SU.BJ-151, HPCI Condensate Storage Tank

Injection, which was performed subsequent to the surveillance test

and confirmed these observations. The inspector questioned the

licensee on this anomalous correlation and the licensee agreed that

the data recorded was not consistent with the HPCI pump curves.

Subsequent investigation revealed a instrumentation problem with the

turbine speed indication. This problem should not affect the

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l validity of the surveillance test but may result in invalidating all

! or part of the power ascension test. The inspector will review the

j licensee's determination of the effect of this problem during a

subsequent routine inspection.

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Findings

No unacceptable conditions were identified.

4.0 Independent Measurements and Verifications

The inspector performed multiple independent measurements and

verifications during the witnessing of power ascension (paragraph 2.4)

and surveillance (paragraph 3.0) testing and during the evaluation of

Power Ascension Test Results (paragraph 2.5). Except as discussed in

paragraph 2.5, the inspector's measurements and verifications agreed with

those of the licensee.

5.0 QA/QC Interface with the Power Ascension Test Program

The inspector reviewed QA Surveillance Report No.86-187 covering the

performance of Power Ascension Test TE-SU.ZZ-041, Full Core Shutdown

Margin Demonstration. The inspector also observed QA involvement in the

review of power ascension test results. No unacceptable conditions were

noted.

6.0 Tours of the Facility

In the course of witnessing power ascension and operations surveillance

test activities the inspector made tours of various areas of the facility

to observe work in progress, housekeeping and cleanliness controls. No

unacceptable conditions were identified.

7.0 Unresolved Items

Unresolved items are matters about which more information is required in

order to determine whether they are acceptable, an item of noncompliance

or a deviation. Unresolved items disclosed during the inspection are

discussed in paragraph 2.5.

8.0 Exit Interview

At the conclusion of the site inspection on July 24, 1986, an exit

meeting was conducted with the licensee's senior site representative

(denoted in Section 1.0). In addition, an interim exit meeting was held

with the licensee on July 17, 1986.

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

report and discussions held with licensee representatives during the

inspection, it was determined that this report does not contain

information subject to 10 CFR 2.790 restrictions.

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ATTACHMENT A

POWER ASCENSION TEST PROCEDURES REVIEWED

TE-SU.ZZ-013, Chemical and Radiochemical 15 to 25% Power Tests, Revision 1,

Approved May 16, 1986.

TE-SU.ZZ-017, Verification of Compliance for Discharge, Revision 0, Approved

June 10, 1986.

TE-SU.ZZ-018, Process Radiation Monitoring Baseline Data, Revision 0,

Approved June ?7, 1986.

TE-SU.SE-103, Intermediate Range Mor.itor/ Average Power Range Monitor Overlap

Verification, Revision 1, Approved April 28, 1986.

TE-SU.SE-112, Local Power Range Monitor Calibration, Revision 1, Approved

April 30, 1986.

TE-SU.SE-122, APRM Calibration at Power, Revision 1, Approved

February 11, 1986.

TE-SU.ZZ-170, Residual Heat Removal System Piping Expansion During Shutdown

Cooling Operation, Revision 3, Approved June 5, 1986.

TE-SU.ZZ-176, Balance of Plant Systems Piping Expansion During Power

Operations, Revision 1, Approved April 30, 1986.

TE-SU.BB-221, Pressure Regulator Test - Control Valves Controlling,

Revision 1, Approved March 26, 1986.

TE-SU.BB-223, Pressure Regulator Test - Bypass Valves Controlling, Revision 1,

Approved March 15, 1986.

TE-SU.AE-231, Feedwater System Level Setpoint Steps, Revision 0, Approved

October 15, 1985.

TE-SU.AE-235, Manual Feedwater Flow Step Change Test, Revision 0, Approved

October 15, 1985.

TE-SU.AB-251, Main Steam Isolation Valve Functional Test, Revision 1,

Approved June 10, 1986.

TE-SU.ZZ-262, Relief Valve Response During Major Trips, Revision 1, Approved

June 25, 1986.

TE-SU.BB-303, Recirculation System Performance, Revision 0, Approved

October 28, 1985.

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TE-SU.AB-331, NSSS Main Stean Piping Steady State Vibration, Revision 2,

Approved May 20, 1986.

TE-SU.BB-332, Recirculation System Piping Steady State Vibration, Revision 2,

Approved May 10, 1986.

TE-SU.AB-341, NSSS Main Steam riping Dynamic Response, Revision 1, Approved

May 23, 1986.

TE-SU.BB-343, BOP Main Steam Piping Dynamic Response, Revision 3, Approved

April 22, 1986.

TE-SU.AB-344, Main Steam Relief Valve Discharge Piping Dynamic Response,

Revision 2, Approved April 15, 1986.

TE-SU.GT-723, Drywell Cooling System Post Trip Performance Test, Ravision 1,

Approved June 19, 1986.

-TE-SU.GT-724, Containment Penetration Cooling Normal Operation Performance

Test, Revision 1, Approved June 16,-1986.

TE- SU . HA-741, Gaseous Radwaste System Performance Test, Revision 0, Approved

October 9, 1985,

1E-SU.ZZ-761, In-Plant Safety Relief Valve Shakedown Test, Revision 2,

. Approved June 24, 1986.

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