IR 05000445/1993002

Insp Repts 50-445/93-02 & 50-446/93-02 on 930104-30. Violations Noted.Major Areas Inspected:Preoperational Test Program Implementation,Deferred Tests & Retests,Qa, Procedures,Results Evaluations & Pipe Supports

ML20044C436
Person / Time
Site:	Comanche Peak
Issue date:	03/12/1993
From:	Constable L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20044C434	List: IR 05000445/1993002 ML20044C433 ML20044C435
References
50-445-93-02, 50-445-93-2, 50-446-93-02, 50-446-93-2, NUDOCS 9303230033
Download: ML20044C436 (63)

Text

D :.

~

!

l f

P t

.

-

APPENDIX B i

L

-

<

U.S. NUCLEAR ~ REGULATORY COMISSION:

i

REGION IV

i NRC Inspection Reports: 50-445/93-02

_;

50-446/93-02 i

Operating License: NPF-87 j

.

Construction Permit: CPPR-127

Licensee:

TU Electric

!

400 North Olive Street Lock Box 81-Dallas, Texas 75201 l

Facility Name: Comanche Peak Steam Electric Station (CPSES)

-

Inspection At: CPSES Glen Rose, Texas Inspection Conducted: January 4-30, 1993

!

Inspectors: Howard F. Bundy. Reactor Inspector, Plant Support Section Division ~of Reactor Safety.

'

!

Lawrence E.- Ellershaw, Reactor Inspector,_ Maintenance Section Division of Reactor Safety

Ronald K. Frahm, Jr., Vendor Inspector. (Rotational Assignment),

Engineering Section, Division.of Reactor Safety-

'

'

Dennis L.~Kelley,-Reactor ~ Inspector, Maintenance Sect' n h

Division of Reactor Safety

'

Claude E. Johnson, Reactor: Inspector, Maintenance Section '

l Division of. Reactor Safety-l Ryan Lantz, Reactor Inspector.(Examiner), Operations Section Division of Reactor Safety

~

j Michael E. Murphy, Reactor Inspector, Plant Support Section Division of Reactor Safety

William McNeill, Reactor Inspector, Engineering Section

-

Division of Reactor. Safety

'

,

i Dr. Dale A. Powers, Senior Reactor Inspector, Maintenance'-

Section, Division of Reactor Safety-

'

Amarjit Singh,' Reactor Inspector, Plant Support Section'

!

Division:of Reactor Safety l

,

.

9303230033 930317

!

PDR-ADOCK 05000445 G.

PDR

,

,

,

,

,

(;

,'~

,.

.

John E. Whittemore,- Reactor Inspector, Plant Support Section Division of Reactor Safety

,

~ [

Approved:

[_

~

6[te-T /2 eJL st e,

hief, Plant Support A

Section, Division of Reactor Safety Inspection Summary Areas Inspected (Unit 1): No inspection of Unit I activities was performed.

'

Results:

Not applicable.

• Areas inspected (Unit 2): Routine, announced inspection of preoperational test program implementation, deferred tests and retests, quality assurance, procedures, results evaluations, and pipe supports and restraints; startup program implementation, quality assurance, and procedures; licensee event reports; and followup of corrective actions for violations.

Results:

The licensee's rationale for deferring certain preoperational tests and

retests was reasonable. The licensee's schedule for developing the -

necessary test procedures and the intended schedular means to control plant mode changes prior to completing the necessary testing were appropriate (Section 2.1).

A quality assurance program for maintenance that conformed to the

appropriate requirenents had been implemented for Unit 2.

Qualified personnel and procedures were available to support maintenance activities. The licensee's program to be used for Unit 2 design changes and modifications is the same as that used for Unit 1, and its implementation should be adequate. The licensee had implemented suitable programs for control and evaluation of surveillance testing, calibration, and inspection as required by technical specification and for calibration of safety-related instrumentation not specifically controlled by technical specification (Section 2.2).

Preoperational Test Procedure 2CP-PT-64-08, " Reactor Trip System Time

• -

Response Summary," was acceptable. The preoperational test procedure:

appeared to be responsive to the needs for collecting, calculating, ~and documenting the. reactor trip system time responses (Section 2.3).

The licensee's acceptance ' criteria were met, or deviations were properly

dispositioned, for all 20 preoperational tests for which results were inspected.

The licensee's result evaluations of the preoperational tests were detailed,' comprehensive, and generally of excellent' quality.

The licensee identified numerous (i.e., several hundreds) administrative-

-2-

-s

F..

.

errors were made during the conduct of the tests by involved startup

,

test engineers. The licensee did not identify two preoperational retests that were preformed prior to receiving management approval. As a result, a violation was identified (Section 2.4).

The licensee had properly maintained data and documentation related to

the result evaluations of tests of piping supports and restraint systems. These systems were found acceptable, or were evaluated or modified, and found to satisfy design requirements.

Pipe supports that were examined were in satisfactory condition (Section 2.5).

,

'

Startup administrative procedures were well written. The procedures

contained explicit and detailed instructions as to the conduct of initial startup and power ascension testing, and should provide for positive and comprehensive control. The administrative procedure for.

.

qualifications appeared to appropriately enable control over test personnel qualifications. The station administrative procedures needed updating to properly identify organizations and titles of responsibilities for the conduct of the initial startup-and power

-

ascension program (Section 3.1).

The licensee had established an adequate program for audit and

,

surveillance of initial startup and power ascension test activities (Section 3.2).

The initial startup and power ascension procedures reviewed appeared to

be technically adequate. They were responsive to Final Safety Analysis Report commitments, and conformed to applicable regulatory guidance, industry standards, and approved administrative controls. The procedures suitably included necessary acceptance criteria and appropriately allowed for the control of work activities by use of sign-

off steps for accomplished tasks. The procedures, in general, presented a cautious and controlled c oach to achieving initial criticality and subsequent startup activities. However, three procedures were inappropriate in that they required the performance of " appropriate,"

" applicable," or unclearly defined portions of ancillary procedures. As a result, a violation was identified (Section 3.3).

Licensee corrective actions for a previous violation were not fully

effective in that the inspectors found a control room copy of a procedure that had not been updated (Section 5).

The licensee's management committed to three actions in response to the

violations. The first action was to have power ascension training for test personnel to assure that they are thoroughly familiar with the administrative requirements of the program, and to review test'results-promptly to provide feedback as to the effectiveness of the training.

The second action was to revise or change initial startup procedures prior to use to provide a description of each section required for any.

" branch out" procedures. The third action was to intensify corrective-3-

e

if.

,

.

.

action to ensure control room procedures are up to date.(Atthchment 1,

,

Section 2).

Summary of Insoection Findinos:

,

Construction Deficiency Reports CP-92-005 and -012 were closed

(Section 4).

Violation 446/9232-01 was closed (Section 5).

• Violations 446/9302-01 (paragraphs 2.4 and 3.3) and 446/9302-02

'.

(Section 5) were opened.

Inspection Followup Item 446/9302-03 (paragraph 2.4) and 446/9302-04

(Section 2.1) were opened.

Attachments:

Attachment 1 - Persons Contacted and Exit Meeting

Attachment 2 - Deferred Tests and Retesti Assessment

'I Attachment 3 - Pipe Supports Examined

Attachment 4 - Procedures and Data Reviewed

,

e-4-

~

U.

c O

TABLE OF C0KTENTS

,

Eitg.it DETAILS...................................................................

I 1.

PLANT STATUS..........................................................

I 2.

PRE 0PERATIONAL TEST PR0 GRAM............................................

2.1 Overall Preoperational Test Program Review Requirements..... I 2.2 Qual i ty Assurance Program.................................. 2 2.2.1 Maintenance................................................

2.2.2 Quality Assurance Program for Design Changes and Modifications............................................

2.2.3 Surveill ance Testing and Calibration Control............... 6 2.2.4 Overall Conclusions of Quality Assurance Program...........

2.3 Procedure Reviews..........................................

2.3.1 Reactor Protection System Test - Preoperational Test -

Procedure Review.........................................

2.4 Preoperational Test Resul ts Eval uati ons.................... 8 2.4.1 Containment Testing........................................

2.4.2 Integrated Hot functi onal Test............................

2.4.2.1 RCS Equipment Supports Thermal Expansion Test.............

2.4.2.2 Hot Functional Piping Systems Thermal Expansion. Test......

2.4.3 Reactor Protection System (RPS)...........................

2.4.4 Loss o f Offsi te Powe r Tes t................................

2.4.5 Engineered Safety Features (ESF) Tests....................

2.4.6 Emergency-Standby Power Supply System Test................

2.4.7 Engineered Safety Features Actuation System (ESFAS) Test.. 20 2.4.8 Main Steam I sol ation Val ve Test........................... 22 2.4.9 Auxiliary Feedwater System Test...........................

2.4.10 Overall Conclusions of Preoperational Test Results E v a l u a t i o n s.............................................. 2 4 2.5 Testing Piping Supports and Restraint Systems.............

2.5.1 Preoperational Test Resul ts...............................

2.5.1.1.1 The rmal Exp an s i on Te s t..................................... 2 5 2.5.1.1.2 Dynamic Equipment Transient Response Testing..............

2.5.1.1.3 Steady-State Vibration Monitoring Test....................

2.5.2 As-Built Observations.....................................

2.5.3 Inservice Inspection Program..............................

2.5.4 Overall Conclusions of Testing Piping Supports and Restraint Systems.......................................

-1-

r

.

.

, 3.

STARTUP PR0 GRAM.......................................................

3.]

Overall Startup Test Program..............................

28-

'

3.2 Quality Assurance for the Startup Program................. 30 3.3 Procedure Reviews.........................................

3.3.1 Initial Fuel Loading Procedure............................ 31'

i 3.3.2 Control Rod Operability...................................

3.3.3 Reactor Coolant System Leakage............................

33-3.3.4 I ni t i al Cri t i ca11 ty........................................ 34 3.3.5 Low Power Physics Testing.................................

3.3.6 Core Re acti vi ty Bal ance.................................... - 37

,

3.3.7 Reactor Power Test Sequence...............................

3.3.8 Load Reduction / Rejection Tests............................

3.3.9 Loss of Offsite Power.....................................

3.3.10 Core Performance and Physics Code.........................

3.3.10.1 Core Performance..........................................

i 3.3.10.2 Core Physics Code.........................................

3.3.11 I nco re Fl ux Mapp i ng....................................... 4 3 3.3.12 Overall Conclusions of Procedure Reviews..................

,

i 4.

ONSITE REVIEW OF LICENSEE EVENT REP 0RTS.............................. 44 5.

FOLLOWUP ON CORRECT IVE ACTIONS FOR VIOLATIONS.........................- 4 5 -

ATTACHMENT 1 - Persons Contacted and Exit Meeting ATTACHMENT 2 - Deferred Tests and Retests Assessment ATTACHMENT 3 - Pipe Supports Examined ATTACHMENT 4 - Procedures and Data Reviewed

.

+

i i-11-

,

.

!

DETAILS

<

!

1 PLANT STATUS During this inspection period, the licensee completed construction of many plant systems, components, and compartments and turned such items over for operation.

Preoperational test program activities, except the deferred tests

and retests, were completed. The licensee's site organization was changed

'

following the completion of various portions of the preoperational test program.

2 PREOPERATIONAL TEST PROGRAM 2.1 Overall Preooerational Test Procram Review Reouirements (70301)

In this area of the inspection, the inspectors reviewed the preoperational and acceptance tests requested to be deferred or retested following issuance of the low-power license. The affected systems and components were identified in-the licensee's letter (TXX-93011) dated January 8, 1993. Subsequently, the

'

licensee updated its letter by letter (TXX-93051) dated January 25, 1993, to include additional affected systems and components.

i 2.1.1 Discussion The inspectors discussed with the licensee's representatives the rationale for deferring the specifically identified tests and retests. The licensee's rationale for each test and retest appeared to be reasonable. The inspectors reviewed each deferred test and retest against Technical Specification (TS)

mode requirements and the scheduled modes for testing and confirmed that the deferred tests and retests should not adversely impact on regulatory and safety requirements.

The licensee's letters indicated that some of the deferred tests and retests had been scheduled for completion. The inspectors' requested copies of documentation that confirmed that the scheduling had been completed. The inspectors learned that the individual task schedules had been placed in the licensee's comitment tracking system and were being placed in the computer-based work scheduling program. The inspectors were also informed that the-

,

'

licensee was establishing in the computer program, which is used for the scheduling of surveillances (PR-ISM, " Plant Reliability - An Integrated System

,

for Management"), the necessary mode restrictions to ensure the completion of the deferred tests and retests when required.

The inspectors determined that the licensee had established which procedures would be necessary for the performance of most of the deferred tests and retests. Those procedures, which were identified to the inspectors, along with the predicted approval dates of needed procedures are documented in

'

Attachr.ent 2 to this report. The inspectors questioned the licensee whether it would use mode change sheets to ensure that the operational staff would not inappropriately change modes without completing the necessary tests and retests.

In response to the question, the licensee's representative indicated

_]_

.

.

.-

-

.

that the work controls and restrictions for deferred tests and retests would be discussed at plin-of-the-day meetings, but would not be placed on mode change sheets.

!

The inspectors questioned the licensee's representatives as to whether any commitments were relevant to the deferred tests or retests [such as responses to NRC staff questions on the Final Safety Analysis Report (FSAR)]. -The licensee's representative subsequently examined the commitment data base and determined that the deferred tests and retests would not contradict any prior licensee commitments. The significant nunber of deferred tests and retests is an NRC concern and the successful completion of these tests will be tracked as

,

an Inspection Followup Item (446/9302-04).

,

2.1.2 Conclusions The licensee's rationale for deferring certain preoperational tests and retests was reasonable. The specified tests and retests to be deferred should not result in any system or component being inoperable when operability is required and should not adversely impact on fuel loading or initial physics testing. The~ licensee's scheduling process for completing the necessary testing was incomplete as of the end of the inspection period, but appeared to be satisfactory to ensure that testing activities were completed prior to the committed times or plant modes.

2.2 Ouality Assurance (0A) Procram 2.2.1 Maintenance (35743)

The purpose of this part of the inspection was to ascertain that the licensee had implemented a QA program relating to maintenance activities for Unit 2 that was in 'conformance with proposed TS, regulatory _ requirements, commitments in the application, and industry guides and standards. An inspection of the QA program relating to maintenance activities associated with Unit I had been documented in NRC Inspection Report 50-445/90-12. Therefore, the scope of this inspection was limited to verifying implementation of the licensee's QA program relating to maintenance activities for Unit 2.

2.2.1.1 Discussion for familiarization with licensee QA requirements, the inspectors reviewed the

>

following sections of the QA Manual:

Section 9.0, " Control of Special Processes," Revision 2;

Section 13.0, " Handling, Storage, and Shipping," Revision 1; and

Section 14.0, " Inspection, Test, and Operating Status," Revision 1.

• The inspectors also reviewed the following nuclear overview department procedures:

NQA 3.09-9.01, " Review of Quality Related Work Orders," Revision 6; and

NQA 3.23, " Surveillance Program," Revision 6.

-2-

r-

-

-

. - -

.

.

,To ascertain implementation of the maintenance program for Unit 2, the inspectors reviewed Report IV7W.U2AVDCDECQ, " Unit 2 Work Orders Vaulted (AVDC)

During December 1992," dated January 5, 1993. The inspectors selected the following work orders (W0s) from this report for further review:

WO l-92-015359-00, " Rework CCW Pump 2-01 Outboard and Inboard Cooler

Flanges to Correct Oil Leakage";

WO l-92-005243-00, " Repack Safety Injection Accumulator 2-03 Valve

2-8875C"; and

'

WO l-92-005151-00, " Refurbish Actuator for Pressurizer Auxiliary Spray

Valve 2-8145."

The inspectors found that the work related to the above W0s was completed and documented in accordance with industry standards and regulatory guidance. The work had been properly approved and performed. Suitable procedures were included in the work packages. The documentation clearly indicated the actions that had been completed. The packages provided traceability for the measurement and test equipment, replacement parts, and materials used.

Appropriate reviews and approvals were indicated in the documentation.

For WO 1-92-015359-00, the crafts personnel did not enter a possible cause for the oil leak. The inspectors observed that the crafts personnel should be encouraged to enter a possible cause for failure of the component, or note that the cause of failure was indeterminate.

To ascertain implementation of the licensee's preventive maintenance (PM)

program for Unit 2, the inspectors reviewed the following reports:

Report 06.49.03, PM Report," dated January 7, 1993; and

Report 06.19.23, "PM Work Orders," dated January.7, 1993.

• The inspectors discussed the PM program with the PM planning and sWduling personnel They estimated that there are approximately 10,000 scheduled PM items which have been entered in the computerized planning system. The program was considered complete. The inspectors reviewed the documentation for several of the PM items which were indicated as complete in the schedule.

In each instance the documentation had been properly completed.-

The inspectors verified the licensee's Unit 2 program implementation of selected special processes by review of documents and interviews with selected personnel. The-follou ng were among the documents reviewed:

Report ID:817-8975714, " Active Listing of Nuclear Overview Department

Personnel Working for Bob Spence," dated January 5,1993; Report " Unit 2 Construction Welder Qualification Update List," dated

December 8, 1992;

_3_

-

..

.

Station Administrative Procedure STA-745, " Nuclear Operations Welding Program," Revision 1; and Welding Program Manual, Revision 64.

• The first document referenced above was a summary of the personnel certified to perform the various nondestructive examination processes at the site. ~From this list, it appeared that sufficient expertise was available on-site to support Unit 2 maintenance activities. According to the station welding engineer, the Unit I and Unit 2 welding programs were merged on January 8, 1993. He stated that on January 22, 1993, the Unit I and Unit 2 welder qualification update lists would be merged. Station Administrative Procedure

'

STA-745 was the governing document for the site welding program. The welder qualification update list indicated the process and procedures each welder was qualified to perform. The Welding Program Manual contained both administrative and specific welding procedures.

Qualifications for the specific welding procedures were on file. The site welding program supported Unit 2 maintenance activities.

2.2.1.2 Conclusions A QA program (maintenance) that conformed to the appropriate requirements had been implemented for Unit 2.

Qualified personnel and procedures were available on site to support Unit 2 maintenance activities.

2.2.2 Quality Assurance Program for Design ~ Changes and Modifications (35744)-

The objective of this inspection effort was to determine whether the licensee had developed and implemented a QA program on the control of design changes cod' modifications, that conformed with industry guides and standards, regulatory requirements, and commitments in the FSAR.

2.2.2.1 Discussion

"

The inspectors reviewed Station Administrative Procedure STA-716, " Site-Modification Process," Revision 8.

The program for management of design was well defined and consistent with FSAR commitments to ANSI N45.2.ll-1974 and ANSI N18.7-1976. A site modification could' be accomplished through a design modification, minor modification, setpoint change, facility change, or repair

.

disposition of an Operations Notification and Evaluation (ONE) Form. A facility change was a large scope modification to the site requiring additional administrative management.

In general changes to safety related structures, systems, or components were design modifications. Minor modifications by definition did not require a change to the FSAR, Fire Safe Shutdown Analysis, or an environmental qualification document package.

Changes that were repair dispositions were controlled by Station Administrative Procedure STA-422, " Processing of Operations Notification and Evaluation (ONE) Forms," and Procedure ECE 5.01, " Design Control Program -

,

General."

-4-

., - _ _ _

- -

-

e

.

.

, Station Administrative Procedure STA-716, which was the Unit I design control program has been revised to scope Unit 2 changes after the " Designated Date,"

December 21, 1992.

Changes that were started in the older design change control process were called " Design Change Authorizations."

These changes were completed following Project Procedure, 2PP-5.01,." Procedure for Processing of Design Change Authorizations (DCAs)."

+

Station Administrative Procedure STA-602, " Temporary Modifications,"

Revision 10, was found to be used to manage temporary modifications during-initial startup testing and operation of Unit 2.

Temporary modifications, if not addressed in an approved procedure or design modification, must be documented on a temporary modification form.

In addition, the forms were logged and filed in the control room. Station Administrative Procedure-STA-602 became effective when a system was turned'over to operations. Then any open temporary modifications from earlier activities were closed.

The inspectors requested a search done of the engineering data base and the document control data base for design modifications, minor modifications, and temporary modifications.

It was found that no design modifications had been issued since the " Designated Date" and only 26 minor modifications were issued. None of the minor modifications were completed. Eleven of the minor modifications dealt with chenging valve positions. The inspectors found six temporary modifications were installed which affected five drawings. The inspectors verified the temporary modifications to be on file in the control room and the drawings were marked up to reflect the temporary modification with one exception.

On January 7,1993, the inspectors found that Drawing M2-0249, Sheet 2A,

" Generator Primary Water," was not marked up to reflect the installation on.

December 3,1992, of a sodium hydroxide pump by Temporary Modification; 92-2-0015. This failure to mark up the drawing had been identified earlier by the licensee on an ONE Form 92-1561 dated December. 17, 1992. This form was closed on December 28, 1992, with " ? issue of the information for marking.up the drawing to the control room. An engineering review on January 4,-1992, found the information to mark up this drawing was in the control room, but the drawing was not marked up. The licensee informed _the inspectors that a Unit I drawing (not the affected Unit 2 drawing) had been marked up in error.

The licensee issued an ONE Form 93-065 on this failure to control temporary modifications, which was a violation of Station Administrative Procedure STA-602, paragraph 6.6.16, and similar to the violation identified in Unit I activities (refer to Violation 445/9262-01 which is documented in NRC Inspection Report 50-445/92-62; 50-446/92-62).

2.2.2.2 Conclusions The program to be used for Unit 2 design changes and modifications is the same as that used for Unit 1.

The licensee's program for the implementation of the.

program should be adequate and will be assessed in the future as a part of the routine inspection program.

-5-

_

I

-.

.

2.2.3 Surveillance Testing and Calibration Control (35745)

The purpose of this part of the inspection was to ascertain whether the licensee had implemented programs for control and evaluation of the following:-

Surveillance testing, calibration, and inspection as required by the TS;

Inservice testing (IST) of pumps and valves as described in

,

10 CFR 50.55a.(g); and l

Calibration of safety-related instrumentation not specifically

'

controlled by the TS.

An inspection of the QA program relating to surveillance testing and calibration control activities associated with Unit I had been documented in NRC Inspection Report 50-445/90-12.

Therefore, the scope of this inspection was limited to verifying implementation of the licensee's QA program relating to surveillance testing and calibration control activities for Unit 2.

2.2.3.1 Discussion The inspectors reviewed the draft report, "TS Surveillance Requirements,"

Revision C, to ascertain that requirements were appropriately listed. The inspectors observed that this report contained both surveillance testing, calibration, and inspection required by the final draft version of the TS and IST required by 10 CFR 50.55a(g). To verify the accuracy of this report, the

,

inspectors verified that responsive test procedures were listed for the following TS paragraphs:

4.2.1.1, 4.3.1.1.2a, 4.3.1.1.10, 4.3.1.1.lBa, 4.3.2.1.lc, 4.3.2.1.2a, 4.3.2.1.3a.1, 4.3.2.1.4a, 4.3.2.1.5a, 4.3.2.1.6a, 4.3.2.1.6d, 4.3.2.1.8c, 4.3.2.1.9a, 4.3.2.1.10b, 4.3.2.1.11b, 4.3.3.2.lb.4, 4.3.3.3b.12, 4.3.3.3b.15, 4.7.1.2a.1, and 4.7.4.1.2.

The inspectors selected several of the test procedures which were responsive to the above specifications and verified that they had been scheduled for performance at the assigned frequencies. The licensee's plant group responsible for performance and the current status were indicated on the necessary documentation.

During discussions with the PM planning and scheduling personnel, the inspectors. learned that calibration of safety-related instruments not identified in TS was part of the routine PM program. The calibration requirements were specified in the master PM schedule and included the following:

Calibration frequency for each component;

Licensee plant _ group responsible for performing the calibration; and

Calibration status.

• The inspectors selected a number of Unit 2 flow instruments, which did not have calibration requirements specified by the TS, for determination of

.

calibration status:

2-FT-2466A, 2-FT-4556, 2-FT-4773-1 and -2, 2-FIS-4756,

-6-

>

,

.

.

,2-FIS-0611, and 2-FT-0918. The instruments were included in the Unit 2 PM program and the calibration frequency and status were given.

2.2.3.2 Conclusions.

The licensee had implemented suitable programs for control and evaluation of:

Surveillance testing, calibration, and inspection as required by the TS;

IST of pumps and valves.as described in 10 CFR 50.55a(g); and

Calibration of safety-related instrumentation not specifically

controlled by the TS.

All TS and plant instruments selected by the inspectors for determination of test and calibration status were appropriately covered by the licensee's programs.

2.2.4 Overall Conclusions of Quality Assurance Program A QA program for maintenance that conformed to the appropriate requirements had been implemented for Unit 2.

Qualified personnel and procedures were available to support maintenance activities.

t The licensee's program to be used for Unit 2 design changes and modifications

'

was the same as that used for Unit 1, and its implementation should be adequate.

The licensee had implemented suitable programs for control and evaluation of surveillance testing, calibration, and inspection as required by TS and for calibration of safety-related instrumentation not specifically controlled by TS.

2.3 Procedure Reviews 2.3.1 Reactor Protection System Test - Preoperational Test Procedure Review (70305)

In this area of the inspection, the inspectors evaluated a preoperational test procedure for the reactor protection system time response measurement.

In particular, the inspectors reviewed the procedure for technical and administrative adequacy and consistency with regulatory requirements, guidance, and licensee commitments. The inspectors primarily directed the review toward determining whether this test would satisfy the licensee's FSAR commitments.

2.3.1.1 Olscussion The inspectors reviewed Preoperational Test Procedure 2CP-PT-64-08,

" Reactor Trip System Time Response Summary," Revision 0.

This procedure did not direct the performance of any field testing. The procedure provided a-7-

.

.-

.

, vehicle to collect data from the performance of twenty-eight other tests.

After the data was collected, this procedure then provided for calculating the reactor trip system and engineered safety features total response times.

Acceptance criteria was provided in detail for each functional unit measurement. The' individual tests satisfied the FSAR testing requirements.

No technical inaccuracies were identified.

2.3.1.2 Conclusions The preoperational test procedure appeared to be responsive to the itcensee's needs for collecting, calculating, and presenting the reactor trip system time

,

responses.

'

2.4 Precoerational Test Results Evaluations The inspectors reviewed the licensee's preoperational test results evaluations to assure that the test data were within previously established acceptance

,

criteria, or that deviations were properly dispositioned; evaluated the

'

adequacy of administrative practices in maintaining proper test discipline during test execution, test alteration, and test recording; and verified that the licensee followed its procedures for review, evaluation,'and acceptance of test results.

2.4.1 Containment Testing (63050,70323)

In this area of the inspection, the inspectors reviewed the licensee's results evaluations for three completed presperational tests for the Unit 2 containment building. The first preoperational test was the containment structural integrity test, the second was the containment integrated leakage

,

rate test, and the third was the local leak rate test. The purpose of the tests was to verify that the containment structure met its design structural requirements and also met the leakage requirements of Appendix J to 10 CFR 50.

.

2.4.1.1 Discussion

,

,

The documents reviewed were Project Special Report PTR-043, " Structural

i Integrity Test (SIT) Report," Revision 0, for Project Procedure 2PP-5.27,

" Reactor Containment Structural Integrity Test Procedure," Revision 0, and Engineering Assessment Procedure 2-EAP-036, " Structural Integrity Test:

Containment Attachments'and Crack Mapping," Revision 0, and " Reactor Containment Building Integrated Leakage Rate Test, Unit 2 Final Report," for Preoperational Test Procedure 2CP-PT-75-02, " Containment Integrated Leak Rate Test," Revision 0.

The third test was Preoperational Test Procedure.

2CP-PT-75-01, " Containment Local Leak Rate Tests," Revision 0.

The inspectors noted no deviations to the testing procedure or the acceptance l

criteria during the review of-Project Special Report PTR-043. During the review of " Reactor Containment Building Integrated Leakage Rate Test, Unit 2 Final Report," the inspectors noted that two data points met the data point rejection criteria, but deletion of the data point had no appreciable effect

.

on the final results; therefore, the points were not deleted.. Additionally,

'

one temperature sensor (TE-10) had data spikes of 1-3*F.

Had the sensor been-8-

-

--

-

-

.

.

, declared failed and the volume fractions reassigned, there would have been no

,

appreciable change in the test results; therefore, the sensor was not declared

~

failed.

Preoperational Test Procedure 2CP-PT-75-01 was performed to verify that all

-

Type B and C tests met the leakage criteria of Appendix J to 10 CFR 50 and the personnel and emergency airlock interlock operability. There were seven retests perforned for the valves that initially failed local leak testing.

All retests were satisfactorily performed. Two of the five startup deficiency reports (SDRs) written were to address administrative deficiencies identified during the licensee's results evaluation. All SDRs were properly closed.

Section 6.3.5 of the procedure concerning recording of test activities in the chronological test log, including retest information, test interruptions, test witnessing, test changes and late entries were frequently not met in that entries were incomplete or inadequate. As mentioned in paragraphs 2.4.2.1.1, 2.4.2.2.1, 2.4.3.1, 2.4.4.1, 2.4.5.1, 2.4.6.1, 2.4.7.1, 2.4.8.1, and 2,4.9.1,

'

the licensee's reviewers, during test results evaluations, identified numerous administrative deficiencies that occurred during the performance of the tests.

These preoperational tests were performed in the approximate timeframe of

,

March to December 1992. The types of deficiencies identified by licensee reviewers included the following:

Failure to record test activities in the applicable test logs;

Failure to ensure appropriate procedure step sign-offs were obtained

prior to proceeding to subsequent procedure steps; and Failure to initiate proper deficiency documentation for test

deficiencies.

None of these administrative deficiencies had any adverse effect on the test results, but were considered as part of the hundreds of examples that constitute a violation of 10 CFR 50, Appendix B, Criterion V for failure to follow the requirements specified in Startup Administrative Procedure (SAP)

CP-SAP-078, "Preoperational Testing," (446/9302-01). Additional examples of this violation are discussed in paragraphs noted above.

2.4.1.2 Conclusions All acceptance criteria for all tests were satisfied. The results evaluations

'

for the structural integrity test and containment building integrated leakage rate test (Preoperational Test Procedure 2CP-PT-75-02) were detailed and professionally written. The licensee's results evaluation of Preoperational

>

Test Procedure 2CP-PT-75-01 was thorough and detailed. A violation (446/9302-01) related to STE conformance to administrative requirements was identified.

-9-

.

>

.

..

,2.4.2 Integrated Hot Functional Test (70324)

2.4.2.1 Reactor Coolant System (RCS) Equipment Supports Thermal Expansion Test In this area'of the inspection, the inspectors evaluated the results evaluation of thermal expansion measurements on reactor coolant system (RCS)

equipment and supports that were taken at various temperature and pressure plateaus during the hot functional test.

Preoperational Test Procedure 2CP-PT-55-09, "RCS Equipment Supports Thermal Expansion Test," Revision 0, dated April 17, 1992, was developed to verify expected thermal displacement

' ehavior of reactor coolant loop piping and major nuclear steam supplier o

system (HSSS) components. The results evaluation for the test was approved by the Joint Test Group (JTG) in meeting number 92-17 on November 19, 1992.

,

2.4.2.1.1 Discussion There were six thermal expansion problem reports (TEPRs) generated as a result of measurements and observations made during the test.

Resulting from these reports were 2 SDRs, which required two retests. The results from the retests were satisfactory in that acceptance criteria were achieved.

During the performance of the test, the licensee made two "non-intent" test procedure changes (TPCs). Subsequently, the licensee reviewed the results evaluation and determined that the particular "non-intent" TPCs should have been considered as " intent" TPCs. The licensee's reviewers also determined that the STE who was responsible for the conduct of the test made numerous other administrative errors (i.e., several dozen). These errors related to various administrative requirements specified in Startup Administrative Procedure CP-SAP-07B, "Preoperational Testing." Consequently, a third SDR (3257) was initiated to resolve these errors.

The inspectors reviewed SDR 3257 and the disposition of the licensee's findings. The findings were properly corrected and the STE was counseled on

the requirements of Startup Administrative Procedure CP-SAP-078. The inspectors agreed with the licensee's review that the lack of adherence to the specific administrative requirements did not invalidate any portion of the testing. The failure of the STE to adhere to administrative requirements of

.

Startup Administrative Procedure CP-SAP-078 were further examples of violation (446/9302-01), which is discussed above.

2.4.2.1.2 Conclusions The results evaluation for Preoperational Test Procedure 2CP-PT-55-09, "RCS Equipment Supports Thermal Expansion Test," was acceptable. Numerous errors were made in violation (446/9302-01) of the administrative requirements for s

the handling of the test activities. The quality of the licensee's review process was excellent in that it identified and amended those administrative errors.

-

-10-

.

.

.

,2.4.2.2 Hot functional Piping Systems Themal Expansion Test In this area of the inspection, the inspectors reviewed the results evaluation for Preoperational Test Procedure 2CP-PT-90-03, " Hot Functional Piping Systems Thermal. Expansion Test," Revision 0, which was approved by the JTG in meeting 92-39 on May 28, 1992. This test was developed to identify any piping -

interferences with other components and to measure piping thermal movements at the designated locations and at the specified temperature plateaus.

2.4.2.2.1 Discussion The test was initiated on July 2, 1992, and concluded on October 28, 1992.

Six sections and one section of the test, respectively, were performed in conjunction with the hot functional test and diesel generator test. The results evaluation for the test was approved by the JTG during meeting 92-143 on December 30, 1992.

The test procedure was considered to be comprehensive and it clearly -

delineated test conditions, locations for measuring thermal movements, test

equipment, and acceptance criteria. The following temperature plateaus and

,

monitoring requirements had been established: pre-test ambient; 250*F; 350*F; 450*F; 557*F, and post-test return to ambient. A total of 11 TPCs were initiated, 4 of which were " intent" changes (i.e., a change to the test objectives, acceptance criteria, or test methods). The inspectors verified that the intent TPCs had been approved by the JTG and that the changes were-incorporated into the test procedure. A total of 20 SDRs were initiated to document and effect resolution of deficiencies identified during performance of the test. Fourteen retests were performed as a result of the SDRs, all.

with acceptable results. The balance of the SDRs dealt with administrative type deficiencies. Three TU Evaluation (TUE) forms and 424 TEPRs and I unsatisfactory inspection report were initiated to document nonconforming conditions, interference / obstruction conditions, and an unsatisfactory snubber

!

spring clearance, respectively.

,

The inspectors reviewed the three TUE forms, one inspection report, and a sample of TERPs to verify that appropriate actions had been taken, the conditions resolved, and the reports closed. Subsequent to the completion of the test, the test procedure and results were reviewed by the following

'

groups: Startup Programs; Performance and Test; Unit 2 Operations; Westinghouse Projects; Bechtel Projects, and Quality Assurance. These reviews resulted in the identification of approximately.186 comments / questions (some were redundant due to the multiple reviews). Due to the nature and number of coments, SDR 3519 was initiated on Dece:aber 12, 1992. The review groups coment sheets were attached to and made part of the SDR.

The general description of the deficiency was stated to be a failure by the STEs to adhere to procedural requirements associated with preoperational testing and reporting of deficiencies and nonconformances. The responsible -

'

STE reviewed and responded to all coments.

It was shown that coments/ questions were determined to be either valid or not valid,. with the valid ones being properly resolved. All coments/ questions received an appropriate justification. The end result was that the identified-11-

,

.

.

, deficiencies were determined to be administrative in nature and had no impact on the test results. The SDR was closed on December 31, 1992, after. receiving all the necessary approval signatures. The specified corrective action was to counsel the STE as to the requirements of Startup Administrative. Procedures CP-SAP-07B and CP-SAP-16, " Deficiency and Nonconformance Reporting," and the need for procedural adherence. The training was provided and documented on a Training / Indoctrination Sign-off Sheet which was attached to the SDR. This failure by STEs to adhere to administrative requirements was a further example of a violation (446/9302-01), which is discussed above.

The test procedure results indicated that all monitored systems either met the

,

acceptance criteria or had been reviewed and approved by Engineering.

It was noted that the auxiliary steam system portion of the test was deferred because extraction steam supply could not be established during conduct of the hot functional test. The inspectors verified that the deferral was identified as an open action item (JTG Action Item 1/92-85) and had been incorporated into Initial Startup Procedure ISU-3088, " Thermal Expansion Power Ascension Phase,"

Revision 0, dated November 12, 1992.

Attachment 4 contains a listing of all procedures reviewed in this area of the inspection.

2.4.2.2.2 Conclusions All acceptance criteria for the test had either been met or had been reviewed and approved by Engineering. The test procedure was comprehensive and well defined. Where deficiencies, nonconformances, and problems had been identified, timely reviews and appropriate actions were taken to. resolve the issues.

Numerous procedural administrative errors were identified during the post-test procedure results evaluation. These errors were in violation (446/9302-01). of the licensee's administrative requirements. The quality of the licensee's review process was considered excellent in that it identified and corrected those administrative errors. The deferred portion of the test

<

was identified as an open JTG action item and has been incorporated in another test procedure for subsequent testing.

2.4.3 Reactor Protection System (RPS) (70325)

The inspectors verified that the RPS had been tested as specified by the FSAR and draft TS. The following requirements were of special interest:

,

Verification that each combination of required channel trips will scram

the reacter; Those protection systems which have two or more methods of tripping the

scram breakers function in all possible combinations for each trip mode; and

,

All components are fail safe on loss of power.

..,

-12-

.

..

.

,2.4.3.1 Discussion The inspector determined that the following tests satisfied RPS preoperational testing requirements:

2CP-PT-64-01, "RPS Time Response Measurement," Revision 0;

2CP-PT-64-02, "RPS Operational Check," Revision 1; and

2CP-PT-64-08, "RPS Results," Revision 0.

• Preoperational Test Procedure 2CP-PT-64-01 had no acceptance criteria.

It was a data collection step for Preoperational Test Procedure 2CP-PT-64-08, which verified that the overall system response times met the required acceptance

-

criteria.

The acceptance criteria appeared to be appropriate and have been. satisfied in Preoperational Test Procedure 2CP-PT-64-02. The itcensee's test results evaluation appeared to be detailed and comprehensive. However, during the test results evaluation process, the licensee identified numerous editorial erurs in the procedure and administrative errors made by the STEs. The following deficiency documents were issued: 33 TPCs, 42 SDRs, and one TUE.

In addition, the following were the number of comments generated during the test results evaluation procc;s: QA/ Nuclear Operations Department (N00) - 167, Other QA - 23, Startup Programs - 90, Unit 2 Operations - 103, Scope B I&C - 12, and Unit 2 Engineering - 11. Because of editorial errors, it was difficult to associate the acceptance criteria in Section 2 with the procedural steps in Section 7.

The licensee issued SDR 3585 to resolve this issue. The licensee developed a matrix of all Section 2 acceptance criteria and associated Section 7 performance steps which satisfied those criteria.

The licensee's actions appeared to verify that all acceptance criteria were satisfied. There were no open or outstanding items associated with the test.

SDR 3584 addressed administrative errors made by the STEs during performance of Preoperational Test Procedure 2C7 ?T-64-02. Most of them dealt with violations (446/9302-01) of Startup Administrative Procedures CP-SAP-07B and

- 16. The following numbers of test results evaluation comments were referenced: Startup Programs - 74, QA/ NOD - 55, and Other QA - 1.

Many of these errors made the test data very difficult to interpret. For instance, the STEs routinely signed failed steps contrary to Startup Administrative Procedure CP-SAP-078, which required the SDR number to be entered in lieu of a

-

signature. For example, in Retest 5, Step 7.2.138 was signed with SDR 2993 noted in the margin. The SDR noted that after a wiring error was corrected, the same step should be signed as a retest. Because of the STEs' convention

of signing failed steps, the record.was not clear that the step had actually been signed prior to the retest. A similar situation existed for Step 7.2.161-

'

and SDR 3001 of the same retest. Because there were numerous omissions, the chronological test log was not very useful in clarifying test failures and retests. As a result of test results evaluation comments, several pages of l

late entries were added to the chronological test log.

Because of the high-number of administrative errors made by the STEs, the inspectors concluded

-

that STE training for conducting tests was inadequate. The inspectors also questioned the value of the licensee's QA surveillances and management-13-

?

'

-

~

r

_,

.

, oversight applicable to this test.

If some of these errors had been observed in-process, it would have been appropriate to stop the test and conduct retraining.

The specification of retesting for SDRs appeared weak. Most of the SDRs stated that testing should be performed in accordance with applicable steps of a given retest. Because some of the retests related to several SDRs, it was difficult to determine the retesting that was actually being specified. The inspectors determined that prior supervisory approval was not obtained for all parts of Retest 10 to Preoperational Test Procedure 2CP-PT-64-02 as required by paragraph 6.8.1.8 of Startup Administrative Procedure CP-SAP-078,

"Preoperational Testing," Revision 3.

Specifically, SDR 2974, which identified several relay contacts as being reversed, was initiated on September 14, 1992. SDR 3033, which identified a discrepancy in the normal configuration of contact blocks, was later initiated on September 24, 1992.

Retest 10 was referenced on each SDR as the required retest by the STE on October 2, 1992, and was approved by the test group supervisor (TGS) on the same date. However, the inspectors noted that the completed steps in Retest 10 were all dated September 30, 1992, which was two days before Retest 10 was approved by the TGS. This is another example of Violation 446/9302-01.

A substantial number of the SDRs related to configuration control problems.

Many SDRs related to electrical cabinet and breaker internal wiring discrepancies.

Several SDRs related to the relay contact blocks not being installed as shown on the drawings.

SDR 2995 involved the closure limit switch actuation arms being left off the high pressure stop valves.

SDR 3018 involved a drawing not showing that fuses were bypassed. SDR 3072 related to a computer database error. SDRs 3119 and 3201 involved interlock screws in the reactor trip breakers and reactor trip bypass breakers, respectively, which should have been removed. The inspectors identified the following SDR numbers as relating to configuration control problems: 2958,.2960, 2967, 2974, 2983, 2993, 2995, 3010, 3018, 3025, 3033, 3047, 3050, 3068, 3072, 3118, 3119, and 3201. All SDRs related to Preoperational Test Procedure 2CP-PT-64-02 were closed and it appeared that safety issues had been appropriately addressed. However, the inspectors questioned the effectiveness of the licensee's component testing program in identifying. errors such as these. The configuration control issue will be pursued as an inspection followup item (446/9302-03) during the' subsequent routine inspection program.

The inspectors reviewed the JTG-approved test results evaluation for Preoperational Test Procedure 2CP-PT-64-08, Revision 1. -The stated test objectives were:

To calculate the reactor trip system total response times and verify

these times are in accordance with TS requirements; and To calculate the engineered safety features total response times and

verify these times are in accordance with TS requirements.

-14-

r

.

.-

,This procedure does not direct the performance of any field testing. The procedure provides a vehicle to collect data from the performance of 28 other tests. The inspectors noted during the review that all SDRs and TUEs had been properly closed and ONE Form 93-0071 remained open. One IPC was an intent change and was properly reviewed and approved. The change had no effect on

,

the testing acceptance criteria. All licensee reviewer comments were properly

!

evaluated and resolved. The licensee's review of this test was thorough and comprehensive. All test acceptance criteria were met except as noted below.

SDR 3448 resulted in a deferred test item. This item is tracked by JTG Action Items 2/92-137 and 2/93-10, and is included in the deferred test item letter.

The deficiency involved the damaged strip recorder trace of Valve 2-HV-2334A and affects Data Sheet 9.4, Items 464, 471, 481, 486, 493, 503, 512, and 522.

ONE Form 93-0071 remains open pending the completion and acceptance of a TUE that will provide for accepting the containment spray pumps data that exceeded

-

the acceptance criteria. This item is being tracked by JTG Action Item 3/93-10.

2.4.3.2 Conclusions for the RPS, all testing was completed and acceptance criteria were satisfieo with two exceptions:

One deferred test which was being tracked by a JTG action item; and

ONE Form 93-0071 was being tracked by a JTG Action Item 3/93-10.

• The test results evaluations were detailed and comprehensive.

However, they raised several issues of concern to the inspectors. One issue was apparent ineffective training of the STEs on administrative requirements for

,

preoperational test performance. This is a further exarnple of Violation 446/9302-01. A related issue was the apparent failure to identify these administrative errors during QA and supervisory surveillance of the test activities. There were so many errors in the performance of Preoperational Test Procedure 2CP-PT-64-02 that it was not clear why testing was not stopped and remedial training provided for the STEs. The significant number of configuration control problems identified by Preoperational Test Procedure 2CP-PT-64-02 is an NRC concern and is being tracked as an inspection followup item (446/9302-03). Specification of retesting in the SDRs was weak.

Examples of a violation involving performance of a retest prior to obtaining approval of the retest procedure were identified (446/9302-01).

2.4.4 Loss of Offsite Power Test (70326)

In this area of the inspection, the inspectors reviewed two completed preoperational test result evaluations for the loss of offsite power test.

One preoperational test was to verify that the emergency diesel generators would be operational during loss of offsite power and the other was to verify the operation of the solid state sequencer under the same conditions.

,

'

-15-

..

.

,2.4.4.1 Discussion The purpose of the tests was to verify that the train A and B diesel generator and the solid state sequencer would meet their design and performance requirements under the following conditions:

Loss of offsite power with no safety injection;

Safety injection with no loss of offsite power;

Loss of offsite power coincident with safety injection; and

Safety injection followed by a loss of offsite power.

'

The two completed preoperational test procedures reviewed were:

2CP-PT-57-05, " Integrated S/G Actuation Test," Revision 1; and

2CP-PT-64-07, " Solid State Safeguards Sequencer, Revision 1."

• Preoperational Test Procedure 2CP-PT-57-05, was included in two other

'

functional groups of the preoperational testing program. The NRC review of the licensee's results evaluation is documented in the preoperational testing functional group titled " Engineered Safety Features Actuation System (ESFAS)

Tests" in section 2.4.7 of this report. The results evaluation review of Preoperational Test Procedure 2CP-PT-64-07 is documented below.

The purpose of this test was to demonstrate the logic of the Train A and B solid state safeguards sequencer and auxiliary relays would function as designed. The specific test objectives were:

Verification of the sequencer input and power supply output voltages,

and output relay coil open circuit detection; Verification of the correct contact position for each blackout signal

and safety injection signal auxiliary; Verification of automatic and operator lockouts during various simulated

input signals; Verification of all associated sequencer blocks, block override and

priority override functions during all testing and simulated valid input

,

conditions; and Verification of load timing sequences during simulated safety injection

and blackout sequences.

This preoperational test was a relatively complicated test due to the number of output devices, signals generated, and the various input conditions. The

,

inspectors noted in the licensee's results evaluation that there were various interruptions during the test due to test equipment, procedural, and plant equipment failures. These problems were properly documented, evaluated and resolved during the test performance. Two SDRs-were written for various administrative procedural errors made by STEs. This is another example of-16-

-

.

-.

,

.

+

, Violation 446/9302-01. All identified deficiencies, test procedural changes and reviewer comments were properly closed out.

2.4.4.2 Conclusions The test package was complete and all acceptance criteria were met. Several of the deficiencies identified during the licensee's results evaluation appeared to have arose because of a lack of attention to detail during the-preoperational test performance documentation. The overall conclusion drawn by the inspectors was that although the test was technically correct.it lacked in quality of documentation, which was a further example of Violation

'

446/9302-01.

2.4.5 Engineered Safety Features (ESF) Tests (70400)

In this area of the inspection, the inspectors reviewed the results evaluation for performance and flow balance test for the centrifugal charging pump Preoperational Test Procedure 2CP-PT-57-02, "CCP Performance and Flow Balance Test," Revision 1.

This test was responsive to ESF testing requirements.

Other preoperational tests and the NRC inspection report that discussed NRC reviews are as follows:

2CP-PT-57-01 in NRC Inspection Report 50-446/92-48;

2CP-PT-57-03 in NRC Inspection Report 50-446/92-38;

2CP-PT-57-04 in NRC Inspection Report 50-446/92-48

2CP-PT-57-05 in paragraph 2.4.7 of this report; and

2CP-PT-57-06 in NRC Inspection Report 50-446/92-48.

• 2.4.5.1 Discussion During the test, the following change and deficiency documents had been issued:

14 TPCs, 12 SDRs, and 4 TUEs. Only one of the TPCs was an intent change. However, it appeared that the number of changes necessary was-excessive. Only one of the SDRs and TUEs impacted acceptance criteria.

TUE 92-4762 provided the disposition for the total developed head for component cooling Pump.2-01 exceeding the maximum allowable. The use-as-is-disposition was justified by the fact that the total injected flow rate during a postulated steam generator tube rupture' accident would be less than the maximum allowed due to system resistances being higher than assumed.

TUE 92-5052 was issued for failure of the STE to issue an SDR or a TPC when Procedure Step 7.3.45 could not be satisfied.

It also referred to a-similar performance failure in Preoperational Test Procedure 2CP-PT-58-01, " Residual-Heat Removal System," Revision 2, Step 7.3.45 involved verifying that.an annunciator window was clear. The STE initiated a temporary modification without first initiating deficiency documentation as-required by preoperational test administrative procedures. -SDR 3131 was issued to address administrative deficiencies in test performance and it also referenced 45 test-review comments. The large number of performance errors by the STEs suggested-inadequate training in preoperational test administrative requirements. These errors were further examples of Violation 446/9302-01.

-17-

,

!

.

.

It appeared that all test deficiencies had been suitably dispositioned and the

.

test results evaluation was complete. The licensee was advised of minor chronological test log errors. identified by the inspectors.

.

2.4.5.2 Conclusions Preoperational Test Procedure 2CP-PT-57-02 had been satisfactorily completed and all deficiencies had been suitably dispositioned. The licensee identified numerous administrative errors in test performance, which suggested that-training for the STEs had been deficient. These errors were further examples of Violation 446/9302-01.

,

,

2.4.6 - Emergency-Standby Power Supply System Test (70400)

i In this area of the inspection, the inspectors reviewed four completed preoperational tests for the emergency-standby power system. Two tests were for the Train A and B diesel generators. One test involved the main contro)

-

room cmergency DC lighting, and the remaining test involved the AC essential lighting. The inspectors reviewed the tests to verify that the tests had met

-

all testing acceptance criteria, that all deviations were properly resolved, that the licensee's results evaluation had been properly conducted, that the tests had been conducted in accordance with the approved administrative procedures, and that any required retests had been performed and evaluated.

2.4.6.1 Discussion Th; purpose of the first two tests was to verify that the Train A and B emergency diesel generators would meet their design and performance requirements as follows:

Proper operation of the engine /gererator controls;

Proper performance of the diesel engine fuel oil storage, transfer and

supply systems; the jacket water cooling and keep warm systeras; and the engine lubricating and air start systems; Diesel generator full load capability and-full load rejection;

Diesel generator 35 start reliability and five start air capacity

verification; Diesel generator space heater SIS /AL load shed; and

Diesel generator room ventilation temperature control.

• The two completed results evaluations reviewed were for Preoperational Tests Procedures:

2CP-PT-30-01A, " Emergency Diesel Generator " Train A," Revision 2; and

!

2CP-PT-30-OlB, " Emergency Diesel Generator " Train B," Revision 1.

-l

!

-18-i

.

.

The results evaluations contained, in addition to the completed preoperational test procedure, the preoperational test review and comment forms, the chronological log, the test review checklists, test procedure change documents, startup deficiency reports, the test review report,.and attachments.

The results evaluations appeared complete and all acceptance criteria were

'

met. All testing discrepancies were identified and appropriately dispositioned by issuance of either SDRs or TUEs. The number of SDRs and TUEs did not seem excessive and had no adverse effect on the tests. The comments on the test review forms appeared to indicate that the quality of the test

,

documentation in the area of the chronological log keeping was poor. The licensee's results evaluation appeared to be complete and detailed.

The results evaluations for the following two Preoperational Tests Procedures verified the. acceptability of AC essential and main control room DC emergency lighting:

2CP-PT-71-01, "AC Essential Lighting Test," Revision 0; and

2CP-PT-71-04, " Main Control Room Emergency DC' Lighting," Revision 1.

• The tests were complete and satisfied the review criteria discussed above.

All design requirements appear to have been satisfied. There were 7 TPCs for

,

Preoperational Test Procedure 2CP-PT-71-01, of which 3 were intent changes.

The number of changes appeared high. There were 6 SDRs and no TUEs. SDR 3600-i cited administrative errors made by the STEs involving 13 test review comments. They involved performance and documentation errors, but did not impact the test results. This large number of performance errors suggests the STEs may not have received apprcpriate training on test performance administrative requirements. These were further examples of Violation 446/9302-01. The licensee was advised of a minor typographical error in the test review report.

For Preoperational Test Procedure 2CP-PT-71-04, there were 5 TPCs, 5 SDRs, and no TUEs. None of the SDRs affected acceptar..

,

criteria, but 2 required retests.

SDR 3644 cited 3 performance errors by the STEs.

2.4.6.2 Conclusions The results evaluations reviewed were complete and all acceptance criteria were met. Several of the deficiencies identified during the licensee's results evaluations appeared to have been the result of a lack of attention to detail during the preoperational test performance, which implied a possible training deficiency and were further examples of Violation 446/9302-01. The.

inspectors' overall conclusion was that although the test results evaluations

.i were technically correct, an unusually high number of_ performance errors were..

-

made by the STEs during performance of some of the preoperational tests and is a further example of the violation discussed above.

-19-

'

.

.

. 2.4.7 Engineered Safety Features Actuation System (ESFAS) Test (70400)

In this area of the inspection, the inspectors reviewed three completed and

JTG-approved preoperational test results evaluations for the engineered safety system. The purpose of the tests was to verify that the engineered safety systems would actuate under various initial conditions in sequence as designed.

'

2.4.7.1 Discussion The inspectors reviewed the licensee's test results evaluations for the

'

following Preoperational Test Procedures:

2CP-PT-57-05, " Integrated Safeguards Actuation Test," Revision 1;

2CP-PT-64-04, " Safeguards Test Cabinets Direct Actuation

Preoperational Test," Revision 1; and 2CP-PT-64-05, " Safeguards Test Cabinets Blocking Circuits

+

Preoperational Test," Revision 1.

For Preoperational Test Procedure 2CP-PT-57-05, deficiency documents (status indicated) were issued as follows:

87 SDRs - none open;

22 TUEs - five open; and

4 ONEs - three open.

• The inspectors perceived that the number of deficiencies were higher than should be expected and interviewed licensee startup personnel to gain a better.

perspective. The inspectors were particularly concerned about the component failures. This could reflect negatively on the preventive maintenance and'

component testing programs. A licensee representative stated that the component failures were undesirable, but not unusually high.

It was also stated that there was an approximately 40 percent failure rate during component testing a year ago. Most of these components were subsequently tested during previous preoperational tests and the failure rate was _much lower.

In response to the inspectors' concerns, the licensee compiled-a list of all component failures with the number of failures for each component.

There were 11 components having' failures on this list.

Because some components had multiple failures, there'were 18 occurrences of. component failures. All components had tested satisfactorily during one of more of the nine test sequences. The inspectors determined that most of the component failures would not have a major impact on plant safety if. they occurred during power operations.

The following deficiency documents remained open: TUE-92-6916. TUE-92-6955, TUE-92-6958, TUE 92-691, TUE 92-6922, ONE 92-1400, ONE 92-1450, ONE 92-1511, and Work Request-AR 92-032466 which resulted from SDR-3371. Work Request AR 92-032466 was being tracked by JTG Action Item 3/93-5. The licensee stated that the ONE and TUE Forms will be transferred to the TRG for tracking and-20-

.

f

..

, that they and JTG Action Item 3/93-5 will be closed prior to plant startup.

It appeared that with closure of these items all test acceptance criteria would be satisfied.

The inspectors reviewed the JTG-approved test results evaluation for Preoperational Test Procedure 2CP-PT-64-04, Revision 1.

The stated test objectives were:

To verify the test status indication and interlocks prevent both trains

or one train and one panel (I and II) from being in test simultaneously; To verify that safeguards test panel II test switches actuate only their

associated slave relays; To verify the function of the latching type relays and the ability of

reset switch S920 to reset these relays; and To verify the interlocks associated with K643, K644, K645, and 741

prevent system misalignment during testing.

The inspectors noted during the review that all SDRs and TUEs had been properly closed. One TPC was en intent change and was properly reviewed and approved. The change had no effect on the testing acceptance criteria. All-licensee reviewer comments were properly evaluated and resolved. There did not appear to have been any major equipment problems encountered during the tests. The licensee's evaluation of this test was thorough.and comprehensive.

All test acceptance criteria were met.

The inspectors reviewed the JTG-approved test results package for Preoperational Test Procedure 2CP-PT-64-05, Revision 1.

The stated test objectives were:

To verify that test status indi~; tion and interlocks prevent both trains

'

or one train and both panels (I and II) from being in the test mode simultaneously; To verify the function of the latching type slave relays and the' relay

reset capability of Reset Switch 2-5821; To verify that the panel I test switches actuate only their associated

slave relays and actuate their associated blocking circuits / relays to -

prevent field actuation while testing; To verify the function of the to-Lo Tavg steam dump interlock lamps and

feedwater split flow bypass valves blocking relay logic; and

,

To verify that the turbine' trip test status indication and interlocks

prevent both trains from being in the test mode simultaneously, and actuate their blocking circuits / relays to prevent field ' actuation while testing.

-21-

r

.

..

f

.

,The inspectors noted during the review that all SDRs and TUEs had been properly closed. One TPC was an intent change and was properly. reviewed and approved. The change had no effect on the testing acceptance criteria. One SDR was written to document that the reviewers identified administrative errors made by the STE in the conduct of the test. This was another example of Violation 446/9302-01. All licensee reviewer comments were properly

,

'

evaluated and resolved. All testing acceptance criteria, except as noted below, were met. The licensee's result evaluation was thorough.and comprehensive.

There were 18 test circuits per train that will be spared by design changes (DCAs) 103759, 103760, 103761, and 103762. The circuits being spared are the

.

'

anti-feedwater hammer test circuits. These circuits had been spared in

.

!

Unit 1.

Due to DCA work in progress, three relays could not be tested. The licensee performed a technical review and determined that the testing of these relays could be deferred and accomplished by the performance of Operations Surveillance Procedure OPT-406B, " Safeguards Slave Relay With Blocking Test,"

Revision 0.

This item was reviewed and approved by the JTG and will be tracked by JTG Action Item 1/93-5.

2.4.7.2 Conclusions The licensee's results evaluation of these three tests were detailed and-complete. All DCAs were properly reviewed and dispositioned. Open ONE and TUE forms were to be tracked by the TRO and closed prior to plant startup.

JTG Action Items 1/93-5 and 3/93-5 were to be closed prior to startup.

Deferred testing was assigned tracking numbers and the method of disposition identified. With the exception of the above open items, which were being tracked by the licensee, it appeared that all ESFAS test acceptance criteria were satisfied. However, as with previous results reviews, an SDR was written to identify and correct administrative errors that occurred during test performance. This was another exan.ple of Violation 446/9302-01.

2.4.8 Main Steam Isolation Valve Test (70400)

In this area of the inspection, the inspectors reviewed the completed and JTG-approved results evaluation for the main steam isolation valves (MSIVs). The purpose of the tests was to verify that the MSIVs would operate in the normal and automatic isolation mode.

2.4.8.1 Discussion The inspectors reviewed the JTG-approved results evaluation for Preoperational Test Procedure 2CP-PT-34-01, " Main Steam Isolation Valve Test," Revision 0.

The stated test objectives were:

To verify the control room and remote shutdown panel operation of all

four MSIVs; To verify that the MSIVs would close upon receipt of an isolation

signal within the maximum time allowed; and

-22-

.

.

To verify MSIV closure upon receipt of an isolation signal after having

,

been isolated from the instrument air system for a minimum of 30 minutes.

The inspectors noted in the review of the results evaluation that the accertance criteria for all tests were met with the exception of MSIV 2-HV-2334A. The inspectors reviewed SDR 3448 and noted that the closure time for this valve was verified to be 3.39 seconds but the recorder trace was damaged during the assembly of the results evaluation and could not be used as a test record. The licensee's resolution of this SDR was to document the retest as a JTG Action Item 2/92-137 to regenerate the test record. The

,

retest is to be performed during Operations Surveillance Procedure OPT-5098, e

"Section XI Testing of Unit 2 MSIVs," Revision 0, which is further described in Attachment 2 to this report. All reviewers' comments were properly evaluated and resolved. There were several administrative errors made by the STE during the conduct of the test that were subsequently identified during the licensee's results evaluation process. One SDR was written to address these errors. The administrative errors made during this test are considered

,

further examples of Violation 446/9302-01.

2.4.8.2 Conclusions All test acceptance criteria, except as mentioned above, were met. The licensee's results evaluation was detailed and thorough. All SDRs were properly closed. The deferral of the retest is considered acceptable (see Section 2.1 of this report). The administrative errors made during this test are considered further examples of Violation 446/9302-01.

2.4.9 Auxiliary Feedwater System Test (70400)

In this area of the inspection, the inspectors reviewed the completed and JTG-approved results evaluation for the auxiliary feedwater system. The purpose of the preoperational test was to verify that the auxiliary feedwater system turbine-driven and two motor-driven auxiliary feedwater pumps, associated valves, and controls would supply feedwater to the steam generators and maintain steam generator water inventory.

2.4.9.1 Discussion

<

The inspectors reviewed the JTG-approved results evaluations for Preoperational Tests Procedures 2CP-PT-37-01, " Auxiliary Feedwater System,"

Revision 0, and 2CP-PT-37-03, " Auxiliary Feedwater Turbine Driven Pump,"

Revision 0.

The stated test objectives were:

To verify the proper operation of instrumentation, controls and

interlocks; To verify the proper operation of system control and isolation valves;

To verify the capability of the station service water system to supply

,

water to the auxiliary feedwater suction header;

-23-

.

.

To verify that the flow limiter will prevent pump runout;

,

To verify that the auxiliary feedwater pumps hydraulic performance meets

design requirements;

,

To verify that the auxiliary feedwater pumps are capable of delivering

water to the steam generators within the acceptable time limits after the initiating signal; To verify the system's ability to restore steam generator water level

after a low-level transient without causing unacceptable feedwater/ steam

'

generator water hammer; To verify that the capacity of the air accumulators for the steam

generator flow control valves is adequate; and To demonstrate the endurance of the auxiliary feedwater pumps by.

• performing a 48-hour endurance run followed by a cooldown, restart and

+

run for a minimum of one hour.

(The turbine driven pump will be run at design conditions for two hours without forced ventilation.)

The inspectors noted during the review that all SDRs and TUEs had been properly closed. Two of the TPCs in each test were intent changes and were properly reviewed and approved. The changes had no effect on the testing acceptance criteria. All licensee reviewer comments were properly evaluated and resolved. There did not appear to have been any major equipment problems encountered during the tests. The licensee's results evaluations were detailed and thorough.

SDRs addressed test performance administrative errors made by the STE during the conduct of the tests. These errors are considered further examples of Violation 446/9302-01.

2.4.9.2 Conclusions All test acceptance criteria for Preoperational Test Procedures 2CP-PT-37-01 and 2CP-PT-37-03 were met. The licensee's results evaluations were detailed

,

and thorough. STE errors made during the tests are considered further examples of Violation 446/9302-01.

2.4.10 Overall Conclusions of Preoperational Test Results Evaluations

'

The licensee's acceptance criteria were met, or deviations were properly dispositioned, for all 20 preoperational tests that were inspected. The licensee's results evaluations of the preoperational tests were detailed, comprehensive, and generally of excellent quality. The licensee identified numerous (i.e., several hundred) administrative errors made during the conduct of the tests by involved STEs. The licensee did not identify two

.

preoperational retests that were performed prior to receiving management approval. These administrative errors were due to apparent lack of training

on administrative requirements for preoperational test performance and lack of l

supervisory or quality control oversight. These instances of a failure to.

,

-24-r

!

..

-.

comply with administrative requirements are considered a violation of 10 CFR 50, Appendix B, Criterion V (446/9302-01).

2.5 Testino Pioina Supoorts and Restraint Systems (70370)

The objective of this inspection was to determine that the licensee has established adequate programs and procedures pertaining to the examination and testing of piping support and restraint systems, and that proper installation and operation of pipe and component support and restraint systems were in accordance with their approved programs, procedures and regulatory requirements.

,

2.5.1 Preoperational Test Results 2.5.1.1 Discussion The inspectors reviewed results evaluations as they related to pipe supports

,

and restraints from the thermal expansion testing, dynamic transient response testing, and the steady state vibration monitoring performed by the licensee.

Some calculations were reviewed for steady state vibrations in which data.did

'

not meet the required acceptance criteria; however, Engineering performed calculations to justify use-as-is. Visual inspections of the tests performed during HFT were documented in NRC Inspection Reports 50-445/92-12; 50-446/92-12 and 50-445/92-27; 50-446/92-27.

In conjunction with the raiew of the preoperational test results evaluations, the inspectors also evaluated the programmatic aspects of the testing and examination of safety-related piping supports and restraints by review of site procedures. This evaluation included the review of the following Project Procedures:

2PP-5.09, " Interface Between Backfit And Other Organizations For Pipe

Stress Analysis And Pipe Support Design Activities," Revision 1; 2PP-5.23, " Piping Vibration Test Guidelines," Revision 0; and

2PP-5.25, " Piping Thermal Growth Test Guideline," Revision 0.

• The inspectors concluded as a result of this review that the licensee had established appropriate controls for the testing of pipe supports and restraint systems during HFT. Other procedures reviewed during this portion-of the inspection are listed in Attachment 3 to this report.

2.5.1.1.1 Thermal Expansion Test The inspectors reviewed the RCS equipment supports thermal expansion test

procedure and test results.

Preoperational Test' Procedure 2CP-PT-55-09, "RCS Equipment Supports Thermal Expansion Test," provided guidelines for obtaining measurements of the RCS equipment support gaps prior to heat-up, during plant heat-up and after cool-down in order to verify the expected thermal displacement behavior of the reactor coolant loop and major nuclear steam-25-

.

..

. supply system (NSSS) components in accordance with guidance in NRC Regulatory Guide 1.68, " Initial Test Programs for Water-Cooled Nuclear Power Plants."

Review of the test data indicated that test results evaluations were acceptable and met the acceptance criteria, or had been reviewed and approved by Westinghouse Engineering.

2.5.1.1.2 Dynamic Equipment Transient Response Testing The inspectors reviewed Preoperational Test Procedure 2CP-PT-90-02, " Dynamic Transient Response Testing," and results evaluations. This procedure monitored load sensing clevis pins installed in rear brackets of selected snubbers and struts to demonstrate that the selected points responded to the transient events in accordance with design. Test records indicated that'some test loads exceed the calculated loads.

Exception reports were written for these conditions and evaluated by engineering.

Engineering concluded by evaluation that the test results for those locations were acceptable.

2.5.1.1.3 Steady-State Vibration Monitoring Test The inspectors reviewed Preoperational Test Procedure 2CP-PT-90-01, " Steady-State Vibration Monitoring Test," and test results evaluations. This test procedure demonstrated that steady state flow piping vibrations of selected ASME Code Class 1, 2, and 3 systems, high energy piping _ systems inside Seismic Category I structures, and high energy portions of systems whose failure could reduce the functioning of any Seismic Category I plant feature to an unacceptable level were acceptable.

Review of the results evaluations indicated that some velocities did exceed the acceptance criteria. Most were evaluated and accepted-by _ Engineering, and others required modifications such as adding additional pipe supports.

2.5.1.2 Conclusions Review of the results evaluations by the inspectors indicated.that required data was documented. All systems tested met design requirements, or were evaluated by engineering and found to satisfy design requirements, or were modified, retested and re-evaluated by Engineering and found to satisfy design requirements. F.eview by the inspectors determined that results evaluations, calculations and Engineering evaluations were maintained and properly reviewed and evaluated.

2.5.2 As-Built Observations 2.5.2.1 Discussion As stated previously, visual observations at ambient, intermediate, and operating conditions to determine thermal expansion and vibration were witnessed by the inspectors during HFT as documented in NRC Inspection Reports 50-445/92-12; 50-446/92-12 and 50-445/92-27; 50-446/92-27. This visual inspection was performed after HFT on various piping systems which

.

-26-

.

'

.

, included dynamic, fixed, and component piping supports at ambient temperature to verify the following:

'

Deterioration, corrosion, deformation or any physical damage was not

evident;

'

All required bolts, locking devices, nuts, and washers were

installed; Pipes, supports, or other associated equipment or components were not in.

• contact with other surfaces as a result of previous vibration or thermal expansion testing;

,

Snubbers and spricq hangers were in their predicted positions; and

Connecting.foints, moving parts, and piston shafts were free from arc

strikes, weld spatter, paint or any other material that would obstruct'

proper operation of the support.

The inspectors examined approximately 47 supports from various systems which are listed in Attachment 3 to this report. Some of the systems were subjected to transient testing. The inspectors verified by visual examination that the above items were not evident. There was one deficiency identified. Support RC-2-Il5-433-C56R was missing a cotter pin. The licensee initiated ONE Form 93-140 to correct this deficiency. This deficiency was considered by the inspectors to be an isolated case.

2.5.2.2 Conclusions In conclusion, the pipe supports examined were in satisfactory condition.

There were no inspection activities by the licensee pertaining to pipe supports in progress because all work activities were ~ essentially complete.

2.5.3 Inservice Inspection (ISI) Program 2.5.3.1 Discussion The inspectors did not review the licensee's ISI program for Unit 2 because

,

this program was not in place at the time of this inspection. -It was-also noted by the inspectors that the licensee did not perform a creservice inspection. The licensee had submitted Relief Request NO. F-I to the NRC staff. The inspectors reviewed the relief request submitted by-the licensee.

and the approval by the NRC staff.

L The inspectors were informed by the licensee.that the Unit 2 ISI program would be identical to the Unit 1 program. The inspectors performed an abbreviated review of the Unit 1 ISI program to obtain an understanding of what the Unit 2 program would involve. The inspectors noted that the licensee's augmented ISI program. pertaining to_ snubbers was removed from TS and placed into the

'

Technical Requirements Manual (TRM), Section 3.1.

Station Administration Procedure STA-742, " Snubber Surveillance Program," Revision 4, will provide-27-

.

- - -

-

- -

- -

- - -

-

.

directions and instructions to control the snubber surveillance program for

,

both units as required by the TRM.

The inspectors noted in Section 3.1 of the TRM that no time interval for the first inservice visual inspection of snubbers was specified as in the Westinghouse Standard Technical Specifications for Unit 1.

Apparently this requirement was removed from the previous revision of the TRM and replaced by

.

'

a statement that was not clear. The licensee has since issued a licensing Document Change Request TR-93-003 to replace the time interval requirement.

2.5.3.2 Conclusions The inspectors could not verify the adequacy of the Unit 2 ISI program because the program will not become effective until 4 to 6 months after receiving an operating license.

2.5.4 Overall Conclusions of Testing Piping Supports and Restraint Systems The licensee had properly maintained data and documentation related to the results evaluations of tests of piping supports and restraint systems..These systems were found acceptable, or were evaluated, or were modified, and found to satisfy design requirements. Pipe supports that were examined were in satisfactory condition, and activities in progress pertaining to pipe supports were essentially complete. The inspectors could not verify the adequacy of the Unit 2 ISI program, which is under development.

3 STARTUP PROGRAM 3.1 Overall Startuo Test Prooram (72400)

The inspectors selected and reviewed five licensee administrative controls for assuring that the startup program would be prepared, performed, and evaluated in accordance with regulatory requirements. Many of the. licensee's controls, procedures, and programs over the Unit I operations will be applicable to

- Unit 2 startup and power ascension activities upon Unit 2 license issuance.

Most of these controls and programs were not reviewed during this inspection; however, several procedures were reviewed as discussed below.

3.1.1 Discussion The licensee's methods for implementing the Initial Startup Test and Power Ascension Programs are described in Station Administrative Procedure STA-801,

" Initial Startup Program," Revision 7.

The procedure assigns responsibilities and specifies various controls for the management, planning, scheduling, and coordination of plant activities. The procedure also denotes various lower-tier procedures to be used for conducting certain activities. The procedure

'

appeared acceptable to satisfy its intended use.

The licensee's methods for specifying the format and content of the Initial Startup Test and Power Ascension Programs is given in Station Administrative Procedure STA-816, " Format and Content of Initial Startup Test Procedures,"

Revision 3.

Paragraph 6.13.6 of the procedure states the following:

I-28-

- - _ _.

.

.

" Instruction steps referencing another procedure should specify the

procedure number and title, and a description of the evolution to be performed. Wording of steps should avoid referencing specific step numbers of the same or other procedures."

As indicated in several sections below, the licensee's procedures did not always satisfy the intent of these above-stated requirements.

Specifically, various test procedures did not provide a clear description of the evolution to be. performed in the lower-tier procedures that were referenced (see l

Violation 446/9302-01). This procedure, however, appeared to be an-acceptable j

procedure for establishing and controlling fomat and content of test i

,

procedures.

l l

The licensee's procedure that specified the process for changing, revising, l

"

and approving procedures and test results was Station _ Administrative Procedure STA-817, " Review, Approval, Revision of-and Changes-to Initial Startup Tests and Results," Revision 2.

The procedure provided detailed guidance for the preparation of test data packages; review by test engineers, TRG, the Site Operations Review Committee (SORC), and the plant manager. The procedure established an acceptable means for the necessary specifications.and guidance, which were detailed and easy to follow.

The inspectors reviewed SORC-approved Station Administrative Procedure STA-818, " Conduct of Initial. Startup. Testing," Revision 1.

This procedure-established requirements and provided guidance for the conduct of initial startup testing. This procedure described the responsibilities of testing personnel and the responsible manager. The instructions section discusses plant :ontrol authorities, test log content, plant equipment operation, procedare data and step sign-off instructions, entry correction, pre-test requirements, test performance requirements, test terminations,_ test discrepancy reporting, retesting, notification of test completion, post-test requirements, instrumentation, and recopying of test data.

The inspectors noted during the procedure review that a great deal of detail was included in the instruction section.

For example, specific authorities for plant control-and testing personnel qualifications were discussed in detail. Test step performance sequence was defined and the conditions for which that sequence may be altered was specified. Additionally, test

.

.

termination authority and conditions for which a test can be terminated were-discussed in detail.

Since initial'startup testing is performed after issuance of the operating license, the procedures stress operation within the boundaries of the operating license, TS, and normal plant administrative and-

,

operating procedures.

The inspectors reviewed Administrative Procedure TRA-312, " Performance & Test:

Qualifications," Revision O.

This review determined that the procedure i

h applied to all non-clerical startup test personnel, including direct hires,

_

-

L contractors, contract personnel, vendor representatives and trainees performing startup test functions. The procedure established and defines three levels of qualification. Level I was defined as_ capable of performing the tests. Level II was defined as having Level I capabilities and additionally capable of planning tests, setting up equipment, supervising or-29-

.

-_ _

-

.

,

. maintaining surveillance over the tests, supervising and certifying lower

,

level personnel, and reporting and evaluating test results.

Level III was

defined as having Level II capabilities and also capable of. evaluating the j

adequacy of specific programs used to train and evaluate test personnel. The stipulated education levels and experience requirements are consistent with the requirements of ANSI N18.1, " Selection and Training of Nuclear Power Plant Personnel."

The inspectors reviewed the preliminary roster of test engineering personnel designated for the Unit 2 startup test program. This list was comprised of 30 individuals,10 of these were not certified to any level of qualification,16

.

,

were identified as Level II qualified and 4 were identified as Level. III i

qualified.

In a discussion with the appropriate manager, the inspectors were informed that the 10 unqualified individuals would be qualified prior to startup testing or they would not be used for these startup activities.

The inspectors selected 8 of the 20 qualified personnel and reviewed the individual qualification packages. The individuals were found to be properly qualified for initial startup testing in accordance with Administrative Procedure TRA-312.

The inspectors noted that the assigned responsibilities and organizations specified in each of these procedures had not been updated to reflect the licensee's current organization.

>

3.1.2 Conclusions

Station Administrative Procedures STA-801, -816, -817, and -818, were well written. The procedures contained explicit and detailed instructions as to the program contents, procedure formats, process approvals, and conduct of

,

initial startup testing. These procedures should provide positive and comprehensive control of the initial startup testing program.

Administrative Procedure TRA-312 appeared to establish and maintain personnel qualifications for test program conduct. The requirements appeared to meet the licensee's commitments. The described program provided for proper training and qualification control.

-

The station administrative procedures need updating to properly identify organizations and titles of responsibilities for.the conduct of the initial startup and power ascension programs.

3.2 Ouality Assurance for the Startuo Procram (35501)

,

The objective of this inspection was to assure effectiveness and implementation of the licensee's QA program that covers startup and power

'

ascension test program activities.

3.2.1 Discussion

<

The inspectors reviewed instructions and had discussions with QA personnel to verify that requirements were established and plans developed for inspection i-30-

.

.

,

of the conduct of testing, tracking of test deficiencies, and test documentation. The applicable procedures were NQA 3.23,_ " Surveillance Program," Revision 6, and NQA 3.07, " Quality Assurance Audit Program,"

Revision 7.

The startup testing phase activities were listed in the FSAR, Chapter 14 as 24 test areas. The initial startup test manual defined these 24 activities in 38 initial startup test procedures. The inspectors found that all initial startup procedures had received approval by the SORC.

Regarding audits, a draft "1993 Quality Assurance Audit Schedule" was written.

'

The schedule addressed such general subjects as test control and design control, both to be performed in April of 1993 on Unit 2 as well as Unit 1.

Each of the 26 audit areas had an " Operational Master Audit Plan" which.

defined the audit requirements. No audits of Unit 2 initial startup testing had been performed.

Regarding surveillance activities of 38 initial startup test procedures, 37

,

had been reviewed by QA. The QA review was for the inclusion of notification points. For each notification point, a detailed instruction was being developed. Approximately one-third of these detailed instructions were drafted with a format of purpose, objectives, philosophy, attribute, and description of the test and inspection activity. With one exception, initial startup testing will begin at the time of fuel load. One Initial Startup Procedure, 150-223B, " Remote Shutdown capability Test," Revision 0, was performed during HFT.

The inspectors found notification points in Initial Startup Procedure 150-223B were signed off. The report on the QA overview of HFT identified that Initial Startup Procedure ISU-223B test had been witnessed and reviewed.

3.2.2 Conclusions The licensee had established an adequate program for audit and surveillance of startup test activities.

3.3 Procedure Reviews (72300)

In this area of the inspection, the inspectors evaluated procedures that will be used in the startup program.

In particular, the inspectors reviewed the,

procedures for technical and administrative accuracy and consistency with regulatory requirements, guidance, and licensee commitments contained in the FSAR, regulatory guides and industry standards, and other licensee correspondence.

3.3.1 Initial Fuel Loading Procedure (72500)

In this area of the inspection, the inspectors reviewed the licensee's procedures for initial fuel loading and core load verification to assure that the licensee's controls would satisfy FSAR commitments; regulatory requirements; and conform to necessary regulatory guidance, codes,_ and standards.

-31-

.

.

-h

.

.

<3.3.1.1 Discussion

The licensee's Initial Startup Procedure ISU-001B, " Initial Fuel Load

'

Sequence," Revision 0, defined the sequence of operations and_ tests to be

'

conducted during the initial fuel loading. The procedure had previously received the TRG approval ~ in meeting number 92-024 on October 16, 1992. The procedure summarized the test, provided acceptance and review criteria,

,

referenced related manuals and documents, specified precautions and limitations, listed prerequisites, and gave detailed sequential step-by-step instructions for the conduct of the test including restoration from the test.

In general, the procedure was a higher-tier procedure that required the :

.

licensee's staff to perform numerous activities specified by other referenced procedures.

Seven of the lower-tier procedures referenced by Initial Startup Procedure

~

150-001B were previously reviewed by NRC inspectors. Also, three of the

,

lower-tier procedures, which have now been revised one or more times, were previously reviewed by NRC-inspectors. The reviews of those' procedures are

discussed in the Fuel Integrity and Reactor Subcriticality (FIRS) inspection (see NRC Inspection Report 50-445/91-57, dated December 27,1991). These-

-

,

procedures were generally of good quality and were not re'-reviewed during the current inspection.

Among the various activities the procedure.specified for performance were

.

'

ensuring containment integrity, maintenance of an inverse count rate; verification of reactor coolant system chemistry, maintaining reactor vessel-water level, housekeeping controls, necessary communications and.personne1'

staffing at work stations, obtaining permission to commence certain-activities, verifying the operability of various fuel handling equipment,

refueling machine indexing, monitoring of RCS temperature, maintaining a fuel

-

status board, nuclear instrumentation calibration, crew briefings on evacuation-procedures, and other necessary activities. The review of the-procedure did not identify any omitted activities.

The inspectors noted that, similar to other initial startup procedures discussed elsewhere in this report, the report wasinot updated in its reference to the licensee's current organization. -~Also as discussed elsewhere

.

in this report, some initial startup procedures-directed licensee personnel to conduct certain " applicable, or, appropriate" portions of other ancillary procedures. The inspectors noted that Initial Startup Procedure.ISU-001 contained such a vague reference in Step 11.1.5.

Step 11.1.5 stated the;

-

following:

,

" Start / verify started the applicable portions of 15U-3088, Thermal.

Expansion, Power Ascension Phase."

This subjective statement. lacked in specificity as to what were applicable portions of. the ancillary procedure.. Consequently, the procedure was,-in-regard to this issue,_not appropriate, and was contrary to the requirements.of; Criterion V, " Instructions, Procedures, and Drawings," to Appendix B'to

'

10 CFR 50. This constituted the first example _ of. Violation 446/9302-01.

-32-

.

,

.

.

3.3.1.2 Conclusions The procedure provided a logical outline of progressive activities necessary to accomplish initial fuel loading. The procedure suitably controlled the flow of work activities by use of sign-offs of accomplished tasks. The procedure was, however, in violation of the requirements of Criterion V to Appendix B to 10 CFR 50, inasmuch as it contained a vague reference to a necessary ancillary procedure.

3.3.2 Control Rod Operability (72564)

In this area of the inspection, the inspectors evaluated two initial startup procedures for rod drop time measurements.

3.3.2.1 Discussion The inspectors reviewed Initial Startup Procedures 150-026B, " Cold Control Rod Operability Testing," Revision 0, and 15U-0278, " Hot Control Rod Operability Testing," Revision 0.

The procedures were found to be written, reviewed and approved in accordance with the licensee's administrative procedures.

Commitments contained in FSAR Table 14.2-3, Sheets 4, 5, 19, 31 and 32, were provided for in the procedures, and appeared to meet the guidelines of Regulatory Guide 1.68.

Initial plant and test conditions were properly established and controlled. Acceptance and review criteria were provided. No technical inaccuracies were identified.

3.3.2.2 Conclusions The initial startup procedures appeared to be responsive to the licensee's FSAR commitments for rod drop time measurements-and conformed to the approved administrative control procedures. The inspectors identified no technical deficiencies.

3.3.3 Reactor Coolant System Leakage (72566)

The inspectors evaluated a precritical test procedure which had been established to determine the RCS leakage rate.

3.3.3.1 Discussion Precritical Initial Startup Procedure ISU-022B, "RCS Leakage Rate Test,"

Revision 0, was developed for the purpose of determining the RCS leakage rate as defined in TS 3.4.5.2 and Chapter 14 of the FSAR. An additional purpose was to verify that no pressure boundary leakage exists. The test was to'be performed prior to initial criticality in Mode 3.with the RCS pressure between 2215 and 2255 psig and the RCS average temperature between 552 and 562*F. The procedure underwent the licensee's standard review process by the TRG, Operations, System Engineering, Performance and Test Group, Design Engineering, Wastinghouse Projects, and the Nuclear Overview Department. The.

review included a 10 CFR Part 50.59 screen on July 23, 1992, and subsequent approval by the TRG.

-33-

.

.

b

, The inspectors verified that the procedure contained acceptance criteria which were consistent with TS requirements and that specified limits were established for normal losses.

It was also verified that the necessary precautions, initial conditions, and test conditions were clearly delineated either in the procedure or in referenced Operations Surveillance Procedures OPT-303, "RCS Water Inventory," Revision 5, and OPT-Il08, " Measurement of.

Controlled Leakage," Revision 0.

3.3.3.2 Conclusions The inspectors concluded that the licensee had developed a procedure that-included all of the necessary criteria which, when implemented, would satisfy

-

verification of the RCS leakage rate and pressure boundary leakage requirements specified in the TS and FSAR.

3.3.4 Initial Criticality (72570)

In this area of the inspection the inspectors reviewed the licensee's methods for achieving initial criticality.

In particular, the inspectors attempted to ascertain that the licensee's procedures for achieving initial criticality presented a cautious and controlled approach.

3.3.4.1 Discussion To ascertain the licensee's methods for achieving initial criticality, the inspectors reviewed the following procedures:

'

Initial Startup 150-101B, " Initial Criticality and Low Power Test

Sequence," Revision 0; Nuclear Engineering Procedure NUC-1068, " Initial Criticality,"

Revision 0; and Nuclear Engineering Procedure NUC-Ill, " Inverse Count Rate Ratio

Monitoring," Revision 4.

Taken together, these procedures presented cautious and controlled methods for achieving initial criticality. The inspectors verified that they satisfied licensee commitments in FSAR Section 14.2.10.2, " Initial Criticality.".'For Initial Startup Procedure 150-101B, the inspectors observed that the following information should be available before commencing the approach to initial criticality:

The date-of the nuclear fuel design report referenced in-

paragraph 6.2.2.1; The proposed letter to NRC concerning changes to the Unit-2 Initial

-

Startup Testing Program, which was referenced in paragraph 6.3.8.3, together with any approvals required; and-34-

-

+

+

,

..

The required initial boron concentration as discussed in

paragraph 10.17.

3.3.4.2 Conclusions The licensee's procedures presented a cautious and controlled approach to achieving initial critically. They generally satisfied regulatory guidance and industry standards, and were responsive to FSAR consnitments.

3.3.5 Low Power Physics Testing (72572)

In this area of the inspection, the inspectors reviewed the licensee's procedures for the determination of the reactivity worth for boric acid, control rods, and the moderator temperature coefficient.

3.3.5.1 Discussion The instructions for the reactivity worth determination of soluble poison, control rods, and moderator temperature coefficient (MTC) were contained in the following Nuclear Engineering Procedures:

NUC-104, " Boron Endpoint Determination And Differential Boron Worth,"

.

Revision 6; NUC-120, " Rod Swap Measurements," Revision 3; and

NUC-207, "Zero Power Isothermal And Moderator Temperature Coefficient

!

Measurements," Revision 5.

The licensee intended to use these procedures that had been originally-implemented for Unit 1, but were now applicable to both units. _ These were lower-tier procedures that would be integrated into.the initial startup testing process. The _ test sequence document, Initial Startup Proceduro ISU-1018, provided for transition to Nuclear Engineering Procedure NUC-104 in Step 11.8.

The inspectors reviewed Nuclear Engineering Procedure NUC-104 and' determined that the procedure was applicable to both CPSES units.

It was also verified that the test performed by the procedure agreed with the description, methodology, and objectives specified in FSAR Table 14.2-3, Sheets 16 and 17.

The detailed procedure review confirmed th:t the Nuclear Engineering Procedure

-

NUC-104 conformed to the required format of the licensee's controlling procedures. An interview with a licensee representative provided the inspectors with details of how the test would be performed'and the expected data obtained. The procedure contained review criteria which specified that endpoint boron concentration be within 50 ppm of the design value for the

-

specified bank configuration. Additionally, _ review criteria specified that the differential. boron worth should be within 15 percent of the design value for the specified bank configuration.

Failure to meet the review criteria

'

l would not invalidate.the test and testing would be allowed to continue. The

,

!

-35-L

.

r

.

t

.

. procedure did not contain acceptance criteria. The licensee was asked to explain why it was not necessary to provide acceptance criteria for the initial startup testing of the core. A teleconference discussion with an on-site representative and the corporate reactor physics group was initiated.

This discussion revealed that acceptance criteria for boron endpoint

,

determination was specified in Nuclear Engineering Procedure NUC-205.

Additionally the value for boron differential' worth was a predictive value -

.

that would be revealed by the determination of other core parameters.

T Therefore, the licensee felt it was not necessary to specify an acceptance criteria because exceeding the review criteria value would trigger evaluation and investigation to assess the effect of failure to meet the review criteria.

,

The inspectors determined that procedure Step 11.2 which required RCS and pressurizer baron sampling was not clear about the number of samples to be drawn and analyzed. The inspectors informed a licensee representative about the unclear step.

The inspectors reviewed Nuclear Engineering Procedure NUC-120 and verified that the procedure met the licensee's requirements and the test to be performed was described in FSAR Table 14.2-3, Sheet 15. The test sequence

-

document, ISU-101B, provided for transition to Nuclear Engineering Procedure NUC-120 in Step 11.10.

The inspectors interviewed a licensee representative to evaluate how the procedure would function to provide the needed data and information. The

,

procedure used the rod swap method to determine differential and integral rod worth. Another procedure was available to obtain the data by the boration and

,

dilution method; however, this method was no longer described in the FSAR.

Nuclear Engineering Procedure NUC-120 required the use of a form that was a part of the previously used Nuclear Engineering Procedure NUC-105. There was no apparent reason to maintain the old procedure other than the need to use the old procedure form with the new procedure.

During the detailed review, the inspectors determined that Step 11.2.5.5 was not clear about how much reactivity should be inserted during control rod manipulation. Additionally, Step 11.1.10 was vague as to how a set of boron samples should be determined as reliable. The step required the results of two boron samples to be averaged and stated that ' judgement should be used to determine if the average result was reliable. The licensee was unable to identify any guidance that would be used to make this judgement on sample i

results reliability. The inspectors pointed out these observations to a licensee representative.

The inspectors reviewed Nuclear Engineering Procedure NUC-207 and determined that the procedure was applicable to both CPSES units.

It was also verified that the test performed by the procedure agreed with the description, methodology, and objectives specified in FSAR Table 14.2-3, Sheet 14.

This procedure provided detailed instructions for initiating RCS cooldown and heatup while plotting the change in RCS temperature, with respect to the change in core reactivity to determine the zero power isothermal temperature.

coefficient (ITC) of reactivity. When the ITC had been determined, the MTC-36-

t

..

.

< was calculated Jy using the value for ITC and other core parameter values that would be supplied by the startup and operations report.

The inspectors questioned one step in the procedure. Step 11.6.4.1 referenced a requirement to maintain pressurizer level constant or slowly increasing during the slow constant cooldown run for data collection, and stated that pressurizer level control should be placed in manual and charging flow slightly increased. The precautions and prerequisites of the procedure collectively required that the volume control tank (VCT) level be high and all makeup precluded prior to commencing the cooldown test run. The inspectors questioned why it was necessary to place pressurizer control in manual and increase charging flow when the control system was designed to maintain a constant pressurizer level for the conditions of the test.

In addition, a question was raised concerning the effect on VCT level which could not be made up to during the test. The licensee responded that since the procedure step stated "should" instead of "shall," it was not required to place the pressurizer level control system in manual. Also, it hau been determined that such a small cooldown (less than 8'F) would have negligible effect on VCT level. The exact reason for the step could not be determined, but the consensus was that the step had been inserted to assure that any difference in pressurizer boron concentration would not affect RCS boron concentration during ITC determination.

3.3.5.2 Conclusions The inspectors concluded that Nuclear Engineering Procedures NUC-104, NUC-120, and NUC-207 would provide the needed data to determine boron endpoint concentration, boron differential worth, control rod differential and-integral worth, and moderator temperature coefficient. The inspectors also identified -

procedural steps that were vague and confusing.

,

3.3.6 Core Reactivity Balance (72576)

In this area of the inspection, the inspectors reviewed the licensee's procedure for core reactivity balance measurements to be taken during initial

.

startup and power ascension of the new core.

3.3.6.1 Discussion Instructions for reactivity balance measurements were contained in Nuclear Engineering Procedure NUC-205, " Core Reactivity Balance," Revision 6.

The purpose of the test was to use reactivity balance calculations to normalize the measured critical boron concentration for hot full power and confirm that the hot zero power boron concentration met acceptance criteria. This procedure was the normal monthly surveillance procedure that was used on the operating unit and would be implemented for Unit 2 after testing was complete.

The overall scheme for initial startup testing intended that Nuclear Engineering Procedure NUC-205 be transitioned into from the test sequencing documents.

The inspectors verified that the procedure adhered to the commitments of the FSAR. The procedure incorporated the objective, prerequisites, and test-37-

.

.

.-

, method summarized in FSAR Section 14, Table 14.2-3, (Sheet 17). The FSAR did not specify acceptance criteria. Acceptance criteria in the form of design values and allowable deviation had been provided to the licensee by the vendor. Section 4.0 of the procedure was titled, " Review And Acceptance Criteria." Startun Administrative Procedure STA-816, " Format And Content Of Initial Startup Test Procedures," Revision 3, defined review criteria as

" additional criteria that do not render the test unsatisfactory, but may require corrective action." Further review revealed that Nuclear Engineering Procedure NUC-205 satisfied the general requirements of licensee's documents that controlled plant procedures and plant testing.

The inspectors identified two initial startup sequencing documents that transitioned to the reactivity balance procedure. Discussions with licensee representatives did not identify any other documents that referenced the procedure. The sequencing documents were Initial Startup Procedure 150-101B,

" Initial Criticality and Low Power Test Sequence," Revision 0, Step 11.8.3, and Procedure 150-2808, "100% Reactor Power Test Sequence," Revision 0,

,

Step 11.2.1.

Both of these documents instructed test personnel to perform the appropriate portion of Nuclear Engineering Procedure NUC-205. The appropriate portion was not identified in Initial Startup Procedures 150-101B or 15U-2808, or Nuclear Engineering Procedure NUC-205. Additionally, Nuclear Engineering Procedure NUC-205 did not contain any guidance for determining the appropriate portion. The three major procedure sections were not listed in the sequence in which they would be performed during the startup testing. The failure to provide adequate instructions in Procedures ISU-101B and 150-280B constitute additional instances of example one of Violation 446/9302-01.

3.3.6.2 Conclusions The inspectors concluded that Nuclear Engineering Procedure NUC-205 would adequately provide the needed data to affirm that the design requirements were met and to normalize boron concentration to the hot full power equilibrium conditions. The inspectors were concerned that a potential existed for error during transition between the test sequence guides and the Nuclear Engineering Procedure NUC-205. These potential errors could result in a failure to perform required testing or the performance of testing with the improper conditions.

3.3.7 Reactor Power Test Sequence (72578)

In this area of the inspection, the inspectors evaluated two startup procedures for evaluation of core performance.

3.3.7.1 Discussion The inspectors reviewed Initial Startup Procedures ISU-2408, "50% Reactor Power Test Sequence," Revision 0, and ISU-260B, "75% Reactor Power Test Sequence," Revision 0.

These procedures defined the order of tests to be conducted during power escalation from 0 to 75 percent reactor power and the tests which are required to be performed at selected steady state power levels during the power increase in order to evaluate core performance. The reviewed procedures did not provide acceptance criteria. Specific acceptance criteria-38-

., -

.

.

,were provided in each specified test to be conducted under the stipulated plant conditions. The procedures were found to be written, reviewed, and approved in accordance with the licensee's administrative procedures. No-technical inaccuracies were identified; however, these procedures exhibited the same lack of specific direction in numerous steps that required the test engineers to accomplish ancillary procedures.

In particular, these steps instructed the licensee's staff to perform " appropriate" portions of other procedures. This was a vague and undefined direction that placed an unnecessary burden on the staff members to be highly conversant with numerous procedures that cover diverse areas requiring knowledge in multiple technologies. These inappropriate references are further instances of example one of Violation 446/9302-01.

3.3.7.2 Conclusions The initial startup procedures appeared to be responsive to the licensee's FSAR commitments and provided test program conduct and control to establish plant conditions required for evaluation of core performance. The inspectors identified no technical deficiencies; however, the procedures contained further examples of Violation 446/9302-01.

3.3.8 Load Reduction / Rejection Tests (72580)

-

During this portion of the inspection, the inspectors reviewed two initial startup procedures. The two tests are to be performed with the reactor at power (greater than 5 percent). The two procedures reviewed were:

Initial Startup Procedure 150-263B, "Large Load Reduction," Revision 0;

and Initial Startup Procedure 15U-2848, " Dynamic Response'To Full Load

Rejection and Turbine Trip," Revision 0.

3.3.8.1 Discussion The purpose of these tests were:

To verify that the primary plant, the secondary plant and the automatic

control systems could sustain a 50 percent load rejection with neither a reactor or turbine trip (ISU-263B);

To evaluate the interactions between control systems to determine if

control system setpoint changes are needed to improve plant transient response (150-263B); and To demonstrate the ability of the primary and secondary plant and their

automatic control systems to sustain a trip from 100 percent power and establish stable conditions 'following the transient (150-2848).

The procedures contained clearly defined acceptance criteria and purposes.

'

The instruction sections appeared technically correct and contained the-39-

.

.

,appropr M e step sign-offs.

Initial plant conditions and power levels were speo fied. TS requirements, ere appropriate, were specified.

Predicted prant response and possible sonormal plant responses were identified with appropriate cautions and actions that may be required. The procedures also contained the appropriate listing of references such as drawings, vendor

.iuals, licensing documents, and ancillary procedures. All test equipment, special conditions, precautions, limitations and clarifying notes were-included in specific sections of the tests. The attachment section of the-procedures contained the necessary data sheets and tables.

>

The inspectors noted that the procedures exhibited a weakness in human factors considerations in that there was a lack of continuity and content in

'

Sections 6.0, References; 8.0, Special Conditions; 9.0, Precautions,

,

Limitations and Notes; and 10.0, Prerequisites.

For example, procedures referenced in Section 6.0 were categorized under the~ heading of documents in

one procedure and specifically categorized by type in the other.

In Section 8.0, the notifications to the shift supervisor were very detailed as to expected possible events except for the possible lifting of the steam generator relief / safety valves.

In Section 10.0, the precaution to ensure that all personnel should be clear of the steam generator safety valve rooms was only included in one procedure when it should have been in both procedures.

In Section 10.0, several steps appeared to contain two action requirements but only one sign-off.

3.3.8.2 Conclusions The inspectors concluded that the procedures were technically adequate and contained the necessary steps to produce data to satisfy the stated acceptance criteria. There were human factors weaknesses identified in the area of consistency in several sections of the procedures. The licensee attributed this to the different authors of the procedures. This may indicate a weakness in the integrated peer review process.

3.3.9 Loss of Offsite Power (72582)

In this area of the inspection, the inspectors reviewed Initial Startup Procedure 150-222B, " Turbine Generator Trip With Coincident loss of Offsite Power," Revision 0.

3.3.9.1 Discussion The purpose of this test was to demonstrate that the plant will respond properly to a turbine / reactor trip with no offsite power available.

The procedure contained clearly defined acceptance criteria and purpose. The instruction sections appeared technically correct and contained the appropriate step sign-offs.

Initial plant conditions and power levels were specified. TS requirements, where appropriate, were specified.

Predicted plant response and possible abnormal plant responses were identified with appropriate cautions and actions that may be required. The procedures also contained the appropriate listing of references such as drawings, vendor manuals, licensing documents and ancillary procedures. All test equipment,

-40-

.

.

, special conditions, precautions, limitations and clarifying notes were included in specific sections of the tests. The attachment section of the procedures contained the necessary-data sheets.and tables.

This procedure also exhibited the same type of human factors weaknesses as described for the two procedures above.

.

3.3.9.2 Conclusions The inspectors concluded that the procedure was technically adequate and-contained the necessary steps to produce data to satisfy the stated acceptance criteria. There were human factors weaknesses identified in the area of consistency in several sections of the procedures. -The licensee attributed this to the different authors of the procedures. This may indicate a weakness in the integrated peer review process.

3.3.10 Core Performance and Physics Code (72584)

The inspectors reviewed the licensee's procedure for the evaluation of core performance during the initial startup testing process.

.

3.3.10.1 Core Performance 3.3.10.1.1 Discussion The instructions for core performance evaluation were contained in Nuclear Engineering Procedure NUC-201, " Surveillance Of Core Power Distribution Factors," Revision 7.

The licensee intended to use this existing surveillance procedure that had been originally implemented for Unit 1, but was now applicable to both units. This procedure was to be integrated into the

,

initial startup testing process. Test sequence documents were reviewed to determine when Nuclear Engineering Procedure NUC-201 would be used. Nuclear Engineering Procedure NUC-201 was entered into from the following test procedure steps:

Initial Startup Procedure 15U-240B, "50% Reactor Power Test Sequence,"

Revision 0, Steps 11.27.3 and 11.40.2; Initial Startup Procedure 150-260B, "75% Reactor Power Test Sequence,"

Revision 0, Step 11.13.5; and Initial Startup Procedure ISU-2808, "100% Reactor Power Test Sequence,"

Revision 0, Step 11.20.5 Nuclear Engineering Procedure NUC-201 was transitioned into a total of four times during the initial startup testing sequence. Core power distribution factors were evaluated at 25-30, 45-48, 75-78, and 98-100 percent power. Each sequence procedure step that transitioned to Nuclear Engineering Procedure NUC-201 stated, " Perform the appropriate portion of Nuclear Engineering '

l Procedure NUC-201." The inspectors asked a licensee representative which

,

portions of the procedure were to be performed. The initial response to this-41-

,

.

.

question was that the procedure should be performed in its entirety every time that it was entered into. The licensee representative volunteered a later response stating that it was only necessary to perform two specific sections of the procedure to satisfy initial startup testing requirements. However, it would be a nuclear engineering group decision if the entire monthly surveillance should be performed at each power level specified for initial startup testing. As a result of these discussions, the inspectors believed that a thorough review of the integrated effects of merging the two procedures had not been performed and is a further instance of example one of Violation 446/9302-01.

The inspectors reviewed Nuclear Engineering Procedure NUC-201 and determined that the procedure was applicable to both units.

It was also verified that the test performed by the procedure agreed with the description, methodology, and objectives specified in FSAR Table 14.2-3, Sheet 22. A recently approved change to this test description eliminated the requirement to perform asymetrical or dropped rod flux testing as the data from previous testing of similar cores could be assumed applicable for Unit 2.

The detailed review confirmed that Nuclear Engineering Procedure NUC-201 conformed to the required format of the licensee's controlling procedures.

Interviews with a licensee representative provided the inspectors with details-of how the test would be performed and the design comparison data obtained.

The procedure provided valid data needed for a conservative and cautious escalation to full power. The overall integrated process required the most recent data supplied by Nuclear Engineering Procedure NUC-201 to be extrapolated to the next stable power level prior to raising power to the next.

testing plateau.

3.3.10.1.2 Conclusions

,

The inspectors concluded that the performance of Nuclear Engineering Procedure NUC-201 would result in the power distribution data needed to make a valid comparison with design parameter values. Also the procedure's performance should provide necessary data. The inspectors noted that the licensee had not considered the integrated performance resulting from the use of the test-sequence documents with Nuclear Engineering Procedure NUC-201. The use of the procedures with the inappropriate references noted above had the potential to raise the core power level without the licensee having evaluated and extrapolating the existing core power distribution. These inappropriate procedure steps were further examples of violation 446/9302-01 of Criterion V of Appendix B to 10 CFR 50.

3.3.10.2 Core Physics Code The instructions for processing of flux map data by executing the detector and burnup routines of the CONF 0RM core physics computer code package were-contained in Nuclear Engineering Procedure NUC-115, "Onsite Execution of the CONFORM Core Physics Code," Revision 4.

The procedure provided review and acceptance criteria for detector runs and a trigger to comply with TS Surveillance Requirement 4.2.2.3.

The procedure was applicable to both units, in all modes of operation.

-42-

.

.

,3.3.10.2.1 Discussion The detailed review conducted by the inspectors confirmed that Nuclear l

Engineering Procedure NUC-115 conformed to the required format of the licensee's controlling procedures. The inspectors noted that although the procedural steps were clear, some steps required subjective engineering evaluations with little guidance beyond the acceptance criteria, which were only applicable to detector runs, not burnup runs. The instructions for execution of the detector code stated that the review criteria should "not necessarily be the only factors reviewed" for flux map evaluations; however, other criteria to be used and under what circumstances were not specified or i

,

referenced. Similarly, instructional guidance on burnup data review and acceptance criteria were not specified.

The procedure was evaluated as effective in triggering compliance with TS Surveillance Requirement 4.2.2.3.

The inspectors were concerned, however, that only the Detector Code instructions referenced the.TS, and the procedure did not explicitly state whether the burnup code should be executed only after the detector code.

The inspectors were also concerned with the emphasis in Section 12.0,

'

" Restoration and Retesting." Specifically Step 12.1 stated, " Deviations from the review criteria may not necessarily.... constitute a safety problem."

The inspectors felt the statement should reflect that " deviations from the review criteria may constitute a safety problem." Similarly in Section 12.3, the inspectors were concerned that failure of Criterion 4.1.1.6 should contain a precautionary reference to TS 3.2.4, concerning Quadrant Power Tilt Ratio.

3.3.10.2.2 Conclusions The inspectors concluded that Nuclear Engineering Procedure NUC-115 would adequately provide an experienced performer the guidance needed to conduct the detector and burnup routines of the CGa.ORM core physics computer code package and reasonably assure the performance of TS Surveillance Requirement 4.2.2.3.

The inspectors noted the lack of specific flux map evaluation and acceptance -

criteria, and heavy reliance on engineering judgment in the procedure.

t 3.3.11 Incore Flux Mapping (72584)

In this area of the inspection, the inspectors evaluated a nuclear engineering procedure that provided for incore flux mapping.

3.3.11.1 Discussion The inspectors reviewed Nuclear Engineering Procedure NUC-102, "Incore Flux Mapping System Operating Instructions," Revision 5.

This procedure detailed the operation of the incore flux mapping system and provided instructions for satisfying the technical requirements for movable incore detectors. This procedure may be used for either unit when core flux mapping is required. The procedure was found to be written, reviewed and approved in accordance with the licensee's administrative procedures. No technical inaccuracies were identified.

,

-43-

.

.

.

,3.3.11.2 Conclusions The nuclear engineering procedure to support the licensee's requirements for evaluating core performance during the initial startup test sequence appeared to be adequate and contained no technical deficiencies.

,

3.3.12 Overall Conclusionslof Procedure Reviews-

'

The initial startup and power ascension procedures reviewed appeared to be technically adequate. They were responsive to FSAR commitments, and conformed to applicable regulatory guidance, industry standards, and approved administrative controls. The procedures suitably included necessary

zcceptance criteria and appropriately allowed for the control of work activities by use of sign-off steps for accomplished tasks. The procedures, in general, presented a cautious and controlled approach to achieving initial critically and subsequent startup activities. However, 7 initial startup or nuclear engineering procedures were inappropriate in that they required the performance of " appropriate," " applicable," or unclearly defined portions of ancillary procedures. These unclear references could result in a failure to perform required activities or the performance of testing with improper conditions. This was in violation (446/9302-01) of the requirements of Criterion V to Appendix B to 10 CFR 50.

4 ONSITE REVIEW OF LICENSEE EVENT REPORTS (92700)

4.1 (Closed) Construction DeficienCv SDAR CP-92-005: " Safety Chiller Control Circuit Voltace Droo" During the _ cable replacement evaluation associated with the fire safe shut down requirements, the licensee discovered that the train A safety chiller control circuit had not been enveloped in the engineering analysis for Class IE miscellaneous 120 VAC control circuits. Upon further investigation, the licensee determined that during a degraded voltage condition, the resulting voltage to the starter circuit would be less than that required to start the chiller unit.

The inspectors examined the licensee's corrective actions associated with this deviation. To resolve the less than adequate voltage condition, the licensee developed and implemented design change DCA 24375, Revision 4, to increase the cable size and reduce the cable length thereby reducing the voltage losses due

,

to the cable. The licensee determined that the reduction of cable loss with the degraded voltage condition would result in adequate resulting voltage to ensure that the train A safety chiller would operate when needed.

Based on the inspectors review, it was determined that the licensee had developed and implemented appropriate corrective action to address the

,

identified deviation.

-44-

,

.

.

.

.

b

.

, 4.2 (Closed) Construction Deficiency SDAR CP-92-012:

"Ferro-Resonant Transformers" During the performance of the preoperational testing of the 7.5 kva inverters, the licensee discovered that all four of the inverters being tested failed to meet the total harmonic distortion (THD) criteria of less than or equal.to five percent.

In addition several spare ferro-resonant transformers drawn from the warehcuse failed to meet the 5 percent THD criteria.

The inspectors examined the licensee's corrective actions associated with this deviation. The licensee requested a technical evaluation from the inverter vendor concerning inverter operation with a THD of greater than five percent.

The same information was requested from the vendor of the loads fed by the inverters. The documented response from the vendor was that no adverse effects on inverter operation or operation of the connected loads would occur at a THD of up to 10 percent. The highest THD documented by the licensee was 5.8 percent. The license therefore determined that no further action need be taken to resolve this issue.

Based on the inspectors' review, it was determined that the licensee had ceveloped and implemented appropriate corrective action to address the identified deviation.

5 FOLLOWUP ON CORRECTIVE ACTIONS FOR VIOLATIONS (92702)

5.1 (Closed) Violation 446/9232-01:

Failure to Maintain Acoropriate Administrative Control of Test Procedures and Startuo Doeratina-Instructions This violation involved the failure to maintain control of documents in the control room. Specifically: 1) Startup operating instructions were expired but not discarded from the files, 2) Current copies of prerequisite testing procedures were not on file, 3) Preoperational test procedures were missing-change notices, and 4) Preoperational test procedures were not changed according to the procedure requirements. The licensee corrected the examples identified in the violation.

In addition, an individual was assigned to the control room to review all documents to assure their status was current and identify changes required to prevent repetition. Startup Deficiency Reports 2783 and 2879 were written to document the last of these problems.

During this reporting period, the inspectors reviewed the licensee's response that was delineated in TU Electric's Letter TXX-92476, dated October 13, 1992.

Presently startup operating instructions have been replaced with system operating procedures, preoperational test procedures have been replaced with initial startup test procedures. Prerequisite test procedures have been phased out of use.

The inspectors reviewed the revision status of system operating procedures, operating testing procedures, and initial startup test procedures and verified that the control room had current information. A sample of 27 system operating procedures and 17 operation testing procedures were reviewed on January 7, 1993. The startup test procedures were all initial issue.

-45-

.

.

,The inspectors noted that Procedure Change Notice 6, effective November 21, 1992, was not in the control room copy of Procedure 50P-313, " Turbine Plant-Cooling Water System." The licensee issued ONE Form 93-131 to document this failure to maintain document control in the control room. This procedure was

" walked down" on December 23, 1992. The errors in the instrumentation list.

that were corrected by this procedure change notice, were documented as exceptions during the " walk down." This means the " walk down" was 'done without the use of the procedure change notice. This was a violation of Procedure ODA-104, Revision 8, with Procedure Change Notice 1, paragraph 6.6.1 which required the control room system operating procedures be maintained.

The licensee's corrective actions identified in letter TXX-92476, dated October 13, 1992, stated a review would be performed on the administrative handling of documents and change identified to prevent repetition of similar problems. This is a repeat of the original violation.

It appeared that the action to prevent repetition was not effective.

'

Based on the above reviews, the inspectors determined that specific corrective actions were implemented to address the identified violation, but there was an apparent lack of effectiveness of the corrective actions to preclude similar problems.

This was identified as a violation of Criterion XVI of Appendix B to 10 CFR 50 (violation 446/9302-02).

,

-46-

.

.

ATTACHMENT 1

.

'

2 PERSONS CONTACTED l.1 Licensee Personnel

• D. Allen, Manager, Initial Startup R. Braddy, Project Manager, Unit 2 D. Buschbaum, System Engineering

,

H. Carmichael, Manager, Engineering Assurance S. Ellis, Manager, Power Ascension Program

'

T. Evans, Electrical Systems Supervisor J. Green, Maintenance Engineer

• J. Greene, Licensing Engineer, Unit 2 Licensing W. Hartshorn, Independent Safety Engineering Group Engineer J. Hitzfeld, Lead Preventive Maintenance Engineering Technician T. Hope, Licensing Manager, Unit 2 N. Hottel, Engineer, Independent Safety Engineering Group H. Harvray, Supervisor, Technical Programs B. Mays, Supervisor, Mechanical Codes & Standards J. McCormick, Operations Quality Specialist G. Merka, Licensing Engineer G. Ondriska, Supervisor, Programs Test Group, Startup J. Patton, Quality Specialist B. Phillips, Principle Engineering Technician

• J. Snyder, Startup Manager S. Swilley, Station Welding Engineer N. Terrel, Supervisor, Station Nuclear Engineering L. Walker, Licensing Engineer

• R. Walker, Manager, Regulatory Affairs C. Welch, Senior Quality Assurance Specialist, Operations Quality Assurance

• J. Wren, Construction Quality Assurance Manager, Quality Assurance-1.2 Contractor Personnel C. Barbehenn, Test Engineer, Initial Startup J. Dorn, Operations Support G. Fanning, Manager, Project Quality Assurance, Brown & Root B. Landry, Test Engineer, Initial Startup J. Lunsford, Quality Control Inspector, Brown & Root M. Pitluk, Project Engineer, Bechtel D. Rencher, Manager, Unit 2 Engineering S. Trickkovic, Startup Engineer, Bechtel In addition-to the personnel listed above, the inspectors contacted other personnel during this inspection.

• Denotes personnel that attended the exit meeting.

2 EXIT MEETING An exit meeting was conducted on January 30, 1993. During this meeting, the insp'ector reviewed the scope and findings of the report.- The licensee did not

~;

..

, identify as proprietary any information provided to, or reviewed by, the

_

'

inspectors.

Previously, on January 21,:1993, a management briefing was= held to discuss the preliminary results of-the portion of this inspection that

,

addressed the " Testing Piping Supports and Restraint-Systems," (see paragraph-2.5).-

_

During the exit meeting, the licensee's management committed:to three actions

in response to the violations. The first action was to have power ascension training for test personnel to assure that they are thoroughly familiar with

.

promptly to provide feedback as to the effectiveness of the training. The

~

the administrative requirements of the program, and to review test results

',

second action was to revise or change initial startup procedures prior'to use.

!

to provide a description of which section of-any " branch out" procedures which

'

are being required to be performed. The third action wasito intensify i

corrective action to ensure control room procedures are up to date.-

.

-t

~

>

>

c J

l l

t-2--

i

.

.

.

ATTACHMENT 2 DEFERRED TESTS AND RETESTS ASSESSMENT

,

Preoperational and acceptance tests and retests that were not complete have been analyzed for plant operability.

It was determined that these tests and retests were not required to be complete prior to the loading of reactor fuel, based on specific criteria given below for individual items and the following criteria applicable to all items:

(I)

The system or component was not required to place the plant in a safe shutdown condition prior to the designated mode or time of testing.

(2)

The system or component to be tested was not required to mitigate the consequences of any accident prior to the designated mode or time of testing.

(3)

The system was not required to prevent release of radioactivity to the environment prior to the designated mode or time of testing.

Inspectors have reviewed requirements for and status of these tests and retests and agreed with the licensee's conclusions. Additionally, the inspectors were briefed on the licensee's intended computer-based scheduling process that will be used to ensure that these tests and retests are completed prior to the required time or plant mode.

Listed below are the systems and components for which testing or retesting have been requested to be deferred until after issuance of the low-power license.

........................................................................+.....

DEFERRED TESTING

..............................................................................

MIXED BED DEMINERALIZER Differential pressure measurements across one of the two mixed bed demineralizers at normal operating pressure had not been taken. The licensee's January 8,1993, letter referred to three instead of two mixed bed demineralizers. Upon questioning by the inspectors, a licensee representative stated that the cation demineralizer (a non-mixed bed demineralizer) was mistakenly included as a mixed bed demineralizer in the letter. The licensee

,

had stated in their letter that this testing would be complete prior to entry into Mode 2.

The other mixed bed demineralizer was available for maintaining chemistry requirements associated with Modes 3 and 4, if needed. The applicable procedure to be used for this deferred testing was Operations Work Instruction 0WI-104, which was licensee approved.

BORON THERMAL REGENERATION DEMINERALIZERS The vibrational response and differential pressure across the boron thermal regeneration demineralizers at normal operating pressure had not been taken.

The licensee had stated in their letter that they would perform this testing prior to utilizing the system for its intended purpose. The applicable procedure to be used for this deferred testing was Performance and Test

'

cs.

..

, Procedure PPT-TP-93B-7, which has scheduled to be licensee-approved on March 1, 1993.

EXTRACTION STEAM PIPING The thermal expansion measurements on the extraction steam piping line to the auxiliary steam system had not been perforued. The licensee had stated in their letter that this testing would be complete prior to commercial operation. The applicable procedure to be used for this deferred testing was Initial Startup Procedure ISU-3088, which was licensee approved.

PUBLIC ADDRESS AND EMERGENCY EVACUATION ALARM SYSTEM The evacuation alarm test and the audibility test of the Gatronics system during the highest expected ambient noise levels had not been conducted.. The licensee had committed to have the communications system verified functional prior to fuel load. The licensee had stated in their letter that.this testing would be complete prior to entry to Mode 2.

The. applicable procedure to be used for this deferred testing was Performance and Test Procedure PPT-TP-92-01, which was scheduled to be licensee-approved on February 21, 1993.

REFUELING CAVITY SKIMMER PUMP AND LOOP FLOW The actuation and operation of the reactor cavity skimmer pump and correct-loop flow had not been demonstrated. The licensee intended to conduct the initial core loading under dry conditions, and did not intend to need this system until the first refueling outage when it will be tested. The applicable procedure to be used for this deferred testing is Performance and Test Procedure PPT-TP-92-22, which was scheduled to be licensee-approved on March 1, 1993.

STEAM DUMP VALVES The licensee had not demonstrated the operability (i.e., valve stroke times)

of the steam dump valves. The licensee had stated in their letter that this testing would be complete prior to entry into Mode 2.

The applicable procedure to be used for this deferred testing was Performance and Test Procedure PPT-TP-93B-5, which was scheduled to be licensee-approved on February 15, 1993.

REACTOR CAVITY HUMIDITY DETECTORS AND ANNUNCIATORS The licensee had not demonstrated the operability of the reactor cavity humidity detectors and associated annunciators. The licensee stated in their letter that this testing would be complete prior to entry into Mode 2.

The applicable procedure to be used for this deferred testing was Performance and Test Procedure PPT-TP-93B-6,-~which was scheduled to be licensee-approved on.

February 15, 1993.

HEATING, VENTILATION, AIR CONDITIONING SYSTEMS FLOW BALANCE The licensee had not demonstrated the adequacy of ventilation flow' to areas serviced by the primary plant ventilation system. The licensee had stated that all testing required by the Technical Specifications had been completed.

The licensee had committed that the remaining flow balancing would be

completed prior to entry into Mode'I. The applicable procedure to be used for this deferred testing was Procedure XCP-ME-14, which is approved.

-2-

,

.

.

, PLANT COMPUTER The licensee had not completed the testing on several modules or hardware items in the plant computer. There are nine such items discussed below for deferred testing.

CPU Usage Time - An upgrade to increase the computer's responsiveness during times when high data flow exists will be installed and tested.

The licensee had asserted that the current computer capabilities for handling data flow will be sufficient for most times below Mode 1, when high data flow is not anticipated. The licensee had committed to have this upgrade tested prior to Mode 5.

The applicable procedure to be used for this deferred testing was Procedure PPT-TP-93B-3, which was

,

approved.

Data Archive - The licensee intended to increase the size of the computer's memory storage. Such an upgrade would reduce the frequency of computer memory downloads from the currently projected every couple of days to every 2 weeks. The licensee had committed that this upgrade would be installed and tested prior to entry into Mode 2.

This upgrade should have no impact on plant operations. The applicable procedure to be used for this deferred testing ws: Procedure PPT-TP-93B-4, which was to be approved on Februaif 20, 1993.

'

Flux Mapping Module - The software used for confirming that flux mapping data conform with Technical Specifications requirements associated with Mode 1 and above 15 percent power had not been declared operational.

The calculations performed by the software could be performed manually; however, the licensee committed to complete this-testing prior to entry into Mode 1.

The applicable procedure to be used for this deferred testing was Procedure PPT-TP-93B-1, which was approved.

Reactor Protection System Monitor Module - The licensee intended to enhance the computer's utility, adding the capability to have a reactor protection system module emulate the logic in the solid state protection system (SSPS). This would enable the licensee to use an alternative means of assessing the SSPS and enhance post-event assessments. The licensee had committed to complete the testing on this

,

module, which would remain off except for testing purposes, prior to entry into Mode 1.

The applicable procedure to be used for this deferred testing had not been developed, but was scheduled for approval on February 15, 1993.

Delta 1 Module - The module that determined and tracked penalty minutes for Delta 1 and would be used to confirm compliance to Technical Specification requirements associated with Mode 1 and above 15 percent power was not fully functional. Specifically, the feature that delays the counting of penalty minutes for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> was not operational. This delay feature could be handled manually; however, the licensee had committed to complete this testing prior to entry into Mode 1.

The applicable procedure to be used for this deferred testing had not been developed, but was scheduled for approval on February 15, 1993.

-3-

-

. -

-

.

_

.

L.

.

Primary Plant Performance Module - The module that performed

-

,

calorimetrics and would be used to confirm compliance to Technical Specification requirements associated with Mode 1 and above 15 percent

,

power was not operational. Currently, the module adequately calculated plant power, but did not properly display data. The licensee committed to have this module operational prior to exceeding 15 percent power.

The applicable procedure to be used for this deferred. testing had not been developed, but w a L iuled for approval on February 20, 1993.

Boron Follow Module - The module that provided the reactor coolant chemical conditions was not functional. This module was designed to improve the economical controls over reactor coolant system chemistry.

'

This module was not to be used for compliance with Technical Specifications. The licensee had not established a schedule for the testing of this module. The applicable procedure to be used for this deferred testing had not been developed, nor had the schedule for the approval for the procedure.

Control Room Printer Fail-Over - The module that diverted data printout upon printer failure to another printer was not operational.

In the event, a printer failure occurred, data wouM not be lost, but the hardcopy might. The licensee had not established a schedule for the testing of this enhancement module. The applicable procedure to be used for this deferred testing had not been developed, nor had the schedule for the approval for the procedure.

Computer Inputs - The licensee had not verified approximately 170 computer input points from field locations. None of these~ input points were used for the safety parameter display system (SPDS) or the emergency response facility (ERF). The licensee had not established a schedule for the testing and calibration of these input points. The applicable existing procedures and work instructions to be used for this deferred testing had not been identified.

............................................................................

DEFERRED RETESTING

..............................................................................

POWER-0PERATED RELIEF VALVES The licensee had not demonstrated the leak tightness of the power-operated relief valves. These valves were reworked following the hot functional-test.

Normal reactor coolant system operating pressure and temperature were necessary to test these components. The licensee stated in their letter that this testing would be complete prior to entry into Mode 2.

The applicable procedure to be used for this deferred testing was Initial Startup Procedure ISU-021B, which was licensee-approved and is scheduled to be changed on February 15, 1993.

PRESSURIZER SPRAY VALVE The licensee had not demonstrated that the leak rate of the pressurizer spray valve was acceptable. This valve was reworked following the hot functional test.

Normal reactor coolant system operating pressure and temperature are

,

-4-

-- -

-

- -

..

.

.

,necessary to test this component. The licensee had stated in their letter that this testing would be complete prior to entry into Mode 2.

The i

- applicable procedure to be used for this deferred testing was Initial Startup Procedure 150-0218, which was-licensee-approved.

MOISTURE SEPARATOR REHEATERS The licensee had not demonstrated the ability to draw chemistry samples from:

the moisture separator reheaters. The licensee stated in their letter that this testing would be complete prior to commercial operation. The applicable procedures to be used for this deferred test'ng were chemistry Procedures CHM-503 and CHM-514, which were licensee-approved.

MAIN STEAM ISOLATION VALVE The licensee had not demonstrated the operability (i.e., valve stroke time) of one main steam isolation valve (MSIV). Normal secondary-side system operating pressure and temperature were necessary to test this component. The licensee

.

had stated in their letter that this testing would be complete prior to entry

'

into Mode 2.

This MSIV was to be maintained shut until the testing in Mode 3 was completed, and at that time the licensee would open the valve. and enter the two relevant limiting conditions of operation (LCO). The applicable procedure to be used for this deferred testing was operations surveillance Procedure OPT-5098, which was licensee-approved.

-5-

,

-.

. -

-

-

- -

.

-

hw'

- _

<

..

7.?

AIIACMENT 3

,

'

PIPE SUPPORTS EXAMINED RESIDUAL HEAT REMOVAL SYSTEM

-

,

RH-2-063-410-S22R-

.;

.

RH-2-007-404-522R

'

>

RH-2-007-407-S22K'

RH-2-063-411-S22R t

RH-2-007-405-S22K

'

'

-

RH-2-007-406-S22R

-

RH-2-063-403-522R RH-2-063-401-522R

RH-2-009-405-S32S

,

RH-2-009-404-S22R RH-2-009-403-522K

-

'

RH-2-007-701-S22K

CONTAINMENT SPRAY SYSTEM

'

CT-2-014-445-S22K f

CT-2-025-405-S22K

CT-2-057-404-S32R

.l CT-2-120-401-S32S

,

-

,

SAFETY INJECTION SYSTEM

.

I SI-2-029-424-Y32K

'

SI-2-071-403-S325 SI-2-071-404-S32K-SI-2-070-403-S22R

<

,

SI-2-070-405-S22K SI-2-070-406-S22R

SI-2-325-700-S32R

,

REACTOR COOLANT-SYSTEM

,

'

RC-2-115-430-C56R

- RC-2-115-431-056K

RC-2-Il5-432-C56K

,

RC-2-115-412-056S

'

RC-2-115-433-C56R-q MAIN STEAM SYSTEM

'

MS-2-026-700-S75K-

,

-MS-2-026-406-S75K-

.

MS-2-027-423-S73R

!

MS-2-027-424-S73R MS-2-027-425-573R'

MS-2-025-700-575K'

MS-2-025-700-575K

'

,

-J

._

If J u...,

a '

pc.

COMPONENT COOLING WATER CC-2-019-702-A43K

'

CC-2-019-703-A435.

,-

CC-2-019-704-A43R-

-CC-2-019-716-A43K

'

~ CC-2-019-707-A43S

.

CC-2-019-705-A43R

'1 CC-2-019-706-A43R-f AUXILIARY FEEDWATER SYSTEM

,

AF-2-047-402-Y46R AF-2-047-404-Y46R-AF-2-047-401-Y46S

,

AF-2-103-411-553R

AF-2-103-430-553R l

.

$

.

,

k b

'I

,

I

,

i:;

'$

i

!

>

)

.

i

'

-2-r

[

,

'

.

--.

y.

-

.,

C

-

ATTACHMENT 4

,

PROCEDURES AND DATA REVIEVED RELATED TO PARAGRAPH 2.4.2.2 0F THE INSPECTION REPORT Preoperational Test Procedure 2CP-PT-90-03, " Hot functional Piping Systems Thermal Expansion Test," Revision 0 Initial Startup Procedure ISU-0228, " Reactor Coolant System Leakage Rate Test," Revision 0 Preoperational Test Procedure CP-SAP-07A, " Prerequisite Test Instruction / Procedure Preparation, Conduct, Review and Approval," Revision 1 Preoperational Test Procedure CP-SAP-078, "Preoperational Testing," Revision 3 Preoperational Test Procedure CP-SAP-16, " Deficiency and Nonconformance Reporting," Revision 19 Operations Surveillance Procedure OPT-Il08, " Measurement of Controlled Leakage," Revision 0 Operations Surveillance Procedure OPT-303, " Reactor Coolant System Water Inventory," Revision 5 RELATED TO PAPAGRAPH 2.5 0F THE INSPECTION REPORT Procedures:

Engineering Procedure 2EP-5.12, * Design Criteria for Pipe Stress and Pipe Support" Engineering Procedure 2EP-5.13, " Design Guidelines for Pipe Stress and Pipe Support Engineering Procedure 2EP-5.14, " Procedure for Pipe Stress / Pipe Support (Scope A) Final Reconciliation" Engineering Procedure 2EP-5.15, " Moment Restraints and Support Qualification -

Procedure and Design Criteria" Projects Procedure 2PP-5.23, " Piping Vibration Test Guideline" Projects Procedure 2PP-5.25, " Piping Thermal Growth Test Guideline" ASME Construction Procedure ACP-ll.5, " Component Support Fabrication and Installation," Revision 11

r;

,o.

o i

'

, Associated Data:

,

,

Preoperational Test Procedure 2CP-PT-90-01, " Steady-State Vibration Monitoring Test," Revision 1 Preoperational Test Procedure 2CP-PT-90-02, " Dynamic Transt'ent ' Response Testing," Revision 1 Preoperational Test Procedure 2CP-PT-55-09, "RCS Equipment Supports Thermal Expansion Test," Revision 0

'.

Station Procedure STA-742, " Snubber Surveillance Program," Revision 4

.

-2-

,