IR 05000445/1990002

Insp Repts 50-445/90-02 & 50-446/90-02 on 900103-0206.No Violations or Deviations Noted.Major Areas Inspected: Unannounced Resident Safety Insp of Applicant Action on Previous Insp Findings

ML20011F703
Person / Time
Site:	Comanche Peak
Issue date:	02/22/1990
From:	Bitter S, Johnson W, Vickrey R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20011F699	List: IR 05000445/1990002 ML20011F698
References
50-445-90-02, 50-445-90-2, 50-446-90-02, 50-446-90-2, NUDOCS 9003070211
Download: ML20011F703 (33)

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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR' REGULATION NRC Inspection Report:

50-445/90-02 Permits: CPPR-126 50-446/90-02 CPPR-127 Dockets: 50-445 Construction Permit 50-446 Expiration Date:

Unit 1: August 1, 1991 Unit 2: August 1, 1992 Applicant:

TU Electric Skyway Tower 400 North Olive Street Lock Box 81.

Dallas,. Texas 75201 Facility Name:~

Comanche Peak Steam Electric Station (CPSES),

Units 1 and 2 Inspection At Comanche Peak Site, Glen Rose, Texas Inspection Conducted:

January 3 through February 6, 1990 Inspectors:

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S.'D. Bitter, Reiidp t Inspector, Date Operations

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R. B. gichre'y, R# actor Inspector, Date

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Region IV

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Feb 22. /990 WT D. Joh)) hon, Schior Resident Inspector, Date Operatitns Reviewed byt rJ Al M!N90

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S. Wiebe, Senior Project Inspector

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-2-Inspection Summary Inspection Conducted) January 3 through February 6, 1990 (Report 50-445/90-02 50-446/90-02)

Areas Inspected:

Unannounced resident safety inspection of applicant action on previous inspection findings; in-office review of event reports (30 CFR Part 50.55[e]); onsite follow-up of event reports (10 CFR Part 50.55[e)); follow-up on violations / deviations; operctional safety verification; monthly maintenance observation; monthly surveillance observation; TMI action items (Safety Issue Management System Items I.C.1, open; I.C.6, closed; I.D.1, open; II.E.4.2.5, closed; II.E.4.2.6, open; II.F.1.4, closed; II.K.3.5.B, closed; III.D.3.4, closed); in-office review of event reports (10 CFR Part 21);' safety evaluation report (SER) follow-up; engineered safety feature system walkdown; initial fuel loading procedure review; and follow-up on NRC Bulletin 89-03.

Results:

Within the areas inspected, the applicant's programs and implementation of,those programs were adequate.

The applicant has made significant progress in preparing for fuel load.

The NRC review of the initial fuel loading procedure was completed with satisfactory results during this inspection period.

The resident inspectors initiated the routine operational safety inspection, the monthly observations of surveillance and maintenance, and the semi-annual walkdown of engineefed safety features.

The results of those inspections were siatisfactory with only minor discrepancies identified.

All of the discrepancies were promptly addressed by management.

The violation involving the applicant's failure to perform a review and approval of an instrument air system temporary

modification was closed.

This closure was based on the applicant's

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significant progress made in rectifying the problems with the Unit 1

temporary modification program.

During this inspection period, two

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open items were identified (paragraphs 2.h and 10).

The first involves the inspector's plans to review the applicant's incorporation of lessons learned in the-design of Unit 1 into Unit 2.

The second

involves review of a 10 CFR Part 21 evaluation.

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l DETAILS 1.

Persons Contacted

• M.

Axelrad, Newman and Holtzinger

• J.

L. Barker, Manager, ISEG, TU Electric j

• J. W. Beck, Vice President, Nuclear Engineering, TU Electric i

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Bhatty, Issue Interface Coordinator, TU Electric

• M.

R. Blevins, Manager of Nuclear Operations Support, i

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TU Electric

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Brau, Operations, TU Electric

• H.

D. Bruner, Senior Vice President, TU Electric

• J.

H. Buck, IAG

• R.

C. Byrd, Manager, Quality Control (QC), TU Electric I

• W.

J. Cahill, Executive Vice President, Nuclear, TU Electric

• H.

M. Carmichael, DBO EA Manager

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• J.

T. Conly, APE-Licensing, SWEC

• B.

S. Dacko, Licensing Engineer, TU Electric i

• D.

M. Ehat, Consultant, Bentham Group

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• D.

R. Ferguson,'NESG, TU Electric

• S.

P. Frantz, Newman and Holtzinger

• B.

P. Garde, Attorney, CASE

• W.

G.

Guldemond, Manager of Site Licensing, TU Electric

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• T.

L. Heatherly, Licensing Compliance Engineer, TU Electric

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C. Hicks, Site Licensing, TU Electric

• C.

B. Hogg, Chief Engineer, TU Electric

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A. Hope, Site Licensing, TU Electric

• J.

J. Kelley, Plant Manager, TU Electric

• J.

L. LaMarca, Manager of Electrical and I&C Engineering, TU Electric

• H.

Lawroski, Consultant, TU Electric

• D.

M. McAfee, Manager, QA, TU Electric

• J.

F. McMahon, Manager Nuclear Training, TU Electric

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• J. W. Muffett, Manager of Project Engineering, TU Electric

• E.

F. Ottney, Project Manager, CASE

+W.

Rosette, Operations, TU Electric

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• P.

Raysircar, Deputy Manager, Project Engineering, TU Electric

• M.

J. Riggs, Plant Evaluation Manager, Operations, TU Electric

• A.

D. Scott, Vice President, Nuclear Operations, TU Electric

+*J.

C. Smith,.alant Operations Staff, TU Electric

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• P.

B. Stevens, Manager of Operations Support Engineering, TU Electric

• J.

F. Streeter, Director, QA, TU Electric lh

• C, L. Terry, Manager of Projects, TU Electric

• T.

G. Tyler, Director, Management Services, TU Electric l

J. R. Walker, Operations / Engineering Training Manager, TU Electric g

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R. Waters, Site Licensing, TU Electric The NRC inspector also interviewed other applicant employees during this inspection period.

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-4-NRC personnel present at exit meetingat

• S.

D. Bitter, Resident Inspector

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D. Johnson, Senior Resident Inspector

• R. M. Latta, Resident Inspector

• H.

H. Livermore, Lead Senior Inspector

• M.

F. Runyan, Resident Inspecter 4*R.

B. Vickrey, Reactor Inspector, Region IV

• R.

F. Warnick, Assistant Director for Inspection Programs

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S. Wiebe, Senior Project Inspector

• Denotes personnel present at the February C, 1990, resident inspector's exit interview.

+ Denotes personnel present at the January 26, 1990, R. Vickrey's exit interview.

2.

Applicant Action on Previous Inspection Fin @inga,,(33,701)

a.

(Closed) Open Item (445/8407-0-01):

Disposition and tracking of_10 CFR Part 21 reports.

This item resulted from the inspector's concern that the applicant did not have an adequate procedure for controlling the dispc.'sition and tracking of 10 CFR Part 21 reporta.

The innpsetor concluded in NRC Inspection Report 50-445/44~07 that the applicant's current practice of using the signific5nt dcficiency analysis report (SDAR) system in accordance with 10 CFR Part 50.55(e) is satisfactory; however, upon Ideensing for operation, using 10 CFR Part 50.55(el will no longer be appropriate.

The inspector reviewed Nuclear Engineering and Operations Procedure NEO 2.01, " Identification, Evaluation and Reporting of Defects and Noncompliance Under 10 CFR Part 21," Revision 0.

NEO 2.01 is to become effective upon receipt of an operating license.

Step 5.2.1 of this procedure assigns responsibility to the manager, site licensing for coordinating evaluations of conditions to determine reportability under 10 CFR Part 21.

Step 6.1.5 of this procedure requires an individual who becomes aware via an outside organization, of a condition that is potentially reportable under 10 CFR Part 21, to immediately notify and I

transmit a copy of the identifying information to the

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manager, site licensing.

NEO 2.29, " Industry Operating Experience Review," (IOER)

Revision 0, effective August 23, 1988, Step 6.1.1 requires that all documents related to the IOER program (this includes incoming 10 CFR Part 21 reports) be transmitted to the industry operating experience coordinator (IOEC).

The IOEC forwards the documents to the appropriate responsible

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organization for evaluation and preparation of necessary actions, plans, and schedules.

Based on the above, the inspector concludes that the

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applicant's program for 10 CFR Part 21 implementation during

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the operations phase is adequate.

This item is considered closed.

b.

(Closed) Unresolved Item (445/8822-U-05):

Procedures

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implementing maintenance programs.

This item arose when the inspector determined that the maintenance and operation data

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system (MODS) did not accurately reflect the size, pressure

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rating, and other attributes for numerous installed flow orifices.

The inspector expressed concern that MODS data

might be used to determine quality classification or configuration status, f

The applicant'has responded to this by explaining that MODS l

has been replaced by the managed maintenance computer program (MMCP).

STA-403, " Identification of Quality-Related Equipment," specifically states that the MMCP is not to be used for determining quality classification,.for obtaining equipment listings, or for determining configuration status.

Instead, the master equipment list (MEL), controlled i

drawings, design basis documents, and Chapter 17 of the FSAR, are to be used.

The inspectot has reviewed the applicant's response and has no further questions.

There is no basis for a violation.

Therefore, this issue is closed.

c.

(Closed) Open Item (445/8937-0-03):

Inadequacies in plant

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paging / announcing system.

This item originated during

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preoperational testing.

The inspector, during his witnessing of several preoperational tests, noticed that the

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paging / announcing system seemed inadequate to alert all personnel in the event of an emergency.

Specifically, he noticed that the voice paging system was inaudible in certain locations.

In response to this open item, the applicant submitted the

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results of the test EGT-TP-89-5, "Gai-Tronics/ Emergency Alarm System Performance Test," for the inspector's review.

y The inspector focused his review on the test's requirement to " demonstrate proper functioning of the Public Address System including the capability of the voice paging channel output to be clearly audible over the highest expected noise levels."- Although not all locations initially tested

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received clearly audible voice output, after rework /

adjustment and retesting, only four locations remained a problem in this respect.

These locations are in areas of high background noise; thus, they were equipped i

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with " visual alarms" (rotating lights) to-permit adequate

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warning of emergency conditions to personnel.

j The inspector has no further questions.

This issue is

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closed.

d.

(Closed) Open Item 445/8940-0-02, "open Issues Involving i

Preoperational Test Matrices."

This item originated during

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a review of the test matrices that the applicant had developed for the preoperational. restart test program.

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During that review, the inspector developed a list of concerns that were to be addressed by the applicant prior to fuel load.

Each of these concerns, together with its resolution, is discussed in the following paragraphs.

(1)

1-0100, 1E DC Power System, was approved by the JTG on

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october 13, 1989.

Comparison of the system test matrix

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to all of the identified applicant commitments (FSAR, Regulatory Guides, etc.) did not identify any missed acceptance criteria.

One exception was noted and is to be addressed by the applicant prior to fuel load:

Derating of the plant batteries was performed after the original test had been completed;

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however, a revised engineering calculation stipulated that the new 225 amp rating (original rating was 300 amps) was still acceptable.

Subsequent performance of a special battery test, i

1-CP-SPT-43, helped to confirm this new rating;

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however, the special test was not included in the system data package as permanent plant test historical dats.

This special test should be f

included in the test data package.

i Resolution:

The applicant has committed to placing the

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record of Special Test 1-CP-SPT-43 in the TU Electric records vault as part of the test package.

(2)

1-0400, Service Water System, was approved by the JTG on October 10, 1989.

A review of all available acceptance criteria (FSAR, Regulatory Guides, commitment Tracking System, etc.) indicated that all were met for the testing performed, with one exception.

This exception is to be addressed prior to fuel load:

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Hydraulic data for service water pumps cpl-SWAPSW-01 and CP1-SWAPSW-02 obtained during prerequisite test XCP-ME-01 and 1CP-PT-04-01 SFT, Revision 0, does not appear to meet the acceptance criteri (F

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Resolution:

The inspector has reviewed the pump data l

and has concluded that it meets the acceptance criteria i

for both service water pumps.

The nonconformance

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report (NCR) that stated that the data did not meet the acceptance criteria should not have been written; it l

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was closed with.no action taken.

The inspector is

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satisfied that the acceptance criteria were met and has

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no further questions.

This issue is closed, j-t (3)

1-1100, Component Cooling Water System, was approved by l

the JTG on October 3, 1989, and appeared to address all

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of the acceptance criteria when compared to the-

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applicant's commitments.

One exception was noted and l.

is to be addressed by the applicant prior to fuel load:

The original criteria required the test to verify the design pressure and flow.

However, due to

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semantic problems, the pressure component of the i

acceptance criteria was deleted in a later revision of the acceptance criteria.

Resolution:

Originally,-the acceptance criteria contained in the FSAR (January 15, 1988) stated that the " Component Cooling Water System-pumps meet or exceed design flow and pressure requirements."

The applicant took the position that this seemed to imply that exceeding the system design pressure was acceptable.

To avoid drawing this conclusion, the

applicant intentionally did not specify " pressure" in the component cooling water (CCW) system test matrix.

This compounded the problem in that to the NRC

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inspector who reviewed the test matrix, it appeared that pressure was not considered in meeting the

acceptance criteria.

Thus, this issue became part of i

this open item.

The applicant has addressed this by submitting a licensing document change request (LDCR) that changes the acceptance criteria in the FSAR to state "The Component Cooling Water System pumps meet design requirements.

System flow and pressure requirements are satisfied."

Furthermore, the test matrix does specify that the CCW pump hydraulic performance is to be verified during the preoperational test.

The applicant has satisfied the inspector that the acceptance criteria have been correctly called out in the test matrix.

There are no further questions.

This issue is closed.

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1-2900, Diesel Generator System was approved by the JTG

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on October 10, 1989.

The inspectors reviewed the test l

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matrix package and compared the identified acceptance

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criteria with the required test commitments.

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review did.not find any undocumented acceptance criteria.

However, the following issues need to be

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The inspectors questioned whether the problems with welds on the diesel generator air start. tanks j

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had been considered during the test matrix review to resolve any concerns with invalidation of any previously performed testing.

r Resolution:

The inspector has determined that the weld repairs on the diesel generator. air start tanks had no effect on'the validity of previously performed diesel generator testing..The inspector has no further

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questions.- This issue is closed.

t Was all previously performed testing and any subsequent testing performed in accordance with the requirements of Regulatory Guides 1.68 and 1.1087

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Resolution:

The inspector has determined that the applicant has successfully completed all preoperational testing in accordance with Regulatory _ Guides 1.68 and 1.108.

The inspector has no further questions.

This issue is closed.

once the preoperational testing requirements were satisfied in 1984, were the requirements of Regulatory Guide 1.108 fulfilled (maintaining the diesel generator operability once testing was completed, i.e.,

surveillance, routine-maintenance, post-work retest program, etc.)?

Resolution:

The requirements of Regulatory Guide 1.108 for maintaining the diesel generator (DG) operability after completion-of testing do not apply until the plant is in normal operation.

The applicant's

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surveillance program ensures that all diesel generator operability testing will be performed prior to i

considering it operable.

The inspector has no further questions.

This issue is closed.

l (5)

1-4800, containment Spray System, was approved by the JTG on October 13, 1989.

The results of a review of the submitted test matrix as compared to the various commitments, regulations, and acceptance criteria appear to be acceptable, with one exception:

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The applicant needs to ensure that the vacuum

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breakers are actuated (operated) and found to be i

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acceptable prior to entry into Mode 4.

TU Electric Work Orders C-890012763 and C-890012764 involve this issue.

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Resolution:

The applicant submitted the completed Work Orders C-890012763 and C-890012764 for the inspector's

review.. These work orders indicate that the vacuum breakers have been manually actuated (operated) to verify opening and closing capabilities.

The inspector.

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has no further questions.

This issue is closed.

(6)

1-4900, Chemical and Volume control-System was JTG

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approved on October 19, 1989.

After review of the test matrix, there are no outstanding comments or concerns.

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Resolution: 'Nono required.

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(7)

1-5500, Reactor Coolant System, was approved by the JTG on August 10, 1989.

Review of the various commitments, regulations, and acceptance criteria appears to be acceptable except for tho hydrostatic test issue involving the reactor coolant system.

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The reactor coolant system hydrostatic test issues was addressed in NRC Inspection Report 50-445/89-13; 50-446/89-13.

Resolution:L None required.

(8)

1-5700, Safety Injection System, was. approved by the JTG on October 5, 1989.

The inspectors reviewed the submitted test matrix and compared it to the identified commitments, regulations,.and acceptance criteria.

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results of the review appear to be acceptable except as

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noted:

l Verification, completion, and ecceptability of all deferred test acceptance criteria need to be performed prior to acceptance of this test matrix.

This must be done prior to fuel load.

These deferrals are listed on Tables I and IA as y

attachments to the test matrix package.

Resolution:

The inspector has reviewed the latest revision of the test matrix.

This review indicates that all the valves listed on Tables I and IA have been i

tested and the results have been approved by the JTG.

The inspector has no further questions.

This issue is

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closed.

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J Verification of test completion and acceptability

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for the following valves needs to be performed prior to acceptance of~this test matrix.

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1-8818A.

Work Order C-890006673

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1-8818B-Work Order C-890006672 l

1-8818C Work Order C-890006549

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1-8818D Work Order C-890006550

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1-8841A Work Order C-890006551

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Resolution:

The applicant has requested deferral (TXX-89824;, December 6, 1989) of preoperational testing

for these five valves post-fuel load.

The inspector

has no further questions.

This issue is closed.

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The test matrix should include the referenced

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system flow balancing data performed during the Unit 1 prestart test program because this data

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forms a basis for the acceptability of the overall system package.

Resolution:

Test Matrix 1-5700 documents that the flow

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balancing data is located in Preoperational

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Tests 1 CP-PT-57-01, -02, -06, and in 1CP-PT-58-01.

In all cases, for each of these preoperational tests, the

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acceptance criteria steps are listed.

The inspector has no further questions.

This issue is closed.

The inspectors noted that the applicant used the Inservice Testing Program, ASME Section XI requirements as part of the justification for valve testing or post-work testing.

It was not clear to'the inspectors as to what portions of the

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test matrix this item was verifying.

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applicant needs to explain this in more detail.

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Resolution:

Test Matrix 1-5700 references TU Electric letter TCF-891807 of October 2, 1989.

This letter lists the valves for which TU Electric intends to apply the ASME Section XI requirements and alternatives toward meeting the preoperational test requirements.

Furthermore, these valves have been deferred per TU Electric letter 89824 of December 6, 1989.

The I-inspector has no further questions.

This issue is closed.

I (9)

1-5800, Residual Heat Removal (RHR) System, was approved by the JTG or. August 10, 1989.

A review of

the submitted test matrix package did not identify any outstanding issues.

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I Resolution:

None required.

(10) 1-6400, Reactor Protection System, was approved by the JTG on October 13, 1989.

The submitted test matrix was

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found to be acceptable with the following exceptions:

j All Post-802 testing must be given a scheduled completion date and performed prior to fuel load.

-Resolution:

The inspector has verified that all

Reactor Protection System (1-6400) Post-802 deferred items have been tested with two exceptions.

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exceptions are:

(a)

The response. time of.the turbine-driven. auxiliary feedwater pump must be reverified because a design

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change was performed on the steam admission

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valves.

This testing.is to be performed during

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Mode 3 following fuel load.

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The response times of the main steam isolation

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valves (MSIVs) must be reverified because of

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extensive maintenance that was' performed on the i

MSIVs following the original MSIV preoperational testing.

This testing.is to be performed during l

Mode 3 following fuel load.

The inspectors request.that the applicant clarify, i

prior to fuel load, the changes to CP-64-08 SFT,

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Revision 0, identified as Temporary Procedure Change (TPC) No. 1, Sheets 1 and 2 of the test.

matrix attachments.

Resolution:

TPC No. 1 to 1-CP-PT-64-08 SFT, Revision 0, " Reactor Protection System," deleted the

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requirement to transcribe, into 1-CP-PT-64-08, three t

Fets of response time data recorded during the Conduct

of various other preoperational tests.

In all three cases, the original data was invalid because equipment malfunctions occurred during the testing, or because major maintenance was performed after the test had been conducted.

Deleting the three sets of data from 1-CP-PT-64-08 SFT enabled the applicant to evaluate, g

close, and approve that preoperational test.

Two of the three sets of deferred data were regathered during various tests conducted via the Performance and Test Program.

The gr.thering of the one remaining set of data was deferred past fuel load.

The three cases are explained in detail below:

(a)

The motor-driven auxiliary feedwater (MDAFW) pump response times recorded in Preoperational

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i-Test 1-CP-PT-37-01 SFT were invalid because of a i

flow regulating valve malfunction.

Therefore, the i

data was not transcribed into 1-CP-PT-64-08 SFT; i

instead, the test that obtains this data was i

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deferred past the turnover of the auxiliary feedwater system to operations.

As of now, this

'l data has been collected via the performance of EGT-720A.-

The data is now being evaluated.

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The turbine-driven auxiliary feedwater (TDAFW)

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pump response time recorded during the performance

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of 1-CP-PT-37-03 was invalid because the steam

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admission valve underwent'a design change following the original' testing.

TU Electric letter TXX-89824 of December 6, 1989, document the

applicant's request to defer the testing that j

gathers this data until power ascension testing.

i (b)

The main steam isolation valve (MSIV)' response

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times recorded in 1-CP-PT-34-01 SFT were invalid because of maintenance that was performed on the MSIVs after completion of the hot functional test program.

TU Electric letter TXX-89824 of December 6, 1989, documents the applicant's request to defer the testing that gathers this data until power ascension testing.

(c)

The response times for the 6.9kv and 480V

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undervoltage relay actuation were unable to be

obtained during the Integrated Sequence (ITS)

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because of scheduling difficulties.

In the meantime, this data has been gathered by testing conducted by the site metering and relay group.

The data is currently being evaluated by the

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Performance and Test Group.

(11) 1-9502, Integrated Hot Functional Test, was approved by the JTG on October 5, 1989.

The inspectors reviewed the submitted test matrix package and did not find any significant deficiencies, except as noted:

The applicant must complete all required testing

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(including deferred testing, i.e.,

post-802

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testing) prior to fuel load.

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Resolution:

The inspector has verified that there is

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only one outstanding item to complete.

This is the testing that demonstrates the effectiveness of l

pressurizer spray.

The applicant submitted, and the l

NRC approved, an FSAR amendment (No. 78, January 15, 1990) that states that the pressurizer spray effectiveness will be demonstrated during power L

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ascension.

The inspector has no further questions concerning this issue.

Verification that the reactor coolant pump seal injection flow must operate.against full reactor coolant flow.

The applicant must verify, prior.to fuel load, that this acceptance criteria was met.

Resolution:

The applicant has submitted documentation that explains how this acceptance criteria was met.

Specifically, charging flow and seal injection flow to

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reactor coolant pumps (RCPs) 1, 2, 3,'and.4 was

recorded on Data Sheet 2 of Preoperational Test i

1-CP-PT-49-02,. Revision 1, "Sealwater and Letdown Flow I

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Performance."

Thisdatawasrecordedwithtgereactor l

coolant _ system pressure at 2235 psig and 557 and the

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RCPs were operating.

Therefore, the data was obtained,.

as required, with normal reactor coolant pressure.

The inspector has reviewed this data and has no further questions.

This issue is closed..

These eleven issues have been satisfactory' addressed by the applicant.

The inspector has no further questions.

This I

item is closed.

c.

(closed) Open Item (445/8945-0-02):

Program for control of temporary and-permanently installed hoists.

This item originated from the inspector's concern over the locations of a temporary hoist and a permanently installed hoist in the diesel generator room while the diesel generator was operating for preoperational testing.

The inspector

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verified that Maintenance Department Administrative

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Procedure MDA-402, " Control of Handling and Holsting Equipment," Step 6.2.2.8, requires personnel to return portable hoisting and handling equipment to its assigned storage location after use.

The inspector also notes that the use and storage of permanent and temporary nonplant equipment is controlled by Station Administrative Procedure STA-661, "Non-Plant Equipment Storage and Use Inside seismic Category I Structures."

STA-661 provides instructions to ensure the nonplant equipment does not affect safety-related equipment.

g To ensure that the permanently installed hoists can not affect safety-related equipment, the applicant has initiated actions to provide for parking locations and bo't-on stops based on seismic interaction analysis.

The inspector understands that these actions will be completed prior to entering the plant mode that requires the affected component or system to be operable.

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closed.

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(closed) Open Item (445/8959-0-03):

Resolution of engineering identified inconsistencies before fuel load.

This item related to engineering identified inconsistencies related to engineering review of cmsrgency response guidelines (ERGS) to determine actual or potential

~ differences between the ERGS and the applicable design-basis documents.

At the time of NRC Inspection Report 50-445/89-59; 50-445/89-59, approximately half of the inconsistencies that required further. review and evaluation had been resolved.

The inspector verified that the licensee had completed the review, evaluation, and resolution of:the above engineering identified inconsistencies of ERGS.

This verification included a random sample rev!.ew of 20 which had been completed.

No discrepancies were identified by the i

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inspector.

The inspector discussed with engineering their future involvement, timeliness, and accuracy for ERG review.

Engineering admitted that they had not initially realized the level of detail required for ERG review but felt that they now had established better methods and references for future reviews.

The'multidiscipline review of ERGS-now includes an engineering review.

This item is considered closed.

g.

(Closed) Open Item (445/8959-0-05):

Correction of control

' room meter markings.

This item concerned perameter errors

.and disagreements in meter markings identified by the NRC

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during review of EOP 0.0, " Reactor Trip cr Safety Injection."

The applicant.had corrected the seven errors and disagreements identified in the inspection report.

The i

applicant identified several other errors and disagreements on the control board indicators and remote shutdown panel during review of corresponding surveillance procedures.

Correction of Technical Specification related items was nearly complete with some new meter scales still on order.

The applicant's progress and status in this area appears satisfactory.

This item is considered closed.

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Applicant action on other comments and commitments.

NRC

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I Inspection Report 50-445/89-59; 50-446/89-59 contained I

several comments and commitments not directly tracked as open items.

These items consisted of approximately 20 issues, 50 technical comments and 35 human factor comments.

The inspector followed-up on approximately 50% of the above issues and found the applicant to be making timely and satisfactory results with respect to their resolution.

The licensee also appeared to be making satisfactory I

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progress towards the resolution of the unresolved item and remaining four open items.

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(closed) open Item (445/8961-0-01):

Operational input to setpoint calculation proceas.

This item resulted from the

inspector's concern that the operators would not be able to

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take manual control of the auxiliary feedwater (AFW) pump i

discharge valvas following an automatic start signal becauce y

the reset setpoints were adjusted such that reset could not

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occur.

As a result, the operators would not be able to

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i control steam generator level or cool-down rate.

The applicant provided additional information which showed

that the operators would have controlled steam' generator Y

level and cool-down rate by isolating the pump (s) when sufficient steam generator level had-been established.

Although this type of design is awkward and appears to be

void of human factors considerations, the inspector i

acknowledges that the operators could safely control the plant cool-down in this fashion.

The inspector also notes that the plant operations staff

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requested a change to this design to allow them to take

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manual control of the discharge valves and, therefore, have

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better control over plant cooldown and steam generator level.

This change was considered by Engineering to be an enhancement, but was included in a modification that was

required prior to obtaining an operating license.

Since the

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' modification was implemented, this item is-considered closed.

However, the lack of an operations perspective, human factors considerations, and a maintenance perspective is displayed by the design and construction engineering

organizations in this example, as well as in the design and construction of manual valves reach rods (NRC Inspection Report 50-445/89-30; 50-446/89-30), and other examples.

l This is of concern because, collectively, if sufficiently extensive, such items could cause a decrease in the safety of the plant even though individually such items may be considered as enhancements for ease of' operations.

This issue is open pending inspector review of the applicant's plans to inject an operations perspective, human factors

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considerations, and maintenance perspective into the design and construction engineering organizations for Unit 2.

The inspector also will review how lessons learned on Unit 1 are factored into the Unit 2 design (446/9002-0-01).

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(closed) open Item (445/8983-o-02):

Power operated relief valve (PORV) closing time.

This issue originated from the inspector's concern that the "as-found" PORV closing time of l

3.35 seconds exceeded the 2 second closing time assumed in

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the setpoint analysis.

In response to this concern, the L

applicant requested Westinghouse to evaluate the effect of

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.w the longer closing time.

The Westinghouse evaluation showed that the resulting pressure undershoot will not cause

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violations of the. reactor coolant pump (RCP) seal pressure limit as long as the PORV closure time is less than

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3.5 seconds.

Since this snvelopes the "as-found" value, the inspector considers this issue closed.

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3.

In-Office Review of Event Reports (10 CFR Part 50.55(e]) (90712)

P The inspector reviewed significant deficiency analysis reports

(SDARs) to dotermine if additional reactive inspection effort or other NRC response is warranted, if corrective action discussed t

in the applicant's report appears appropriate,.if the information a

satisfies reporting requirements, if generic issues are present,

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if the event is appropriate for citiusification as an. abnormal

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occurrence pursuant to Section 208 of the Energy Reorganization

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Act of 1974, and if-the event is appropriate for the licensing board, the appeal panel, or commission notification.

SDAR CP-89-024, " Emergency Diesel Generator (EDG) Turbocharger Bolt Failure," (Units 1 and 2) was determined to not warrant

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additional NRC action and is considered closed.

4.

Onsite Follow-Up of Event Reports (10 CFR Part 50.55(e])

(92700)

a.

(Closed) Construction Deficiency (SDAR CP-86-18):

" Safety Chilled Water Units."

This item originated when the applicant evaluated numerous difficulties in starting and i

operating the safety chilled water chiller units when the entering component coolant water (CCW) temperature was not

maintained at, or above, the chilled water return

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temperature.

The applicant reported this significant

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- deficiency under the provisions of 10 CFR Part 50.55(e).

As corrective action, the applicant has performed the following:

(1)

Installed an air-operated water regulating valve in the

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CCW outlet from the chiller condenser along with a pressure controller to maintain condenser refrigerant pressure above evaporator pressure.

)

(2)

Prevented CCW from circulating through the chiller condenser prior to condenser startup.

(3)

Installed a 30 minute air accumulator tank for the air-operated water regulating valve and installed a hand wheel to manual operate this valve after 30 minutes if instrument air is not available.

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(4)- Installed similar modifications for the nonsafety

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chillers (excluding the 30 minute air accumulator).

(5)

Installed a bypass _line to bypass the instrument air i

compressor jacket after-cooler.

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The inspector had left this item open pending the NRC's

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verification that plant operating and surveillance procedures adequately addressed these modifications.

During

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this inspection period, the inspector reviewed the following:

ABN-301A Revision 3

" Instrument Air System Malfunction" ALM-0041A Revision 3

" Alarm Procedure 1-ALB-4A"

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ODA-301-07 Revision 2

" Equipment Log-Auxiliary and i

Fuel Building - Common" OPT-209A Revision 1

" Safety Chilled Water System Operability Test" SOP-502A Revision 6

" Component Cooling Water System"

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SOP-815A'

Revision 5

" Safety Chilled Water System" t

The inspector has determined that these procedures adequately address the concerns involving the modifications.

There are no further questions.

This item is closed.

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(Closed) Construction Deficiency (CP-86-26):

" Containment s

Spray System Piping."

By letter from the Gibbs and Hill

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Vice President of Quality. Assurance, dated March 17, 1986, the NRC was notified of a.nonconformance or defect under-l 10 CFR Part 21.

The letter stated that the containment j

spray system recirculation piping is classified as non-nuclear safety.

If this piping failed in an accident, there is'a possibility that the containment spray. pumps would runout, overload their motors, and trip.

By letteriTXX-4799 dated May 14, 1986, the applicant informed the NRC that the piping classification is in g

conformance with the regulatory requirements.

Furthermore, the applicant performed an analysis that demonstrated that if the piping failed, the pumps would not run out.

Therefore, the applicant determined that this item was not reportable and no corrective action was needed.

NRC Inspection Report 50-445/89-65; 50-446/89-65 addressed

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SDAR CP-87-55, " Containment Spray Pump Recirculation Piping."

SDAR CP-87-55 identified the same deficiency n

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18-(i.e., the piping and supports were installed as.non-nuclear safety piping).

The' applicant informed the NRC in letter TXX-7046 dated December 8, 1987, that the FSAR and ANSI MB.2 require this piping to be designed and installed as ASME Class 2 piping.

As a result, the deficiency was considered reportable.

By internal letter NE-27362 dated July 27, 1989, the utility documented an' attempt to resolve the differences in the evaluations of SDAR Cp-86-26, and SDAR CP-87-55.

This letter concluded that both evaluations were accurate and acceptable.

The inspector, while questioning the premise that the piping system would both be in conformance with FSAR commitments (SDAR-86-26) and not be in conformance with FSAR commitments'(SDAR-87-55),

recognizes that the second SDAR was identified as a result of the corrective action program which was developed to identify such problems.

In addition, the piping was upgraded in response to the second SDAR.

As a result, the inspector determined that it is not worthwhile to pursue the questions concerning the adequacy of the first SDAR's evaluation.

In addition, the findings could not be representative of the current plant environment.

The inspector, therefore, considers SDAR CP-86-26 closed.

5.

Follow-up on Violations / Deviations (92702)

(Closed) Violation (445/8972-V-02):

Failure to perform the review and approval of a temporary modification (TM) on the instrument air system as required by Procedure STA-602.

A TM had been installed at the boundary of the Unit 1/ Unit 2 interface of the instrument air system to provide instrument air from Unit 1 to certain Unit 2 components.

The instrument air system was turned over to Operations with the TM still installed, but without performing the required reviews and approvals of the TM.

This TM had been incorrectly identified as a Unit 2 TM and was not included in the Unit 1 instrument air turnover package.

In response to this discrepancy, the applicant removed this TM and reviewed other existing TMs to identify any impacting Unit 1 systems.

This review identified two other TMs.

Training for system engineers was conducted and a review was performed to ensure that adequate controls were in place to prevent

inadvertent interaction with Unit 1 systems by personnel working

on Unit 2 components, g

During recent months, the applicant has greatly reduced the number of TMs installed on Unit 1.

The remaining 10 TMs have had a safety evaluation performed and have no discrepancies.

This

violation is closed.

Another violation related to Unit 1 TMs (445/8986-V-01) remains open.

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6.

operational Safety Verification (71707)

The inspectors routinely toured the facility during normal and backshift hours to assess general plant and equipment conditions, housekeeping, and adherence to fire protection, security, and i

i radiological control measures.

Ongoing work activities were

monitored to verify that'they were being conducted in accordance

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with approved administrative and technical procedures and that i

proper communications with the control room staff had been established.

The inspector observed valve, instrument, and i

electrical equipment lineups in the field to ensure that they

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were consistent with system operability requirements and

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operating procedures.

During tours of the control room, the inspectors verified proper staffing, access control, and operator attentiveness.

Adherence

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to procedures and limiting conditions for operations were

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evaluated.

The inspectors examined equipment lineup and operability, instrument traces, and status of control room

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annunciators.

Various control room logs and other available licensee documentation were reviewed.

The inspectors observed and reviewed maintenance and problem investigation activities to verify compliance with regulations and procedures.

Involvement

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of QA/QC, safety tag use, personnel qualifications, fire protection precautions, retest requirements, and reportability were assessed.

• Checks were made to determine whether security conditions met regulatory requirements, the physical security plan, and approved procedures.

Areas checked included security staffing, protected

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and vital area barriers, personnel' identification, access control, badging, and compensatory measures when required.

On Janv.ary 9, 1990, the Train B emergency diesel generator was tagged out for maintenanco under Clearance 1-90-0082.

The NRC inspector checked several of the components tagged under this clearance to verify that they were in the proper position.

The fuel oil booster pump handswitch on the diesel engine control panel was in the " AUTO" position although the attached danger tag

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required its position to be "OFF."

The unit supervisor was informed and the switch position was promptly corrected.

Since TU Electric typically tags both a switch and its associated breaker, and because the associated motor control center breaker

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kg was tagged open in this case, the safety significance of the mispositioned switch was minimal.

The applicant responded well to the mispositioned switch.

Actions completed included:

Performing an investigation to determine the cause of

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established the clearance.

Promptly performing a walkdown verification of the

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proper status of all clearances on safety-related electrical equipment and engineered safety feature

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equipment.

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Performing a 100 percent shift operations audit of E

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Unit 1 and Unit 2 clearances on January 18, 1990.

l Performing a 100 percent'radwaste group audit of

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clearances on radwaste systems.

Performing a human factors review of the diesel

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generator engine control. panel.

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Providing lessons learned memos to the maintenance and I

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operations groups.

No additional mispositioning discrepancies were identified during the above clearance audits.

The applicant concluded that the switch was most likely mispositioned by inadvertent bumping.

The NRC inspector agreed with this conclusion.

The lessons learned memos should help prevent future similar events.

Routine clearance audits by operators are intended to identify clearance problems.

During plant tours, the inspectors noted several minor discrepancies such as vent or drain valve leaks.

These were identified to applicant representatives for initiation of corrective action.

On January 9, 1990, the inspector noticed that the breaker for the Train B emergency diesel generator (EDG) space heater was tripped.

Approximately a week later, the corresponding breaker for the Train A EDG space heater was noticed as being tripped.

Applicant representatives initiated a ONE form to evaluate these spurious breaker trips.

The inspector determined that this action was adequate to assure resolution of the problem.

g on January 9, 1990, the inspector attended a System Operability Review meeting.

Recently, this type of meeting has been held on a daily (approximately) basis to review system readiness.

At this meeting a small group of electrical systems was reviewed by plant management to verify that the systems were ready to be declared operable.

On January 12, 1990, the inspector attended a regular meeting of

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the Station Operations Review Committee (SORC).

At this meeting i

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-21-the SORC reviewed and discussed 10 CFR 50.59 safety evaluations, a task team report on the inadvertent safety injection initiation of December 7, 1989, procedure changes, and SORC action item status.

A plant employee informed the inspector that he had left a concern with SAFETEAM relating to what he considered a " cheat

sheet" in the radiation count room.

Based on the employee description, the " cheat sheet" appears to have been a handwritten informal procedure-for operating the counting equipment.

The Operations Department response to SAFETEAM stated that no " cheat sheet" could be located; therefore, SAFETEAM considered the matter closed.

The inspector discussed this issue with the radiation protection manager who confirmed'that no informal procedures should be in use.

Instead, the equipment procedure should be used; or, if necessary, an approved posted operating procedure should be used.

The inspector considers this matter closed.

No violations or deviations were identified.

7.

Monthly Maintenance observation (62703)

Station maintenance activities for the safety-related systems and components listed below were observed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with

.the draft Technical Specifications.

The following items were considered during this review:

the limiting conditions for operation were met while components or systems were removed from servico, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, quality control records were maintained, activities were accomplished by qualified personnel, parts and materials used were properly certified, radiological and fire prevention controls were implemented.

Work requests were reviewed to determine the status of outstanding jobs and to ensure that priority is assigned to safety-related equipment maintenance which may affect system g

performance.

Maintenance activities observed included:

Calibration of LI-2478B, the condensate storage tank

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level indicator at the remote shutdown panel (Work Order C00000168).

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Opening and inspection of the electro-hydraulic i

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converter for the Unit 1 main turbine.

Repair of tube leaks on secondary system feedwater

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heaters.

Repair to component cooling water (CCW) relief l

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valve 100-1063.

j No violations or deviations were identified.

i 8.

Monthly surveillance observation (61726)

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t The NRC inspector observed the Technical Specification required surveillance testing.on the various components listed below.

The l

inspector verified testing was performed in accordance with

adequate procedures, test instrumentation was calibrated, limiting conditions for operation were met, removal and

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restoration of the affected components were accomplished, test

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results conformed with Technical Specifications and procedure

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requirements, test results were reviewed by personnel other than the individual directing the test, and any deficiencies

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identified during the testing were properly reviewed and resolved by appropriate management personnel.

The NRC inspector witnessed portions of the following test activities:

Monthly operability test of the Train B EDG

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(Procedure OPT-214A, Work Order S890001376).

EDG field voltage waveform check

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(Procedure MSE-Pl-0872-13, Work Order C890017440).

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Analog channel operational test on RWST level

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u Channel 0933 (Procedure INC7881A, Work i

Request S89-1525).

During this test, the inspector

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noted that the connunications interf ace between the r

Instrumentation and Control (I&C) technicians and the L

operators was very effective.

No violations or deviations were identified, hg 9.

TMI Action Items (SIMS)* (25565)

• The Safety Issue Management System (SIMS) tracking number is the same as the TMI Action Item number.

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a.

(Open) TMI Action Item I.C.1, "Short Term Accident and Procedure Review."

This item includes Temporary Instruction (TI) 2515/65 items I.C.1.1, "Small-Break LOCA," I.C.l.2.B,

Inadequate Core Cooling," and I.C. l. 3.B, " Transients and l

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Accidents."

TI 2515/65 requires review of the procedures and verification of their implementation.

NRC Inspection

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Report 50-445/89-59; 50-446/89-59 documented the NRC inspection of emergency operating procedures in accordance with TI 2515/92, " Emergency Operating Procedures Team inspection."

The inspector, therefore, considers the i

inspection required by TI 2515/65 to be complete.

The insoector notes, however, that the associated SIMS item is

still open.

The item awaits completion of the description

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of the program for conducting the plant function and task i

analysis.

As documented in supplement safety evaluation

report (SSER) 21, the results are to be submitted to the NRC j

staff before startup for the second operating cycle of j

Unit 1.

1 b.

(Closed) TMI Action Item I.C.6:

" Verify correct Performance of Operating Activities."

This item requires the applicant to have procedures that include an effective system for verifying the correct performance of operating activities.

l NUREG-0737, " Clarification of TMI. Action Plan Requirements,"

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contains the guidelines as to how to meet the requirements of this item.

Furthermore, the NRC staff has addressed the applicant's progress in meeting these requirements in NUREG-0797, "CPSES SER," and Supplement I to the SER.

Prior to the current inspection period, all NRC concerns, with the exception of two, had been resolved.

These two

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concerns were:

Which personnel are designated as being qualified to

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perform independent verifications (second checks) for tagging, return-to-service, etc?

l Do plant procedures contain sufficient controls to

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ensure that all surveillance procedures will contain steps that direct a " qualified operator" to perform independent verifications (second checks) of components priot' to returning a component (system) to service.

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During this inspection period, these two concerns were addressed to the NRC's satisfaction.

With respect to the first concern, the applicant has submitted an FSAR change request that allows qualified personnel other than licensed g

operators to perform independent verifications.

The inspector is satisfied that the qualified personnel are adequately trained to conduct independent verific&tions.

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The inspector reviewed several procedures that control independent verifications.

Specifically, ODA-404, Revision 3, " Guideline on Component Positioning and Independent Verification," has been changed to direct organizations other than Operations to perform independent

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verifications as'necessary to support system lineups specified by plant operating procedures.

It should be noted q

that during this inspection period, an NRC Operational

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Readiness Assessment Team (ORAT) identified a potential violation involving the implementation of.this policy by the instrumentation and control (I and C) organization.

The

inspector has reviewed the applicant's response to this issue as part of the programmatic review that involves this concern.

Implementation of the independent verification policy will be reviewed as a: follow-up to the ORAT's

inspection report.

To resolve the second concern, the inspector reviewed

ODA 207-1, Revision 3, " Procedure Verification Checklist."

j This checklist keys the procedure reviewer to consider the

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need for specifying an independent verification.

The inspector is satisfied that both concerns have been adequately addressed.

There are no further questions.

This item is closed.

c.

(open) TMI Action Item I.D.1, " Control Room Design Review."

l This item has been partially addressed in NRC Inspection Reports 50-445/89-02; 50-446/89-02 and 50-445/89-40;

50-446/89-40.

NRC Inspection Report 50-445/89-40;

50-446/89-40 identified two issues which remained open.

The status of these two issues and an additional issue are

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discussed below.

(1)

The human engineering deficiency (HED) (Control No. 354) concerning high temperature in the remote shutdown area requires additional evaluation of the temperature, humidity, and airflow to determine

corrective action.

This evaluation can not be

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performed until data is taken when the plant is hot and operating at high power during the hot, humid summer

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months.

The inspectors understand that the evaluation will be performed during the first summer in which the plant is operating at high power.

(2)

The HEDs that require environmental surveys were still open because the corrective action was not yet-complete.

The applicant has now completed the

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g corrective action-and performed the environmental surveys.

The temperature and humidity surveys were completed with minor deficiencies which the applicant determined do not affect the validity of the test.

The inspector agrees with the applicant's determination and has no further questions concerning the temperature and humidity surveys.

The lighting and noise surveys showed that additional corrective actions may be

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necessary.

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.i By letter TXX-90005, from William J. Cahill, Jr. to the NRC, the applicant committed to complete the final

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control room environmental survey and provide to the

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NRC an action plan to implement any necessary corrective. actions prior to exceeding 5 percent power

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for Unit 1.

(3)

Also, by letter TXX-90005, the applicant committed to review the emergency operating procedures to verify that the parameters required by procedures are L

consistent with the actual display parameters (e.g.,

terminology and labeling).

This review-is to be performed and a report submitced to the NRC prior to-exceeding 5 percent power for Unit 1.

t This TMI item remains open pending' applicant's resolution and NRC review of the items identified above.

d.

(Closed) TMI Action Item II.E.4.2.5, " Containment Pressure Setpoint."

NRC Inspection Report 50-445/89-17; 50-446/89-17 left this item open because of an apparent conflict between the Final Safety Analysis Reoort (FSAR) setpoint and the i

setpoint specified in the Technical Specifications.

The inspector verified that the revised FSAR and the revised Technical Specifications now list the containment pressure setpoint at 3.2 psig.

The inspector considers this item closed, e.

(Open) TMI Action Item II.E.4.2.6:

Containment Purge Valves.

This item was addressed in NRC Inspection Report 50-445/89-17; 50-446/89-17 and had remained open pending resolution of the operability of the 18-inch Containment Pressure Relief System valves and pending

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implementation of the applicant's commitment to include in

the Technical Specifications a limitation on the use of the Containment Pressure Relief System to less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per y

year during plant operating Modes 1, 2, 3, and 4.

The NRC has reviewed the analysis contained in the applicant's December 16, 1985, letter and the technical evaluation report dated October 5, 1989.

The NRC concluded that these documents demonstrate the ability of the 18-inch purge valves to close from a maximum opening of 65 degrees against the rise in containment pressure following a postulated accident.

The NRC, therefore, concludes that the operability requirements are satisfied.

As a result, the limitation on the use of this system to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year is no longer required.

The inspector verified that the FSAR l

commitment was removed.

This item remains open pending inspector verification thag the 18-inch valves are limited to a maximum opening of 65.

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(Closed) TMI Action Item'II.F.1.4, " Containment Pressure."

f This item was discussed and closed in NRC Inspection Report

50-445/89-67; 50-446/89-67.

This item is'being readdressed

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to clarify the applicant's action with regard to ensuring these instruments read correctly, i

The wide range pressure instruments are not used for manual

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operator actions, but are used to monitor for potential containment failure.

They are, therefore,' classified as

Type C variables and are not required to be' listed in the Technical Specifications.

The inspector = verified that Procedures INC-7857A and INC-7858A contain the channel t

I calibration for the two wide range instruments.

The applicant has informed the inspector that the calibration

and channel check intervals will be the same as those

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specified-in the Technical Specifications for the intermediate range containment pressure instruments.

This item is still considered closed.

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(Closed) TMI Action Item II.K.3.5.B:

Modification for

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automatic trip of reactor coolant pumps (RCPs).

This issue remained open in NRC Inspection Report 50-445/89-24; 50-446/89-24 pending NRC review of supplementary information, completion of applicant action, and subsequent

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NRC review in accordance with Temporary Instruction 2515/65,

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"TMI Action Plan Inspection Followup."

The applicant has adopted the Westinghouse Owners Group-(WOG) methodology that

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was approved by the NRC in Generic Letter 85-12.

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The NRC.has concluded that the WOG methodology has

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significantly improved reactor safety and that there are no major safety-significant concerns with the CPSES plant-specific analysis submitted by letter dated September 24, 1986.

As a result, the issue is considered satisfactorily resolved.

Since no modification was required

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to resolve this issue, the inspector considers the inspection required by TI 2515/65 to be complete.

h.

(Closed) TMI Action Item III.D.3.4, " Control Room Habitability / Modification."

This item is listed in

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Temporary Instruction 2515/65, "TMI Action Plan Followup";

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however, the inspection instructions state that no

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inspection is required.

This item is considered closed.

No violations or deviations were identified.

10.

In-Office Review of Event Reports (10 CFR Part 21)

(90712)

The inspector reviewed 10 CFR Part 21 reports to determine if additional reactive inspection effort or other NRC response is

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warranted; if corrective action discussed in the report appears appropriate; if the information satisfies reporting requirements;

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if generic issues are present; if the event is appropriate for classification as an abnormal occurrence pursuant to Section 208 of the Energy Reorganization Act of 1974; and if the event is appropriate for licensing boerd, appeal panel, or commission notification.

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The applicant's evaluations for the following reports with Region IV tracking numbers were reviewed and determined not to warrant additional NRC actiont

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RIV Tracking Related Document i

No.

Company Date or-Remark

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86-07 Validyne Engineering 09-17-86 N/A 86-13 Foxboro Company 10-07-86 N/A

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86-14 Westinghouse CP-85-17 12-15-86 Addressed 87-03 Foxboro Company 06-04-86 N/A i

87-11 General Electric 11-17-86 N/A l

87-21 United Engineers and 12-17-86 N/A l

Constructors L

87-28 Niagara Mohawk 01-26-87 N/A 87-30 Niagara Mohawk 02-02-87 N/A 87-38 Morrison Knudsen 01-13-87 N/A 87-42 Basler Electric 11-25-86 N/A

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87-46 Isomed IX 03-30-87 Addressed l

87-51 Static O-Ring 04-27-87 N/A 87-54 System Energy Sources 04-09-87 N/A 87-55 Southern California 05-07-87 Addressed l

Edison 87-64 Limitorque 07-27-87 N/A 87-76 General Electric 11-12-87 DR-C87-4859 87-80 Public Service Electric 10-23-87 SN-412 and Gas 87-83 Morrison Knudsen 09-27-87 N/A 87-84 Combustion Engineering 10-02-87 Addressed

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87-86 Carolina Power & Light 09-22-87 N/A

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87-88 Indiana and Michigan 08-26-87 Addressed Electric Company 87-90 Detroit Edison 07-06-87 SN-412

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88-01 Gulf States Utilities 11-16-87 NE-18764 88-02 IMO Delaval 12-03-87 NE-18764 88-06 Nebraska Public Power 03-07-88 N/A 88-08 IMO Delaval 04-29-88 NE-19269 88-16 Automatic Switch Co.

10-18-88 N/A 88-17 IMO Delaval 10-05-88 Unit 1 Corrected 88-19 Limitorque 11-3-88 Addressed 89-02 Automatic Switch Co.

1-30-89 N/A 89-04 Control Component, Inc.

04-04-89 N/A 89-05 ABB ASEA Brown Boveri 04-17-89 SN-478 v

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A 20' percent sample of the applicant's evaluations for the following 10: CFR Part 21 reports without NRC Region IV tracking

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numbers was reviewed to assess'the adequacy of the applicant's l

response.

Based on the results of the review, the NRC inspector

' determined that the applicant's evaluations and actions were

appropriate.

Letter-Related CPSES Documents Company.

Date or Remark g

Westinghouse Electric 04/26/85 SN-233 Foxboro Company 08/22/86 SN-203

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Westinghouse Electric 10/23/86 Addressed Weidmuller 03/03/87 SN-224 Gamma-Metrics 07/13/87 Tel-Con GM-TU 08/26/87 Fisher Controls 11/18/87 CR-C87-4858 IMO Delaval 12/02/87 DR-C87-5232 Westinghouse Electric 12/03/87~

DR-C87-5371 Westinghouse Electric 12/30/87 DR-C87-5370

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Westinghouse Electric 01/29/88 DR-C88-091 Westinghouse Electric 02/18/88 SER-45-83 General Electric 02/19/88 DR-C88-1169 Westinghouse Electric 02/26f88 NCR-88-3017

Hilti 03/18/88 Plant Evaluation IN-88-25 Westinghouse Electric 05/19/88 Reactor Engineering Evaluation Westinghouse Electric 07/29/88 Plant Evaluation i

Fisher Control 10/14/88 Plant Review of IN-88-73 i

Rockwell International 01/23/89 SN-437 Foxboro 05/19/89 TSL-89195

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Fisher Control-06/08/89 TSL-89184 Fisher Control 06/15/89 TSL-89184 Westinghouse Electric 06/15/89 N/A Ops /Results Foxboro Company 08/03/89 TSL-89254 During the above review of TU Electric evaluations, the NRC inspector noted that the applicant's evaluations of two 10 CFR Part 21 reports were ongoing.

The applicant had taken adequate actions to identify and control the identified components; h

however, the applicant's final evaluation had not been performed.

The NRC inspector reviewed those actions and determined them to be adequate and not impact fuel load for Unit 1.

The items are identified by NRC Region IV tracking numbers 86-01 and 87-09 and will be an NRC open item 50-445/9002-0-02 pending satisfactory NRC review of the final evaluations and corrective actions if require.:

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In. summary, during this inspection and a previous NRC inspection-(see NRC Inspection Report 50-445/89-86; 50-446/89-86) the NRC

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inspectors verified that the applicant has evaluated-10 CFR Part 21 reports and that the evaluations and corrective actions appeared appropriate.

Specifically reviewed were those 10 CFR-Part 21 reports received from the NRC and identified by NRC Region IV tracking Nos. 86-01'through 89-10'as well as other applicable Part 21 reports rect'.ved from~ vendors.

No violations o

or deviations were noted.

One open item was-identified above for subsequent review of the applicant's;tesolution of two of the Part 21 reports.

Since receipt'and review of 10 CFR Part;21 reports is sn ongoing-applicant activity, additional NRC inspection of that activity will be performed.

No violations or deviations were-identified.'

' 11.

Safety Evaluation Report Follow-up (92719)

Section 9.5.4 of Supplemental Safety Evaluation Report (SSER) 22 indicated that the fuel oil suction lines connected to the bottoms of the EDG fuel oil day tanks would be modified.

The

modification would allow for removal of water and sediment from the; lines.

The NRC inspector examined the EDG day tank fuel oil suction line arrangement and'found that capped sample points were-isolated by valves 1-DO-0410 and 1-DO-0411.

Procedure OPT-214A,

" Diesel Generator Operability Test," was reviewed-to verify that it included.a requirement to drain the water from the EDG day tanks every 31 days as required by Technical Specification 4.8.1.1.2.b.-

Step 9.44 included this requirement.

During this inspection period, the NRC inspector observed-the testing of the Train B EDG including observation of the sampling of fuel oil from the associated day tank.

No water or sediment was observed in'the sample.

The NRC inspector also reviewed-Procedure SOP-609A, " Diesel Generator System," to verify that it required draining accumulated water from the EDG fuel oil day tank following diesel runs of one hour or more.

Step 5.3.1.R contained this requirement.

No diolations or deviations were identified.

12.

Engineered Safety Feature System Walkdown (71710)

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The inspectors performed a walkdown of both trains of the Unit 1 residual heat removal' system to confirm that the applicant's system lineup procedure matches plant drawings and the as-built configuration and to identify equipment conditions and items that might degrade system performance.

Attachments 1, 2, and 3 of Procedure SOP-102A, " Residual Heat Removal System," Revision 5, were used for the walkdown.

Drawings used included:

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M1-0260 Revision CP-14

M1-0263 Revision CP-12 i

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M1-0263B Revision Cp-5

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M1-0261 Revision CP-12 The following minor discrepancies were noted:

The seal. tag for 1RH-0023, the 1-8702A bypass,

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identified it as the 1-8702B bypass.

The_ applicant corrected this tag.:

i Approximately nine valves were capped,-although neither:

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the drawing nor the procedure showed them as capped.

t The applicant stated that caps will=be used on vents, drains, and test connections when the. pipe has been-threaded to accept caps.

The' procedure lineup will specify caps.only for those locations shown as! capped-on'the drawings.

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There was a minot discrepancy between the valve

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description in the procedure lineup and on the valve

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_4 tag for 1RH-0007.

The applicant stated that the~new labeling program currently being implemented ~would j

correct this.

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1SI-0109 and 1SI-0110 were each locked in position with'

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a_ padlock, although this was not prescribed by the procedure lineup or by the drawing.

The applicant removed the locks from these valves.

. Valves 1SI-8967A and 1SI-896713 were not in the RHR

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-procedureLline.up, although-they are located within the j

O applicant added these valves to.the RHR lineup in. The

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system scope covered by the RHR procedure' lineup.

Revision 6 of the procedure and stated that they would be removed from the safety injection system lineup 11n-SOP-201A.

i Several vent and drain valves had minor leaks..

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Applicant personnel wrote work requests to initiate corrective action for these.

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The RHR pump discharge sample points were not labelled.-

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Appropriate. labels were subsequently attached.

t None of the identified discrepancies had the potential to adversely impact the operability of the RHR system.

No l_

violations or deviations were identified.

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Initial Fuel Loadino Procedure Review (72500)

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.This-inspection effort was continued from previous inspection periods.

The remaining items have been addressed and the necessary procedure revisions have_been issued.

A prerequisite ~

to: verify that the containment purge system is operable prior to

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fuel loading was not necessary since this system is required by the Technical Specifications to be operable only in Modes 1 through 4.

Containment isolation capability is verified on Form ISU-001A-1.

Administrative controls over the number of.

l personnel and type of activities to be permitted in the control

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room, fuel building, and containment during fuel loading

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operations include clean area boundaries in the containment and fuel buildings.

Controls over these boundaries include material'

and personnel entrance controls and-logging.- The integrated work and test schedule will be used to ensure that inappropriate work activities are not permitted in these areas.

The technical support manager _has been designated as the person responsible for coordinating fuel: loading and Westinghouse will provide technical assistance as required.

RFO-302 has been revised to include the weight of a fuel assembly with a silver-indium-cadmium control rod assembly.

RFO-102A and Form ISU-001A-1 have been revised to move the Technical Specification compliance checklist from RFO-102A to

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Form ISU-001A-1.

A rehearsal session involving the insertion of a dummy primary sourew rod into a dummy fuel-assembly was observed by the inspector.

The personnel involved appeared proficient in the use i

of the equipment and procedures.

The reactor engineer conducted a briefing session for the crew prior to the. start of the evolution.

This, too, was observed by the inspector.

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During.the briefing and evolution, the inspector noted two strengths.

These involve the quality of the briefing and the performances of the fuel handling supervisor and the personnel from all departments involved (radiation protection, operations, maintenance, and security).

The briefing covered the pertinent steps of the evolution and was

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followed by an opportunity for the crew to ask questions.

The resulting dialogue proved beneficial to the successful completion g

of the evolution.

During the evolution, personnel from all departments involved exhibited an enthusiasm and willingness to work together.

Particularly noteworthy was the fuel handling supervisor's ability to coordinate his personnel and work calmly under pressure.

A training session involving movement of a dummy fuel assembly to core locations and to containment storage locations was observed by the inspector.

The personnel involved appeared to be

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containment.

The inspector made suggestions to.the fuel handling supervisor concerning improvements in the control of tools and equipment-in the clean area.-

The following procedures were reviewed during this inspection i

period:

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Procedure Revision Title

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RFO-101:

2-1 Refueling Organization RFO-106

Development and Implementation of the

Reload Fuel Shuffle Sequence Plan RFO-207 1-1 Surveillance of Fuel Assemblies and l

Insert Components L

RFO-302-4-2 Handling of Fuel Assemblies l

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ISU-001A

Initial Fuel Load Sequence

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No violations or deviations were identified.

14.

Followup on'NRC Bulletin 89-03'(92703)

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'NRC Bulletin 89-03, " Potential Loss of Required Shutdown Margin

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During Refueling Operations," was issued on November 21, 1989.

It pointed out the potential for loss of shutdown margin and listed three actions to prevent problems in this' regard..

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TU Electric responded to the bulletin with a letter (TXX-89873)

dated January 5, 1990.

This letter committed to complete'the required actions for Unit 1 initial fuel load prior to Unit 1 fuel load.

-For Action 1, the NRC inspector found that the fuel vendor.had-provided guidelines to TU Electric in a letter (89TB*-G-0061)

dated December 20, 1989.

TU' Electric evaluated' intermediate fuel assembly and control rod. configurations resulting from the planned fuel load sequence against these guidelines to assure l

compliance.

Procedure RFO-106, " Development and: Implementation of the Reload Fuel Shuffle Sequence Plan," was revised.to

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~ incorporate.the vendor guidelines as Attachment 8.B.

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Precaution 6.2.1 requires that the Attachment 8.B guidelines BEL followed when planning. fuel movements to assure that core reactivity is maintained within the core design analysis.

g For Action 2, the NRC inspector verified that RFO-106 requires verification that proposed fuel shuffle sequence changes meet the

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reactivity control guidelines of Attachment 8.B.

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i For-Action 3, the NRC inspector reviewed the attendance, lesson plani objectives,' handouts, examination, and test grades for.

Ltraining of fuel handling supervisors and reactor engineers in criticality control'during refueling-This~ training, completed

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on January 18, 1C90, appeared to meet the requirements of the bulletin.

The act' ions required by Bulletin 89-03 related to Unit 1 initial

' fuel loading ~have been completed.

No violati'ons or deviations-were identified, m-15..

open Items

open items are matters which have been discussed with the applicant, which~will be reviewed further by the. inspector, and

w h i c h i n v o l v e s o m e a c t i o n o n t h e p a r t o f t h e N R C (nr r. p p l i c a n t ' o r both.

Two open items identified during the inspection are discussed in paragraphs 2.h and 10.

16.

Exit Meeting =(30703)

Exit meetings'were: conducted on January 26 and February 6,.1990, with the applicant's representatives identified in paragraph 1 of this report.

No written material was provided to the applicant by the inspectors during this reporting period.

The applicant-did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.

During

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these meetings, the NRC inspectors summarized-the scope and

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findings of the inspections.

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