IR 05000445/1989024
| ML20247G659 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/18/1989 |
| From: | Bitter S, Burris S, Joel Wiebe Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247G577 | List: |
| References | |
| TASK-2.F.1, TASK-2.F.2, TASK-2.G.1, TASK-2.K.1, TASK-2.K.3.05, TASK-2.K.3.10, TASK-2.K.3.12, TASK-2.K.3.25, TASK-3.D.1.1, TASK-3.D.3.3, TASK-TM 50-445-89-24, 50-446-89-24, IEB-80-06, IEB-80-6, IEB-85-003, IEB-85-3, NUDOCS 8905310038 | |
| Download: ML20247G659 (25) | |
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APPENDIX B U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION NRC Inspection Report:
50-445/89-24 Permits: CPPR-126 50-446/89-24 CPPR-127 Dockets: 50-445 Category: A2 50-446 Construction Permit 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:
April 5 through May 2, 1989 Inspector:
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/,8 S.
D. Bitter, Resitent Inspector,
/ bate Operations Mit!4 'l Inspector:
S.
F. Burris, Senior Resident Inspector,
/ Date l
l Operations Si/
l Reviewed by:
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~. Wiebe, Senior Project Inspector
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8905310038 890518 PDR ADOCK 05000445 Q
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-, Inspection Summary Inspection Codducted: April 5 through May 2, 1989 (Report 50-445/89-24; 50-446/89-24)
Areas Inspected:
Unannounced-resident safety inspection of
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applicant action on previous inspection findings, follow-up on
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. violation / deviation, action on 10 CFR'Part 50.55(e) deficiencies identified by the applicant, NRC bulletins, preoperational retest program activities,.TMI-action items (Safety Issue Management System items II.F.1.2.1, closed; II.F.1.2.B, open; II.F.1.3, open; II.F.1.4., open; II.F.1.5, open; II.F.1.6, open; II.F.2.1.A, open; II.F.2.3, open; II.G.1.1, closed; II.K.1.3, open; II.K.3.1.B, closed;'II.K.3.5.B, open; II.K.3.9, open; II.K.3.10, closed; II.K.3.12.B, closed; II.K.3.25.B.2, closed; III.D.1.1.1, open;
.III.D.3.3.1, open;.III.D.3.3.2, open.), and plant tours.
Results:
Within the areas inspected, significant weaknesses were noted in the operations and preoperational retest programs.
These weaknesses are discussed in paragraph 6 of this report.
One violation was identified which is discussed in detail in paragraph 6 of'this report.
One unresolved item is discussed in paragraph 8 and two open items are discussed in paragraph 8 of this report.
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DETAILS 1.
Persons Contacted
- G.
K. Afflerbach, ASM Startup, TU Electric
- J. W. Beck, Vice President, Nuclear Engineering, TU Electric
- M.
R. Blevins, Manager, Technical Support, TU Electric
- H.
D. Bruner, Senior Vice President, TU Electric
- R.
J. Daly, Manager, Startup, TU Electric
- J. W. Donahue, Operations Manager, TU Electric
- D.
E. Deviney, Deputy Director, Quality Assurance (QA),
TU Electric
- W.
G. Guldemond, Manager of Site Licensing, TU Electric
- T.
L. Heatherly, Licensing Compliance Engineer, TU Electric
- J.
J. Kelley, Manager, Plant Operations, TU Electric
- D.
M. McAfee, Manager, QA, TU Electric
- E.
F. Ottney, Program Manager, CASE
- P.
W.
Pellette, Operations, TU Electric
- D.
M. Reynerson, Director of Construction, TU Electric
- A.
B.
Scott, Vice President, Nuclear Operations, TU Electric
- J.
C.
Smith, Plant Operations Staff, TU Electric
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- M.
A. Thero, QTC Intern
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- O.
L. Thero, QTC Consultant to CASE
- R.
D. Walker, Manager of Nuclear Licensing, TU Electric
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The NRC inspector also interviewed other applicant employees during this inspection period.
- Denotes personnel present at the May 2, 1989, exit interview.
2.
Arplicant Action on Previous Inspection Findings (92701)
l a.
(Closed) Open Item (445/8802-0-01):
Containment hydrogen analyzer calibration gas leakage.
This item stemmed from
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a 10 CFR Part 21 notification concerning a calibration gas leakage in the Salem Unit 1 and 2 containment hydrogen analyzer systems.
The notification indicated that CPSES is one of three other sites using the same system.
The applicant has determined that the gas leakage problem is the result of a generic design deficiency in systems manufactured by Exo-sensor, Inc. (now Whittaker Corpora-tion).
As corrective action, the applicant documented the problem in Deficiency Report (DR) C-88-01137.
The DR was dispositioned by initiating Design Change Authorization (DCA) 75307.
This DCA calls for implementing vendor-recommended modifications.
These include installing, in a new configuration, a new solenoid valve and a new regulator in each of the four analyzer units.
The parts for the modification have been received and the modification is scheduled to be completed prior to fuel
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load. LThe inspectors have reviewed these actions and are
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satisfied with the applicant's resolution of this issue.
Therefore, this-item is closed.=
b.
(closed) open Item (445/8844-0-03): " Quality Surveillance Group Documentation of Conditions Adverse to Quality."
-This was originally identified in NRC Inspection' Reports
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50-445/88-39; 50-446/88-33 and;50-446/88-44; 50-446/88-40 which: identified that not all of the concerns of.a quality surveillance report had been officially documented'within an approved applicant program.
Review of this item and'
the applicant's subsequent resolution to the inspectors'
concerns are as follows:
Surveillance program Procedures STA-402 and QAI-017-
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have been retired and replaced by Procedure NQA-3.23,
" Surveillance: Program."
The applicant has added additional detailed directives to this procedure to
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provide the.necessary guidance for documenting unsatisfactory conditions identified during a surveillance.
Section 6.4 of Procedure NQA-3.23, " Documenting
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Unsatisfactory Conditions" states, in part,
" Identified unsatisfactory conditions involving quality related items / activities shall be documented on the surveillance report as deficiencies unless one of the following actions has been taken:
"a.
The: Unsatisfactory condition has been reported Lpreviously via an appropriate corrective action document.
"b.
The unsatisfactory condition had no apparent generic implications and was corrected as allowed by applicable governing precedures prior to issuance of a surveillance report.
"c.
The unsatisfactory condition was documented for resolution in accordance with methods estab-lished in applicable governing procedures."
The inspectors discussed item b (above) with QA management to' determine how and subsequently who would have the final authority and responsibility for determination of whether or not an identified item had generic implications.
The inspectors were informed that the final determination
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as to whether or not an item had "no apparent generic (
implications" would be made at least at the supervisor level (review by independent person).
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The inspectors consider the applicant's actions adequate to resolve this issue.
This item is considered closed.
c.
(Closed) Unresolved Item (445/8879-U-01):
Containment ventilation piping classification.
This issue arose when an NRC inspector noticed that, on Drawing M1-0301, Revision CP-10, the applicant had classified the ventilation piping between the inside containment isolation valve and the debris screen as safety Class 5 (nonseismic).
On the other hand, the debris screen itself was' classified as seismic Category I.
The NRC inspector.
then concluded that unless the ventilation piping was also classified as seismic Category I, the debris screen could be rendered useless if the ventilation piping were to fail.
The applicant has. reviewed this concern and determined that the drawing used by the NRC inspector contains a typographical error.
The ventilation piping of concern is actually qualified as seismic Category I.
Ac corrective action, the applicant has issued a design change authorization (DCA) that corrects the typographical error.
Furthermore, the NRC inspector hat reviewed the seismic calculation for the ventilation piping in question.
The NRC inspector is satisfied that the ventilation piping is actually qualified as seismic Category I and has no further questions.
Therefore, this item is closed.
3.
Follow-up on Violations / Deviations (92702)
(Closed) Violation (445/8844-V-01):
" Identification and Resolution of Deficient Conditions," was issued based on three examples of the applicant failing to properly document conditions adverse to quality.
The inspectors reviewed the applicant's response TXX-88649 dated September 6, 1988, which outlined the following corrective action:
Need for a clear, conservative threshold for reporting
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deficiencies.
Directive to all management personnel outlining the need
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for adequate timely resolution for identified problems.
Review of the temporary modification log and appropriate
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procedures are to be revised requiring verification walkdowns at least every six months.
Training on use of the cross-reference in STA-606 which
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provides guidance for documenting potential problems.
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The-inspectors reviewed the above referenced corrective actions
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and discussed these responses in detail with the applicant's senior management.
This effort found that the applicant's current threshold for reporting deficiencies to be conservative in nature and thia philosophy was reflected in all subordinate personnel
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interviewed.- The onsite inspection staff currently reviews all L
site generated problem reports and to date have not found similar problems with documenting nonconforming conditions.
Based on this review, the inspectors consider this item closed.
4.
Action on 10 CFR Part-50.55(e) Deficiencies Identified by the Applicant (92700)
a.
(Closed) Construction Deficiency (SDAR CP-86-64):
" Coatings for Diesel Fuel Oil Tanks."
Originally this item'was identified while cleaning the Unit 2 diesel generator fuel oil. tanks (DGFOT) when the applicant noticed a corrosion area approximately two feet in diameter.
The original architect and engineering' firm (Gibbs and Hill) did'not consider the DGFOT interior coating to be sEfety-related.
However, when the coating became governed by Specification AS-31, " Protective Coatings," it was regarded as safety-related.
Later when the specification for containment coating was reclassified as not safety related and containment painting became nonsafety-related, the DGFOT coating was similarly reclassified.
DGFOT coating removal was found to be advantageous to ensure the reliability of the diesel generator system.
The applicant removed the applied coating and subsequently performed extensive ultrasonic examination of the DGFOT to ensure the minimum wall thicknesses were not exceeded.
Review of the documentation found that currently the
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diesel generator fuel oil tanks are considered by both TU Electric'and the NRC to be acceptable and operable; therefore, tnis item is considered closed.
b.
(Closed) Construction Deficiency (SDAR CP-87-37): " Failure I.
of Westinghouse Valves to Completely Close Under Full Pressure Conditions."
This item was brought to the applicant's attention by Westinghouse in an October 1980 (WPT-4111) letter which stated, in part:
"Preoperational tests disclosed that Westinghouse Electro-Mechanical Division three inch gate valves, Model 3GM88, 1500 lb. class, failed to completely close under preoperational test conditions (i.e., approximately 2700 psid as fluw approaches zero).
These conditions were less
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severe than the equipment specification design conditions (i.e.,
2750 psid as flow approaches zero).
At that time, Westinghouse also reported that the later redesigned version of this valve, the Model 3GM99, may also be subject to the same problem, althcugh no testing was performed on this model."
The applicant generated a Deficiency Report (DR) C88-2289 which identified that all siz 3 of Westinghouse motor-operated Pete valves (sizes 3, 4, 6, 8, 10, 12, 14, 16, and 18 inch) might not completely close under conditions that are less severe than the equipment specification design conditions.
Subsequent correspondence from Westinghouse (WPT=4286)'
stated, in part,
. the changes required to guarantee
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full closure can vary..They could involve any combination of the following items:
Torque switch ad3ustment;
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Increasing the operator gear ratio to guarantee
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adequate thrust capacity at 80 percent voltage; Rewiring the operator for limit closing control;
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Changing the operator torque switch spring pack;
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Changing a larger foot-pound rated motor; or
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Changing to a larger size Limitorque operator.
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The applicant has implemented these recommended changes
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l which in most cases was to adjust the torque switch setting and/or motor upgrade.
The inspectors have reviewed the work documents and verified that these valves would be retested in accordance with the current preoperational retest program requirements.
In cddition, the applicant is taking actions in response to NRC Bulletin 85-03, " Motor Operated Valve Failures."
Based on this review, this item is considered closed.
c.
(Clciad) Construction Deficiency (SDAR CP-87-91):
"Startup Transformer Overload."
During review of the Design Basis Documents (DBD-EE-038) for the startup transformers, the differences between licensing commit-ments and the transformers specification (2323-ES-2A&B)
indicate that the loading condition for the transformers may exceed their overload capacity.
Comanche Peak FSAR, Section 8.3.1.1.1 states that the transformers are rated for 49% overload which is an
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overload rating of 63,800 KVA; however, during a postulated overload condition, the calculated KVA load is 72,600 KVA.
The reason for this design difference was that the original startup and auxiliary transformer Load Study' Calculation (III-7, Revision 3) and Bus Feeder Calculation (VII-2, Revision 3) did not take into account the impact to the transformer bushings or the plant voltage profile.
This problem was determined by the applicant to be limited to the startup transformers.
The inspectors reviewed various documents that incorporated the approved changes to Unit 1 and common switchgear arrangement for an additional startup.
transformer and its physical incorporation into the original designed AC distribution system.
These documents included DCAs, DBDs, instructions, procedures, work orders, transformer data sheets, and test procedures.
Based-on this review, the inspectors feel confident that the applicant has corrected the original concern based on the detailed multidisciplinary review performed by the Joint Test Group and final acceptance of associated preoperational tests (lCP-AT-03-01 and 1CP-AT-03-2).
Based of this review, the inspectors cor. sider this item closed.
d.
(Closed) Construction Deficiency (SDAR CP-87-130):
" Service Water System Water Hammer."
Design review of the station service water system (SSWS) found a potential for an adverse effect due to water hammer problems.
Specifically, the design validation process of the SSWS revealed piping high points at the. inlet and outlet of the emergency diesel generator jacket water coolers at which water column separation can occur subsequent to a service water pump trip.
When the column rejoins, it could result in a water hammer and pipe and/or pipe support damage and thereby cause a loss of the emergency diesel generator.
The applicant reviewed this issue and determined that the corrective action to preclude the loss of the emergency diesel generator power would be to install vacuum breakers.
DCAs 55662, Revision 4, and 62476, Revision 1, implemented Stone and Websters' calculation 15454-NP(B)-F34 for the appropriate sizing requirements of the vacuum breakers.
The applicant obtained (Field Requisition FR-6R-345728) the vacuum breakers and installed them in accordance with the approved work documents (MPP-88-0512-0400, MPP-88-0514-0400, MPP-88-0522-044, and SWA 47549).
Subsequent system operability verification was performed in accordance with SSW water hammer test, REI-701, Revis, ion 0, and preoperational test completion, which
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provided reliable assurance that this system was not susceptible to water hammer problems.
Based on this review,.this item is considered closed.
e.
(Closed) Construction Deficiency (SDAR CP-88-18):
" Post Accident Monitoring Instrumentation."
This issue was identified by TU Electric during'a design review of the reactor coolant system (RCS).- RCS pressure and temperature is defined in DBD-EE-004 and Regulatory Guide 1.97, Revision 2, as Type A and Category I variables.
The DBD stipulates these variables must be Class 1E indications (T hot and T cold) for each reactor coolent loop.
As designed, the monitoring systems for RCS temperature and pressure was determined to be deficient as follows: there were only two qualified channel recorders for;T hot and'two qualified channel recorders for T cold; in addition,~there are only two monitoring channels for-
RCS pressure.
Failure of:only one channel of pressure
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variable cn: loss of one of either the T hot or T cold i
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recorders may cause the operator to be mislead by his indications asEto the true system status.
The applicant has reviewed this issue and determined that corrective action for this issue is as.follows: install a third qualified monitoring channel for RCS pressure replacement of the existing T hot and T cold recorders with Class 1E indicators.
The applicant generated documents to change the hardware applications to meet the original DBD and applicable Regulatory Guide.
The inspectors reviewed the original DR C87-5368, Travelers CP1-ECPRCB-5-8, 1-PT-0437, and DCAs 62908, Revision 1; 75681, Revision 1; 77123, Revision 1; 77743, Revision 3;
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and 77817, Revision 1; and finally Nonconformance Reports
(NCRs) 89-01521, Revision 1, and 89-01670, Revision 2, to verify that these changes were accomplished as described and were properly controlled during work activities.
Of those activities reviewed, the inspectors did not find any problems or concerns.
Based on this review, the inspectors consider this item closed.
5.
Follow-up on NRC Bulletins (92700)
(Closed) IEB 80-06, " Containment Ventilation Isolation Signal Blocked by Reset":
The original concern dealt with the use of
i the SI reset push buttons which could result in certain ventilation dampers changing position from their safety or
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emergency mode to their normal mode of operation or condition.
i The staff review, as identified in Comanche Peak Steam Electric Station Safety Evaluation Report and Supplement No. 2, Section 7.3.2.1, found the applicant review and subsequent modification to be acceptable pending final acceptance of the
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preoperational test which will be verified.by the onsite NRC staff.
This item is considered closed.
6.
Preoperational Retest Program Activities (70300, 70311, 70312, 70438, 71302 NRC inspections of the applicant's preoperational retest and operational preparedness phase activities were performed through direct observation, personnel interviews, and review.of preoperational test activities by verifying that:
All management and administrative controls and procedures,
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including QA requirements which were required for the necessary operation, had been implemented, followed, and documented.
Applicant management and supervision were aware of the
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current status at all times and were maintaining control over all areas of preope_ational testing.
Systems and components important to the safety of the
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plant were being fully tested to demonstrate their-operability and design requirements.
The NRC inspectors accomplished these goals by reviewing available test procedures, witnessing selected ongoing test activities and reviewing completed test results.
The inspectors used the following criteria to perforin the pretest review to ensure that:
Prerequisite conditions were established, adequately
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defined, and easily understood.
Test equipment used specified the appropriate custody
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control and required calibration data.
Procedure format was clearly written and appeared to be
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able to be easily followed.
Test objectives met the referenced Regulatory Guide and
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FSAR Section 14 commitments.
Acceptance criteria were identified and clearly defined.
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Administrative content, format, and requirements were
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incorporated in the final approved procedure.
Test witnessing of the identified systems was accomplished to ensure that all testing was performed in accordance with approved procedures and to verify the adequacy of test program records including preliminary evaluation of test results.
The NRC inspectors accomplished these purposes by ensuring that:
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The latest revision of the test procedure was in use by
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test personnel.
Testing was performed in accordance with an approved
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procedure.
Criteria for interruption of testing and continuation of
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testing was adhered to during all witnessed portions of the test.
All deficiencies were documented in accordance with
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program requirements.
All temporary modifications, such as jumpers, strainers,
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spool pieces,.or blank flanges were installed and tracked per established administrative controls.
Test equipment required by the procedure was calibrated
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and in service, if applicable.
All crew manning requirements were met.
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All test prerequisites were met.
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Proper plant systems were in service.
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The NRC inspectors reviewed the following test:
Remote Shutdown Capabilities Test:
The NRC inspectors
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obtained copies of the applicant control Procedure ISU-223A, Revision 1,
" Remote Shutdown Capabilities Test,"
and Operations Procedure ABN-905A, Revision 1, " Loss of Control Room Habitability," to ensure that the intent of Regulatory Guide 1.68.2, " Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water Cooled Nuclear Power Plants," and the current Chapter 14, FSAR commitments would be met when this test was performed.
Regulatory Guide 1.68.2 and FSAR commitments identify certain specific requireruents which must be met in order for the applicant to assure the Commission that the plant can be safely shutdown following a loss of the control room.
The purpose of this demonstration is threefold.
(1)
To demonstrate that the design of the plant is adequate to meet the requirements of General Design Criterion 19.
(2)
To demonstrate that the procedure or procedures to be used in performing the shutdown from outside the control room are sufficiently clear and comprehensive and that the operating personnel are familiar with their application.
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(3)
To demonstrate that the number-of personnel available to conduct the shutdown operation is sufficient'to
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. perform the many actions required by the procedure in-a timely, coordinated manner.
To satisfy.the above purposes, the.following objectives
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must be met when the1 applicant performs this' test.
(1)
Verification that'the nuclear power plant can be:
safely; shutdown-from outside the control room.
(2)
Verification that the nuclear power ~ plant can be maintained in a hot shutdown condition from outside control room.
(3)
Verification that the nuclear power plant has.the potential for being safely cooled from hot to cold shutdown conditions from outside the control room.
Based on discussions between the NbJ inspector and
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TU Electric management, the applicant plans on performing this test in two phases: (1) Precore Load Phase (PCL 1)
and (2) Post Core Load Phase (PCL 2).
PCL.1 phase will be performed at the end of Hot Functional testing to:
(1)
Verify that the plant can.be safely shutdown from outside the control room (Mode 3 to Mode 4 conditions).
(2)
Verify that the plant has the potential for being safely cooled from hot to cold shutdown conditions from outside the control room (Mode 4 to Mode 5 conditions).
(3)
Verify that all of the objectives of Regulatory Guide and FSAR commitments have been met.
PCL 2 phase is currently scheduled to be performed after
. fuel load..This portion of the test will be done to:
(1)
Verify that the plant can be safely shutdown from outside the control room.
(3)
Verify that the plant can be safely shutdown by tripping the reactor from-between 10% - 25% reactor power.
(4)
Verify that the plant can be maintained at Hot L
standby (Mode 3) and the potential exists to reduce the plant to Hot shutdown.
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The NRC inspectors informed the applicant that they would
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review and witness the major aspects of this proposed test l
to ensure that the previously identified objectives would be mer.
As identified during both this and the previous months
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inspection periods, certain activities failed to meet 10 CFR 50, Appendix B requirements.
The following four l
discrepancies are examples of failure to follow procedures.
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The first three items were identified in NRC Inspection Report 50-445/89-17; 50-446/89-17 as unresolved items.
The fourth item was identified during this inspection period.
a.
445/8917-U-03, " Failure to Obtain Shift Supervisor Signature."
The applicant identified this discrepancy during the performance of preoperational test Procedure 1CP-PT-02-01 SFT, Revision 1, "118 Volt Class lE AC Inverters."
Several days after initiating this test, the shift test engineer (STE) who was performing the test discovered that he had failed to obtain the shift supervisor's permission to perform the test.
Furthermore, he had not obtained the shift supervisor's approval to implement three temporary procedure changes (TPC-1, TPC-2, and TPC-3).
b.
445/8917-U-08, " Unauthorized Work During Preoperational Testing."
The applicant identified this discrepancy during the performance of preoperational Test Procedure 1CP-PT-37-01 SFT, Revision 0,
" Auxiliary
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Feedwater System (Motor Driven Pumps)."
While performing this test, the shift test engineer (STE)
discovered that he could not perform step 7.1.2.18 because the required test equipment (a comparator and associated equipment) was no longer connected to the flow transmitter.
The equipment had been connected earlier, at step 7.1.2.4.
No step between steps 7.1.2.4 and 7.1.2.18 called for its removal.
The STE documented this discrepancy and investigated the ca.rcumstances.
He determined that the Startup technicians did not want to leave the test equipment connected while it was not in use.
Therefore, the equipment was disconnected and removed.
Both actions took place without the STE's knowledge or approval.
c.
445/8917-U-09, "Preoperational Testing in Wrong Cabinet."
This discrepancy was identified by the applicant during the performance of preoperational Test Procedure 1CP-PT-64-03 SFT, Revision 0, " Turbine Runback and Reactor Control."
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When the STE directed that step 7.3.7 be performed, a bistable switch was manipulated in the wrong process control cabinet.
Instead of manipulating the switch in process control Cabinet 1, as the step called for, the same functional switch was manipulated in process control Csbinet 2.
The STE discovered this discrepancy immediately because the wreng alarm actuated at the main control board.
The STE investigated and determined that his assistant had performed the step in the wrong cabinet.
d.
During this inspection period, the applicant reported that an equipment operator was assigned to perform steps 5.5.3.1 and 5.5.3.2 of CDSES Standard Operating Procedure (SOP) 304A, Revisi c 5, so that the turbine-driven auxiliary feedwater pump could be realigned to the test header.
The operator failed to perform these two steps in the specified sequence.
Consequently, because of this procedural error, coupled with the failure of several check valves in the auxiliary feedwater system, steam generator water flowed backward through the auxiliary feedwater piping, into the condensate storage
tank.
This resulted in discolored and blistered paint on
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the affected piping.
The multiple check valve failure is
currently the subject of applicant investigation as well as the subject of an augmented inspection team (AIT) by the NRC.
These four examples of a failure-to-follow-procedure are, collectively, a violation of 10 CFR Fart 50, Appendix B, Criterion V, which requires, in part, that activities affecting quality shall b2 prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
CPSES Test Department Administration (TDA) Manual Procedure TDA-101, Revision 0, Section 5.11 states that test engineers shall provide direction to support personal and others during performance of preoperational tests.
CPSES TDA Manual Procedure TDA-303, Revision 0, Section 6.2.7, requires that preoperational testing be performed in accordance with approved test procedures and instructions.
CPSES TDA Manual Procedure TDA-304, Revision 0, Section 6.1, states that all changes to preoperational test procedures must be approved by the shift supervisor.
Furthermore, Section 6.2 requires the test engineer to obtain the shift supervisor's approval to initiate preoperational test _ _ _ _ _ _ _ _ _ _ -
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Procedure ODA-407, Revision 1, Section 6.1, requires that plant systems and subsystems be operated in accordance with written approved procedures during normal, abnormal, and emergency conditions except as allowed by Procedure ODA-407.
Section 6.1 of ODA-407 also requires that any individual who cannot follow an operating procedure as written, shall place the system or component into a safe or stable condition and notify the shift supervisor immediately.
Finally, Section 6.1 of ODA-407 states that t
if the desired or anticipated results in a step are not achieved, the individual should not proceed.
The four discrepancies described in items a, b, c, and d above are contrary to the requirements of procedures TDA-101, TDA-303, TDA-304, and ODA-407, in that they collectively constitute a failure-to-follow-procedure.
The unresolved items 445/8917-U-03; 445/8917-U-08, and 445-8917-U-09 are closed and comprise three of four.
portions of the violation (445/8924-V-01):
" Failure to Follow Procedures During Preoperational Testing."
7.
TMI Action Items (SIMS)* (25565)
- The Safety Issue Management System t<hb) tracking number is the same as the TMI Action Item number.
l a.
(Closed) TMI Action Item II.F.1.2.1:
Interim Iodine /
l Particulate Sampling.
Since the FSAR in Section II.F.1 commits to completely implementing the Iodine / Particulate sampling prior to fuel load, the interim actions are not necessary.
The inspection requirements of Temporary Instruction 2515/65 are considered complete and this item is closed for Unit 1 and Unit 2.
b.
(Open) TMI Action Item II.F.1.2.B:
Long Term Inclementation of Iodine / Particulate Sampling.
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NUhrG-0797, Chapter 22,Section II.F.I, Attachment 2,
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documented that the staff considered the applicant's response to this item incomplete.
Supplement 3 to i
NUREG-0797 in Chapter 22,Section II.F.I, Attachment 2, documented the staff's evaluation of additional information presented in FSAR Amendments 27 and 31.
The staff concluded that the iodine and particulate sampling
devient met the provisions of NUREG-0737, " Clarification of TM1 'ction Plan Requirements" and are acceptable.
A The NRC inspector needs additional information to be able to close this item as follows:
(1)
NUREG-0797, Chapter 22,Section II.F.1, Attachment 2, states that in Amendment 14 to the FSAR, the
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16-applicant committed to include the iodine / particulate sampler in the operating procedures and training
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program prior to fuel load.
The NRC inspector needs to review documentation that this commitment has been met.
(2)
The NRC inspector notes that the iodine / particulate sampler is not included in Table 3.3-6 of the Final Draft Version of the CPSES Technical Specification.
The_NRC inspector needs to review the applicant's program for maintaining operability of this monitor including requirements for calibration and action taken when the sampler is inoperable.
This item remains open pending NRC review of the above identified information.
c.
(Open) TMI Action Item II.F.1.3:
Containment High Range Monitor.
This item involved the provision of two radiation monitoring systems in containment.
The FSAR,Section II.F.1, stated that the redundant monitors will be located in the containment building at elevation 905'-9",
g and be located at least 90 apart, and will not be located adjacent to process piping.
The NRC inspector verified that the applicant's commitment with regard to location was met.
NUREG-0797, Chapter 22,Section II.F.1, Attachment 3, and Supplement 3 to NUREG-0797 (same section) documented the staff's review of this item.
The staff concluded that the monitors meet the provisions of NUREG-0737 and are acceptable.
The NRC inspector verified that the monitors' availability and calibration are assured by their inclusion in Table 3.3-6 of the Final Draft Version of the Technical Specifications.
The NRC inspector needs additional information before this item can be closed.
Since the monitor indicators in the control room are digital, the range of the monitors is not apparent.
The NRC inspector needs to review documentation to verify that monitor range is 1 R/hr to 1E7 R/hr Gamma.
I d.
(Open) TMI Action Item II.F.1.4:
Containment pressure.
This item involved the provision of pressure measurement and indication capability of three times design pressure of the containment and down to -5 psig.
NUREG-0797, Chapter 22,Section II.F.1, Attachment 4, documented the staff's acceptance of two separate channels of wide-range pressure indication (0-160 psig) and four channels of narrow range pressure indication (-5 to + 60 psig) at the
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main control board.
The NRC inspector verified that the narrow range indication meets the range accepted by the staff.
However, the wide-range pressure indication on the control board indicates from 0 to 150 psig.
The NRC inspector notes that the operability and.
calibration of the narrow range indicators are assured br their inclusion in Table 3.3-6 of the Final Draft Versich of the Technical Specifications.
The NRC inspector needs to review the applicant's
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provisions for assuring operability and calibration of the wide-range indicator.
This item remains open pending completion of this review and acceptable resolution of the discrepancy concerning widc+rangc instrument range.
e.
(Open) TMI Action Item II.F.1.5:
Containment Water Level Monitor.
This item involved the provision of a continuous indication of containment water level in the control room.
- NUREG-0797, Chapter 22,Section II.F.1, Attachment 5, states that the applicant has committed to provide narrow-range (0 to 30 in.) and wide range (0 to 160 in.)
containment water level indications on the main control i
board.
Both the narrow-range and wide range instruments were to be designed to meet the provisions of Regulatory Guide 1.97, Revision 2.
Based on this commitment, the
staff determined that the CPSES Containment Water Level Monitor was acceptable.
The FSAR,Section II.F.1, indicates that the CPSES design will include redundant wide range containment level indication (808' - 3" to 817' - 111") on the main control
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board.
The indication is to meet the provisions of
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Regulatory Guide 1.97, Revision 2.
The design is c.lso to include normal sump level indication (0 to 3 feet or 36")
on the Main Control Board.
The NRC inspector infers that the normal indication will not meet either Regulatory Guide 1.97, Revision 2, or Regulatory Guide 1.89 requirements.
The FSAR commitments do not appear to meet the commitments identified and accepted by NUREG-0797 nor the provisions of NUREG-0737.
This item remains open pending applicant resolution of the conflicting commitments and NRC review and acceptance of the correct commitment.
f.
(Open) TMI Action Item II.F.1.6:
Containment Hydrogen
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l Monitor.
This item involved the provision of a continuous indication in the control room of containment atmosphere hydrogen concentration.
NUREG-0797, Chapter 22,
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Section II.F.I, Attachment 6, documented the staff's acceptance of redundant indication with a range of
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-0% to 10% at the. main control board.. The NRC' inspector-verified that the installed instrumentation has the range specified.
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The NRC inspector notes that the containment hydrogen
. monitors'are not included in Table-3.3-6.of.he Final Draft version of the technical specification.
The NRC inspector needs to review the applicant's provisions for assuring the operability and calibration of this= indicator. -This item remains open pending this review.
g.
(Open) TMI Action Items II.F.2.1.A and II.F.2.3:
Instrumentation for Detection of Inadequate Core Cooling.
-The applicant's response to this item was included in the FSAR. dated. January' 30, 1981.
Several items were left open
'and were to be provided at a later date.
The applicant
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submitted further information on January 3, 1984, and May 21, 1984.
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Supplement 6 to-NUREG-0797 documented the staff's review of the additional information.
The staff determined that the applicant's detection system for Inadequate Core Cooling (ICC) is acceptable including an exception to position C.6 in Regulatory Guide 1.118, Revision 2, which allows'for the lifting of some core exit thermocouple system leads for testing.
The. applicant'had stated that the instrumentation for the detection of inadequate core cooling (ICC) would be installed prior to fuel load except for the_ Heated Junction Thermocouple (HJTC) system which was to be installed prior to startup from the first refueling outage.
The staff determined that this schedule was acceptable.
The staff's, determination of acceptability was subject to the following:
Technical Specifications relating to the final ICC
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instrumentation are submitted and approved.
An implementation letter report mast be provided for
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staff review 90 days following completion of the preoperational testing of the HJTC system at the beginning of Cycle 2.
By letter dated January-22, 1988, the applicant submitted a revised response to NUREG-0737, Item II.F.2.
The applicant revised the implementation schedule such that the HJTC system will also be operational prior to fuel load.
In addition, the applicant identified two deviations from generically approved systems.
The first
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deviation involves the lack of a backup display of Core Exit Thermocouple (CET) temperatures.
The second deviation involves the periodic testing of instrumentation-channels.
Prior to closecut of this item, the NRC will review this submittal for acceptability.
The NRC inspector verified that the instrumentation for i
the subcooling monitors, the CETs, and the HJTC system as l
installed in the plant has the range as committed and that the~e instruments are included in Table 3.3-6 of the Final Draft Technical Specifications.
i These items remain open pending NRC review of the applicant's January 22, 1988, submittal.
h.
(Closed, Unit 1 only) TMI Action Item II.G.1.1:
Emergency Power for Pressurizer Equipment.
NUREG-0797, Chapter 22,Section II.G.1, documents the staff's review of the
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applicant's response to this item.
The staff determined that:
The Power Operated Relief Valves (PORVs) motive and
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control components can be supplied with emergency power if offsite power is not available.
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The PORV block valves' motive and control components
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can be supplied with emergency power if offsite power is not available.
This emergency power is from a different bus than that which supplies emergency power to the PORV.
The PORV and PORV block valve (ncnsafety) are
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connected to their respective emergency buses (safety-related) through devices that are qualified in accordance with safety-grade requirements.
The pressurizer level indication circuits are
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safety-grade and post-accident qualified.
The power for these circuits is supplied from emergency buses with automatic backup from the emergency buses as committed.
The staff concluded that the design for providing emergency power for pressurizer equipment is consistent with NUREG-0737 and is acceptable.
The NRC inspector verified that the installed pressurizer level instruments are supplied from the emergency buses as committed.
The inspection requirements of Temporary Instruction
2513/65 are considered complete and this item is closed for Unit 1.
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1.
(Open) TMI Action Item II.K.l.3:
NRC Bulletins.
NUREG-0737 lists three NRC Bulletin actions that are applicable to Westinghouse plants.
These actions are
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discussed below along with the applicant's commitments and j
the actions needed to close this item out.
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Action:
Review all valve positions, positioning
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requirements, positive controls, and related test and maintenance procedures to assure proper engineered safety feature (ESF) functioning.
Applicant Commitment:
The applicant in the FSAR,
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Section II.K.1 committed to the above action.
Closecut Action:
NUREG-0797 noted that an operating
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license would not be issued until the NRC verifies that applicant action is complete.
The NRC inspector needs to review documentation that indicates how this review was accomplished, the results of the review, and how subsequent revisions (including the present
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procedure upgrade effort) will be controlled to preserve the essential aspects of this commitment.
Action:
Review and modify, as required, procedures
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for removing safety-related systems from service (and restoring to service) to assure operability status is known.
i Applicant Commitment:
The applicant in the FSAR
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Section II.K.1 committed to the above action.
Closecut Action:
NUREG-0797 noted that an operating
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license would not be issued until the NRC verifies that applicant action is complete.
The NRC inspector needs to review documentation that indicates how this review was accomplished, the results of the review, and how subsequent revisions (including the present procedure upgrade) will be controlled to preserve the essential aspects of this commitment.
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Action:
Trip pressurizer level bistable so that
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pressurizer low pressure (rather than pressurizer low pressure and pressurizer low-level coincidence) will
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initiate safety injection.
Applicant Commitment:
The applicant in the FSAR
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Section II.K.1, states that pressurizer low-level trips are not utilized.
Closecut Action:
The NRC inspector verified that the
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pressurizer low-level trip is not utilized.
The NRC inspector had no further questions on this issue.
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This TMI Item remains open pending NRC review of the two issues identified above.
j.
(Closed) TMI. Action Item II.K.3.1.B:
Automatic Power Operated Relief Valve (PORV) Isolation-Test / Installation.
Supplement 6 to NUREG-0797 concluded that the provisions of NUREG-0737 are met with the existing PORVs, Safety Valves (SVs) and high-pressure reactor trip setpoints.
Therefore, there is no need for an automatic PORV isolation system at Comanche Peak.
The inspection re-quirements of Temporary Instruction 2515/65 are considered complete and this item is closed for Unit 1 and Unit 2.
k.
(Open) TMI Action Item II.K.3.5.B:
Modification for automatic trip of reactor coolant pumps (RCPs).
Supplement 12 to NUREG-0797, documented the staff's conclusion that additional information specific to Comanche Peak was needed.
By letter dated February 17, 1986,.the applicant submitted their response to Generic Letter 85-12, " Implementation of TMI Action Item II.K.3.5.
Automatic Trip of Reactor Coolant Pump." _By letter dated May 2, 1986, the NRC documented a conference call and requested supplementary information on this subject.
This item remains open pending NRC review of the supplementary information, completion of applicant action, and NRC review in accordance with Temporary Instruction 2515/65.
1.
(Open) TMI Action Item II.K.3.9:
Proportional Integral Derivative (PID) Controller Modification.
This item involved modifying the controller to remove the derivative function.
This will reduce the number of challenges to l
the PORV as a result of noise on the pressure signal.
The f
applicant in FSAR,Section II.K.3.9 indicates that the controller has been modified to eliminate the derivative
function.
l The NRC inspector needs to review the completed design change authorization for this modification prior to i
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closecut of this item.
m.
(Closed) TMI Action Item II.K.3.10:
Proposed anticipatory trip modification.
This item was only directed at those licensees / applicants who proposed to modify their antici-patory trip function (reactor trip on turbine trip) such that its use was confined to high-power levels.
The item involved ensuring on a plant specific basis that the probability of a small-break loss of coolant accident-resulting from a stuck-open PORV is substantially unaffected by the modification.
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The NRC inspector verified that the anticipatory, trip function is-not blocked above 10% power.
Therefore, this item is not applicable.
The inspection. requirements of Temporary Instruction 2515/65 are considered complete and this item is considered closed for Unit 1 and Unit 2.
n.
(Closed) 'DMI Action Item II.K.3.12.B:
Confirm existence-of anticipatory reactor trip upon turbine trip.
This item-involved the installation of an anticipatory trip'func-tion.
The NRC inspector verified that CPSES has the anticipatory trip function (also see paragraph m. above).
Therefore, no modification is necessary.
The inspection requirements of Temporary Instruction 2515/65 are con-sidered complete and this item is closed for Unit 1 and Unit 2.
o.
-(Closed) TMI Action Item II.K.3.25.B.2:
Effect of loss of AC. Power on Pump Seals.
This item involves ensuring that RCP seal cooling is maintained following a loss of offsite
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power.
NUREG-0797 states that an acceptable solution is to supply emergency power to the component cooling water (CCW) pump.
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The NRC. inspector verified that the CCW pump is automatically loaded on the emergency-diesel generator on a loss of offsite power.
Therefore, no modifications are required..The inspection requirements of Temporary Instruction 2515/65 are considered complete and this item is closed for Unit 1 and Unit 2.
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p.
(Open) TMI Action Item III.D.1.1.1:
Primary coolant outside containment / leak reduction.
This item involved implementing a program to. reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels.
In the FSAR,Section III.D.1.1, the applicant commits to
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the following:
At intervals not exceeding refueling outages and most
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probably quarterly (every 92 days) operating pressure tests will be performed on appropriate portions of residual heat removal (RHR), safety injection (SI),
core spray (CS), coolant volume control system (CVCS)
(emergency core cooling portion), primary sampling (PS), and waste gas systems.
The initial leakage rates of the systems will be
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determined by hydrostatic (or pneumatic) test prior to plant operations.
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The primary testing method will be by' system walkdown
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at normal operating pressure with quantified measure-ments cbtained at all visually observed leakage paths.
Systems.not readily testable in this manner will require either leakage makeup or pressure drop testing.
Individual leakage rates in gaseous systems
. will be quantified most probably by helium leak detectors.
~ The limiting-leakage value will be initially
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established at 10 gpm with no more than 5 gpm from any one system.
NUREG-0797 documented the results of the staff's review of this item.
The staff found that the proposed leak reduction program covers the necessary systems and subsystems and the frequency of retesting is acceptable.
Supplement 4 to NUREG-0797 documented the staff's conclusion that the applicant's limiting leakage value for liquid cystems is acceptable and that the leakage value for gaseous systems was to be determined by pneumatic tests.
The NRC inspector needs additional information to close
out this item as follows:
The.NRC inspector needs to review the applicant's
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initial leakage test program and results.
The NRC inspector needs to review the applicant's
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program for maintaining the leakage value as low as practical.
The NRC inspector needs to review the applicant's
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plans for conducting the leakage tests during the prestart test program.
This item remains open pending NRC review of the items identified above.
q.
(Open) TMI Action Items III.D.3.3.1 and III.D.3.3.2:
Provide means for detecting and accurately measuring radiciodine.
This issue involved providing a capability to detect and measure radiciodine under accident conditions.
NUREG-0797 and Supplement 6 to NUREG-0797 documented the staff's review of the applicant's response to this item.
The staff concluded that the applicant's description of their plans for performing iodine analysis under accident conditions are acceptable.
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In'FSAR,Section III.D.3.3, the applicant committed to having:
All procedures and instructions associated with l
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in-plant iodine monitoring under accident conditions prepared, approved, and implemented prior to Unit 1 fuel load.
All training programs and procedures associated with
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sample acquisition, sample quantification and data evaluation pertaining to in-plant iodine monitoring activities under accident conditions shall be i
prepared, approved, and implemented prior to Unit 1 fuel load.
The NRC inspector needs to review the training programs, procedures, and instructions described in the applicant's
commitments prior to closecut of this item.
This item
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remains open pending the completion of.this review.
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8.
Plant Tours (71302)
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The NRC inspector reviewed documentation and performed a walkdown of the turbine generator primary water system in advance of the applicant adding tritium to this system.
Tritium is used as a tracer to detect water leakage into the
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turbine generator cooling gas (hydrogen).
The NRC inspector notes that although this system contains radioactivity, it is not a radwaste system and; therefore, it does not clearly fall under 10 CFR Part 50, Appendix A, Criterion 60 and the applicant's commitments regarding radwaste systems in Chapter 11 of the Final Safety Analysis Report.
The NRC inspector needs more information concerning the design basis of the primary water system and applicable design standards before a determination can be made as to whether the design adequately considers the radioactive contents of this system.
This item will be tracked as an unresolved item (445/8924-U-02).
The NRC inspector noted that the demineralized water system provides makeup to the primary water system through two hard piped fill lines.
One fill line connects to the high pressure portion and one fill line connects to the low pressure portion of the system.
The NRC inspector is concerned that leakage through the provided check valve and isolation valve would allow contaminated water to flow into the demineralized water system.
In response to this concern, the applicant installed a temporary blank flange in the fill line to the high pressure portion of the system.
The fill line to the low pressure portion is used for normal makeup and could not be blank flanged.
Before the NRC inspector can determine the adequacy of the applicant's action, the NRC inspector needs to:
(1) review the applicant's procedural controls over removal and
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installation of the blank flange to ensure the blank flange is not' removed without the necessary approvals and system conditions and to ensure that the blank flange is reinstalled.
when required,-(.) review the applicant's procedure ~ controls
over the systems to ensure that the fill line pressure is always higher than the primary water system pressure at the fill. point, (3) review the isolation features between the two systems to ensure that NRC guidance concerning cross
. connections with radioactive. systems has been adequately considered.
These items will be tracked as an open item (445/8924-0-03).
During walkdown of.the system, the NRC inspector noted that the turbine plant cooling water system inlet header to the primary waterisystem coolers appears to move excessively as a result'of normal system' water turbulence.
Several small drain lines are supported by the inlet header and as a result exhibit excessive
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movement and vibration.
At.least one of these drain lines is from the primary water system.
The NRC inspector is concerned that this movement could cause fatigue failure of the drain lines or the' coolers and and result in a spill of tritiated water.
This item will be tracked as an open item (445/8924-0-04).
9.
Open Items open items are matters which have been discussea with the applicant, which will be reviewed further by the NRC. inspector, and which involve some action on the part of the-NRC or licensee or both.
Two open items disclosed during the (_
inspection are discussed in paragraph.8.
10.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or deviations.
One unresolved item disclosed
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during the inspection is discussed in paragraph 8.
11.
Exit Meeting (30703)
An exit meeting was conducted on May 2, 1989, with the applicant's representatives identified in paragraph 1 of this report.
No written material was provided to the applicant.by i
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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 this meeting, the NRC inspectors summarized the scope and findings of the inspection.
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