ML20209A973
| ML20209A973 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/22/1987 |
| From: | Domer J TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8704280327 | |
| Download: ML20209A973 (28) | |
Text
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.K TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 37401 SN 1578 Lookout Place APR 221987 U.S.- Nuclear Regulatory Comission ATTN: Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of
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Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT - NRC INSPECTION AND ENFORCEMENT REPORT NUMBERS 50-327/86-55 AND 50-328/86-55 The subject inspection report on the Design Baseline and Verification Program (DBVP), transmitted to TVA on February 3, 1987, has been reviewed by TVA.
As requested in the inspection report, responses to the identified observations are' enclosed (enclosures 1 and 2).
Open items from the following NRC Inspection and Enforcement Reports are addressed:
(1) 50-327/86-27 and 50-328/86-27; (2) 50-327/86-38 and 50-328/86-38; (3) 50-327/86-45 and 50-328/86-45; and (4) 50-327/86-55 and 50-328/86-55.
Information provided herein or activities associated with it are available for inspection by NRC.
New comitments are identified in enclosure 3.
To the best of my knowledse, I declare the statements contained herein are complete and true.
Very truly yours, TENNESSEE VALLEY AUTHORITY J. A. Domer, Assistant Director Nuclear Safety and Licensing Enclosures cc:
See page 2 l
gi 8704280327 870422 p' ' g PDR ADOCK 05000327 G
PDR An Equal Opportunity Employer
. U.S. Nuclear Regulatory Commission cc (Enclosures):
Mr. G. G. Zech, Assi. Leant Director for Inspection Programs Office of Special Projects U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 l
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5 ENCLOSURE 1 1
The TVA responses to observations identified in NRC Inspection and Enforcement j
(IE) Report Nos. 50-327/86-55 and 50-328/86-55 follow and are identified by number and title of the observation.
The NRC observation t' ext is not repeated within this enclosure. Observations from previous inspections are addressed in enclosure 2.
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i Observation No. 2.4 - Design Criteria vs. Final Safety Analysis Report (FSAR)
Component Cooling System Heat Loads-I
Response
i j
A January 2, 1987 memorandum, J. C. Key to H. L. Jones, transmitted the requested FSAR revisions. The requested FSAR revision includes reduced
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Component Cooling Water System (CCS) heat loads and raw cooling water flow j
rates, and it is currently in the review and approval' cycle for inclusion during the current annual FSAR update. However, the Essential Raw Cooling
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Water (ERCW) was not addressed in that previous requested revision.
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Therefore.an additional revision to the FSAR to include revised ERCW flow j
rates and heat loads will be initiated and submitted with the 1988 FSAR update.
j Observation No. 2.5 - Reactor Coolant System Design Criteria vs. Technical Specification (Four Loop Operation)
Response
The Reactor Coolant System (RCS) design criteria originally indicated that the plant is designed for critical operation with less than four' loops a
operational. As the original design included three-loop operation but the operating license only allows four-loop operation, this information could have been misleading to someone unfamiliar with the plant technical specification.
TVA has revised the design criteria (DC), SQN-DC-V-27.4.
l
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Observation No. 2.6 - Flow Rate Assumption Used in Calculation
Response
j i
In the preparation of Calculation EPM-SMJ-022886, performed to determine the j
design pressure of the Refueling Water Storage Tank (RWST) suction header, the I
engineer was required to determine the maximum flow through this header. The l
engineer was aware that System operating Instruction (SOI)-74.1D limits flow l
through this header by restricting flow through valve FCV-74 during RWST~
refill. To be conservative, however, the pump runout condition, as taken from I
the Residual Heat Removal (RHR) pump drawing (V-8x20WD36X1 Contract 91934),
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was used in the calculation. The actual runout condition was not verified by i
development of a system curve for the calculation. The actual runout i
condition must be determined in order to ensure that the flow and the related pressure are within the assigned design conditions. Calculation
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EPM-SMJ-022886 has been revised.
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' Observatica No. 2.7 - Drawing Control Respons,e 1.
It has been observed that the locations of temperature ' nd pressure a
changes have not been consistently applied; i.e.,
these changes in some cases are shown to take place on the wrong side of a valve. This could lead to the installation of a component of the wrong pressure class or a piping analysis which is in error.
In order to ensure that an error or an inconsistency in the application of i
these pressure or temperature changes has not led to installation of inadequate components or invalid analyses, a review of all mechanical flow diagrams was performed. Condition Adverse to Quality Report (CAQR)
SQP870194 was generated to resolve differences between the pressure and temperature designations shown on the flow diagrams and the actual valve ratings as used in the systems.
If a discrepancy had a potential for causing inadequate material to be used or would invalidate analyses, the condition was. investigated.
Two problems were identified and have been documented in Problem Identification Report (PIR) SQNMEB8770 and CAQR SQP870192. These CAQs y
identify discrepancies between the temperatures and pressures shown on the i
drawings and the actual valve ratings provided by the manufacturers. They l
will be dispositioned in accordance with the TVA CAQ process.
I 2.
A review of Waste Disposal System (WDS) drawings (47W560 Series) indicates i
that drafting errors and inconsistencies exist. This is due to the excessive number of times these drawings have been revised and to the complexity of the system. Although TVA has found that other drawings j
contain an occasional drafting error, which is corrected when found, these errors seem to occur more frequently on WDS drawings.
A PIR (SQEMEB8718) was written to identify discrepancies on this series of drawings. A review of these drawings was made during the development of this PIR.
None of the discrepancies identified in this review and spelled out in the PIR have any impact on system function. These discrepancies, as well as any other drafting errors discovered while revising the drawings, will be corrected as the Configuration Control Drawings (CCDs) are made in accordance with Sequoyah Engineering Procedure (SQEP)-17.
These drawings are scheduled to be completed by October 1, 1987.
SQEPs-13 and -17 were developed to control the design change for modifications.
In the future, changes will be done on Drawing Chango Authorizations (DCAs) which should preclude recurrence of this problem.
I i.
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- Observation 3.5 - Vent Condenser Heat Exchanger Flange Installations t
Response
This observation relates to ECN L6499.
3 Concern lA:
1(A).1 An examination of the analysis isometric (Calculation N2-70-A-301A) reveals that the coordinate system is inconsistent with that shown on the mechanical isometric (Drawing 47W464-13, Rev. 21).
The Z' direction shown does not follow the right-hand rule, and the 2
directions of the two north arrows are not consistent.
While the Z coordinate direction is reversed, all Z direction piping is also shown reversed. Thus, the piping appears opposite hand when compared to the physical isometric.
The apparent mirror-image condition exists only because of the analysis isometric coordinate system definition. A point-by-point comparison of both isometrics reveals consistent routing, except for that addressed in 1(A).2.
Generically, this pictorial confusion occurs whenever piping is shown in isometric form on the physical drawings, and the coordinate system-on the analysis isometric does not match the one on the physical isometric. A review has been performed of all existing alternate analysis isometrics, and none now utilize a left-hand coordinate system.
The coordinate system definition and drawing have been reworked to
+
l utilize a right-hand coordinate system.
Piping analysts / checkers were
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notified by the section supervisor that a left-hand coordinate system j
is not acceptable practice for future work.
i 1(A).2 The comparisons in 1(A).1 revealed that piping shown in the X-Y plane between elevations 675 feet 8 inches and 676 feet 0 inches on detail F-13 is incorrectly shown in the Y-Z plane on the original analysis isometric (Calculation N2-70-A-301A).
The specific discrepancy with regard to the skewed pipe routing is an i
i isolated case, and the isometric has been reissued to show the skewed piping in the X-Y plane. Analysts / checkers will be informed to pay r
i particular attention as to how skewed piping directions are shown on analysis isometric drawings.
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- _4 Concern IB:
1(B)
The analysis isometric does appear to be shown opposite hand from the unit 1 physical layout. This is because of the condition addressed in 1(A).1.
Boric acid evaporator packages A (unit 1)'and B (unit.2) are not physically opposite hand (reference:
Chemical and Volume control
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System (CVCS) Drawing 47W555-7, Rev. 21), but are side by side as j
shown on Drawing 47W464-8,' details B8 and C8.
One isometric, detail D8 applies to both units. Therefore, the analysis appears to be opposite hand from the unit 1 physical isometric, but for the same reasons as indicated for unit 2 (see concern 1(A) above).
Because unit 1 is not opposite hand from unit 2, this concern does not represent a problem. Consistent' coordinate system definition eliminates the probica.
Concern 2:
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l A TPIPE Code thermal evaluation was performed, but an error was discovered in the checking process. Because a thermal evaluation of piping with a maximum temperature of less than or equal to 200 F is 0
i not required, per the alternate analysis review program restart criteria, the thermal evaluation was not corrected.
This is an isolated case that went beyond the scope of the alternate j
analysis required. Because of this specific concern, a thermal j
evaluation has been performed and checked. The evaluation confirms j
the acceptability of the original qualification and the calculation package will be revised postrestart, in accordance with SQN-AA-001, to l
include the thermal evaluation and qualification.
I i
l An examination of the original analysis geometry revealed that the I
piping was modeled correctly. This was also confirmed by a computer plo.t of the taath model geometry that matched the configuration of detail F13 on the mechanical isometric drawing (47W464-13 Rev 21).
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Also, a thermal evaluation was performed and checked to confirm the
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acceptability of the initial qualification. With the corrective action'and action to prevent recurrence established for the isometric il deficiencies identified, the implications to safety and quality of design are minimal in regard to the concerns of this observation.
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-S-Observation No. 3.6 - Primary Containment Leak Rate Test Lines
Response
i This observation relates to ECN L5750.
The preparer and checker made a judgment evaluation that the slip joint end of the penetration would act equivalent to an anchor. This evaluation was not documented properly. A i
review of penetration drawing 47W331-2 indicated that this is a unique situation in the plant.
Calculations were performed to address this documentation deficiency, and the calculation package has been revised. This confirms t he technical adequacy of this evaluation.
The Sequoyah Rigorous Analysis Handbook will be revised by April 30, 1987, to provide guidance to analysts.
i The potential impact of the identified deficiency is not significant according 4
to criteria contained in Nuclear Engineering Procedure (NEP)-9.1.
It is a document change only. The calculation shows that there is daylight clearance only and the piping within the penetration is rigid; therefore, the inboard headplate acts as an equivalent anchor.
Observation No. 3.7 - TVA-Supplied Limit Switch Brackets
Response
]
This observation relates to ECN L6556.
j 1.
The Division of Nuclear Engineering (DNE) assumed that the Plant Modifications Group would use " common" grade bolts-(ASTM'A307) sized to fit the tapped or drilled holes.
i' This situation has been identified as a generic concern, " Lack of Detail in Design Output," by the Gilbert / Commonwealth (G/C) report
. Technical Review of Sequoyah Nuclear Plant Modifications.
j dated March 3, 1986.
1 The Electrical Engineering Branch (EEB) will revise drawing 47A348-260, 3
Rev. 2, to show bolting sizes and material specifications (Refer to Field
{
Change Request (FCR] 4289, Rev. 1, ECN L6556).
DBVP SQEP-45 punchlist item 9493 (Sequoyah unit 2) has been written to identify and correct the deficiency. 'This item will be dispositioned in accordance with SQEP-45.
The bolt size and material specifications called out on FCR 4289, Rev. 1, j
agree with the "conson" grade assumption.
DNE calculction demonstrates j
that the bolts are adequate.
The Sequoyah Project Manual and the Civil Engineering Branch (CEB) i Interface Review Procedure have been revised to ensure that, in the
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future, sufficient details will be shown on TVA and modified vendor j
drawings.
(Refer to G/C Report No. 2614, Technical Issue Data Sheet 1
Nos. 4 and 19, and TVA letter from R. L. Cridley to J. Nelson Crace dated July 28, 1986, addressing Deficiency D3.1-1.)
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- i 2.
Lack of seismic qualification for components shown on drawing 47A348-260, Rev.1, is not a valid concern because the seismic qualification evaluation of minor drawing changes was performed in accordance with the applicable TVA engineering procedure (EN DES EP 3.02) at the time of drawing issue. Component Seismic Qualification Enginee'es reviewed 47A348-260, Rev. O, and indicated approval by initialling the drawing.
Drawing 47A348-260. Rev. 1, was issued for FCRs 2645 and 2847 (ECN L5881).
In accordance with EN DES EP 3.02, EEB assessed the FCRs.
However, seismic qualification of components shown on drawing 47A348-287, Rev. O. (ECN L6556) has not been documented (review dated i
November 19, 1986). The apparent lack of seismic qualification documentation for 47A348-287 Rev. -0, is an isolated case because the FCR 4216 was open and was being tracked (in accordance with SQEP-AI-11A) by the FCR Late Report at the time of this observation. Seismic 4
)
qualification documentation was issued for drawing 47A348-287 Rev. O, on i
December 15, 1986, to address the deficiencies identified by the DBVP in l
PIR SQNCEB8657.
Interface Review Procedure CEB-DI-121.03 has been revised to require coordination of FCRs and is endorsed by the Sequoyah Project Manual.
Training classes have been conducted for CEB-DI-121.03.
SQEP-AI-11A has j
been revised to require written FCR approval and prohibit informal verbal-l approval. The Sequoyah Project Manual and the CEB Interface Review 1
Procedure have been revised to ensure that sufficient details will be
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shown on TVA and modified vendor drawings.
(Refer to note above in item 1, regarding G/C Report and TVA response.) A representative sample of past design changes was evaluated. The evaluation established a 95-percent confidence level that seismic design is adequate.
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The implications to safety are minimal, based on the representative sample j
of past design changes. Drawings 47A348-260, Rev. O and Rev. 1, were reviewed and approved in accordance with the applicable engineering
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procedures at the time of drawing issue.
ObservationINo. 3.8 - Solenoid Valve Mounting Seismic Qualification Responsa j
This observation relates to ECN L6487.
1.
Hanger location isometric 0-WD-215, Rev. 1, is incorrect. A walkdown shows two segments of piping installed opposite hand from what is shown on l
drawing 0-WD-215.
The analyst used the hanger location isometric to evaluate this piping and was not aware that the pipe routing was wrong.
Based upon the pipe routing and support scheme information provided by the walkdown, the engineering evaluation is still valid.
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. Walkdowns that have been performed show that the hanger location isometrics have the routing correct. However, hanger location drawing (0-WD-215) has been revised and documented in analysis problem No. N2-77-A-314A.
Further walkdowns are required for future work, in accordance with SQEP-13.
The item has negligible impact to the analysis; therefore, no other consideration is necessary at this time. All piping in safety-related systems will be walked down and analytical isometric drawings provided in Phase II (postrestart) of the CEB Alternate Analysis (AA) Program, in accordance with SQN-AA-001.
2.
The screening evaluation performed on valve 2-FCV-77-20 is in accordance with AA program instructions SQN-AA-006, paragraph 3.5.1, and SQN-AA-001, paragraph 3.3.2.
The fact that the span of six feet nine inches exceeds the two-foot-maximum span given in one of the three previously used
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criteria does not establish a stress or support load problem. This was reviewed and checked by experienced stress analysts and was determined not to be a problem (reference: problem No. N2-77-A-314A).
Because most all the support loads would be generated by the valve, the increased contribution from the additional piping is negligible. Because the support was only six inches from the center of the valve and the spectra acceleration for this model would be less than one g, there was no concern for qualification of the pipe, supports, or the valve. The operator was supported in two directions.
In accordance with SQN-AA-001 dated July 1, 1986, paragraph 3.4.3, large concentrated weights are to be addressed as a potential deficiency in the postrestart (Phase II) part of the AA program.
The screening process was typical of several evaluations. This analysis meets all programmatic requirements and is analytically adequate as determined by the evaluation.
Discussions with the NRC Office of Nuclear Reactor Regulation (NRR) have taken place regarding this observation. A telephone conference call on December 16, 1986, between representatives of TVA and NRR concluded that TVA's defined AA program, with inclusion of five confirmatory items, is acceptable.
This decision is reflected in NRC's draf t Safety Evaluation Report (SER) on the Interim Acceptance criteria for Small Bore Piping (reference:
NRC letter from B. J. Youngblood to S. A. White dated December 19, 1986).
3.
The cause of item 3 resulted from a breakdown in the interdiscipline coordination of configuration changes to components. The conduit outlet body and one-inch conduit were not shown on the sketch attached to the FCR. The designor evaluating the support variance did not include the weights of the above items because he was not aware of such an attachment.
. Calculations SQEP-C2-37 and SQEP-C2-L6787 have been issued to document the testing and qualification of the as-built configuration. Calculations on FCR Variance No. 47A054-33A-A174 have been revised to reference the qualification testing.
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Implementation of SQEP-13 ensures that structural components such as supports, approved per FCRs, are installed in accordance with the evaluation performed.
Because testing was performed to qualify the as-built configuration, along with issue of QIR-CEB-87-048, which has been issued to limit the size of conduit junctions attached to electrical devices, the identified deficiency will not jeopardize plant safety.
4.
The cause of item 4 is that the FCR cover sheet did not indicate that a component seismic qualification engineer was involved. Therefore, the FCR was closed without issuance of seismic qualification documentation. The problem should be treated as an isolated case because TVA has established a 95-percent confidence level of seismic design adequacy for changes to safety-related components and equipment at Sequoyah.
A calculation has been issued to address this specific deficiency.
Interface procedure CEB-DI-121.03 has been revised to include coordination of FCRs. CEB-DI-121.03 is endorsed by the Sequoyah Project Manual.
Training classes have been held for proper implementation of procedure CEB-DI-121.03.
The implications to safety and quality of design are minimal because TVA has qualified similar solenoid installations by test.
5.
The cause of item 5 is a lack of forethought related to the depth of detail required in the FSAR.
This section of the FSAR describes criteria related to field routed process piping and instrument lines that do not require complete analysis as defined in paragraph 3.9.2.5.
The FSAR will be revised in the 1988 update to address the use of copper tubing for field routed instrument lines.
The potential impact of the identified deficiency is negligible because typical support detail drawings identify requirements for field location and support of copper tubing based on the design data and qualification procedures established in CEB Reports 75-9 and 80-05.
Observation No. 3.9 - High Energy Line Break Designation
Response
This observation relates to ECN L5724. A review of piping isometrics (47K432-Series) revealed the following conclusion:
the energy level was incorrectly identified only on isometric 47K432-51. Rev. 4.
No additional discrepancies were identified and this appears to be an individual error on a specific isometric with no generic implications.
Isometric drawing 47K432-51, Rev. 4, identified in the subject action item, will be revised to show correct energy level per Sequoyah Rigorous Analysis
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Handbook, SQN-RAH-103.
DBVP-SQEP-45 punchlist item 0057.(Sequoyah unit 1) has been written to identify and correct this deficiency.
It is a document change only, and will be dispositioned in accordance with SQEP-45 and NEP-9.1 criteria. A copy of this response will be routed through the appropriate section to all piping analysts by April 30, 1987, to more accurately identify the need for quality consciousness in the preparation of-isometric drawings.
Section training and revision of the checklist in the Rigorous Analysis j
Handbook (providing guidance to the analyst) to be conducted by April 30, 1987, will ensure that isometries will be reviewed for pipe rupture j
considerations.
The action to correct will ensure that each analyst is made aware of the need to coordinate pipe rupture evaluations.
This will prevent j
recurrence of this discrepancy.
Because this was found to be an isolated occurrence and the corrective action identified above ensures adequate future review, the concern has no impact on i
plant safety.
Observation No. 5.7 - Diesel Breaker Trip
Response
i j
EEB concurs with the findings in Observation 5.7 and has written a CAQ to address the concern. The CAQ number is Significant Condition Report (SCR)
SQNEEB86206RO. The corrective action for the CAQ has not been determined at this time, but the CAQ will be dispositioned in accordance with the TVA CAQ j
process.
Observation No. 5.8 - DBVP Review Per Procedure SQEP-12 Checklist i
Response
i i
The NRC inspection team's review of the completed SQEP-12 checklist indicated i
that electrical attributes such as voltage drops, thermal rating of cables, and accuracy of instrument loops were not evaluated (as denoted by "not j
spplicable" [NA) on the checklist) and documented on the SQEP-12 checklist.
The team also found that the reviewer assumed the existence of i
calculations / analysis which addressed these attributes, but failed to verify I
1 this and failed to provide a cross-reference. All checklists prepared by EEB were affected.
They required rereview and retision to ensure that this problem, as well'as other EEB checklist problems were corrected. Checklists a
prepared within the mechanical, nuclear, and civil disciplines were found to j
be unaffected.
i The checklists identified as deficient have been revised.
Also, all i
checklists prepared by EEB were revised to correct the deficiency of l
improper / incomplete checklists prepared by their personnel. The personnel l
involved with preparation of EEB checklists were retrained in the proper manner and degree for preparing and documenting SQEP-12 checklists.
l Appropriate cross-references to other Sequoyah programs, e.g., Equipment j
Qualification Program (EQP), are provided as necessary and documented on the SQEP-12 checklists.
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The DBVP scope of verification of technical adequacy when interfacing with other TVA programs is as follows:
The DBVP took credit in its assessment, where appropriate, for technical work in several other DNE programs. The programs include: Weld'ing Project, Alternate Analysis, DNE Calculations Effort, and Equipment Qualification.
These interfaces are referenced in SQEP-12.
Observation No. 6.9 - Cable Design Criteria Temperature Limits
Response
As a result of the latest environmental qualification effort, environmental parameters were redefined. Regardless of the original parameters, which were stated in Design Criteria SQN-DC-V-11.3, all cable required to be qualified has been evaluated for compliance with 10 CFR 50.49 and properly dispositioned by Sequoyah EQP in accordance with Project Manual (PM) EQP-01.
This PM required evaluation of documentation against the 47E235-Series Environmental Data Drawings (most recent data) and a baseline list of cables issued. All ECNs are reviewed by SQNEQ (refer to SQEP-ll) for applicability of EQ requirements in the change process.
The cable yard inventory was reviewed, and nonqualifiable cable identified for removal from the site. Cable which may be qualifiable and is available for modification was also identified.
Each future replacement or new installation in a 10 CFR 50.49 circuit requires evaluation and qualification to the environmental parameters at the specific installation location.
Because the generic parameter values given in SQN-DC-V-11.3 are no longer applicable SQN-DC-V-11.3 has been revised to refer a user to the Environmental Design Criteria SQN-DC-V-21.0 and the 47E235-Series Environmental Data Drawings for the appropriate environments. Parameter values were deleted from SQN-DC-V-11.3.
Observation'No. 6.10 - Instrument Calibration Data Consistency
Response
This observation identified an inconsistency in the recording of technical data for instrument calibrations in workplan 11916 (ECN L6551).
It was also noted that the design basis for the calibration tolerance was not referenced on the calibration data sheets.
The taking of "as found" technical data is required during plant surveillance testing of safety-related protection equipment, as defined by IEEE 279-1971 and NQAM, Part III, section 4.5, paragraph 3.4.9.
The applicable surveillance and/or calibration test procedures at Sequoyah are consistent with this requirement. However, this observation involves a modification which changed the instrument setpoint. The original setpoint was considered inadequate for the demonstrated accuracy of the equipment.
Therefore, taking "as found" data for the original setpoint was unnecessary.
"As found" data for the equipment in the workplan was collected as a good practico, and omission of the data for some equipment was a minor oversight which does not compromise plant safety features.
- i The design basis for the calibration tolerance is documented in instrument accuracy calculations which are part of.the Environmental Qualification Binder I
SQNEQ-ITS-001. To prevent recurrence, the instrument engineers will be instructed per Sequoyah procedure AI-19, Part IV, to reference any source documents used to develop any accuracies, setpoints, etc.,'used in workplans.
I This procedure will be revised by April 30, 1987.
Observation No. 6.11 - Environmental Effect Calculation for In-Containment Non-Class-1E Loads Connected to Safety-Related 480-Volt f
Shutdown Boards i
Response
1 l
The analysis to-identify nonsafety-related electric equipment important to j
safety, as required by 10 CFR 50.49 b(2), was performed by two separate calculations. SQN-0SG7-038 analyzed process control interfaces, and j
calculation 10 CFR 50.49 b(2)-1 analyzed distribution equipment and wiring.
The initial issue of the latter calculation addressed the subject changes on a concept level with a subsequent issue on February 25, 1987, for a detailed treatment. The initial issue on November 7, 1985, did not classify any of the subject equipment and wiring as important to safety on the basis of adequate circuit protection for harsh environment induced faults and protective device j
location in a mild environment.
i Harsh environment induced failure mechanisms'as the cause of simultaneous i
faults on the Class 1E auxiliary and control power systens have been analyzed. These faults are considered to occur as singular power train faults with sufficient clearing time before any subsequent faults.
Therefore, there is no detrimental environmental effect for in-containment nonclass-1E loads connected to safety-related, 480-volt shutdown boards i
i The NRC review team noted in this observation that the lower compartment -
cooler (LCC) and the control rod drive mechanism (CRDM) fan motors are not l
tripped on a safety injection signal (Phase A).
The design of the circuitry is such that the motors are tripped on a containment isolation signal i
Phase B.
This differed from the design of the circuitry for the pressurizer heaters which.do indeed trip on a Phase A signal. NRC noted that these nonqualified motors would not be disconnected from their respective 480-volt board before being loaded onto the diesels.
1 containment cooling is provided by both safety-related systems (ice condenser, air return fans, containment spray) and nonsafety-related systems which are designed for high reliability (LCC, CRDM, etc.).
During a large loss of coolant-accident (LOCA) or high-energy line break (HELB) inside containment, a Phase B signal is generated on high-high containment pressure. This, in turn, initiates the safety-grade cooling systems.
For these events, the LCC and CRDM fan motors are load shed (Phase B) per the DC Load Analysis.-
There are other events, however, such as a small break LOCA or an HELB outside containment, which may or may not result in the generation of a Phase B signal. These events would not necessarily challenge the ice condenser j
system, which is a passive heat sink, nor would they initiate containment s
j spray. Thus, the normal heat removal systems are designed to remain operable l
to provide containment cooling. For this reason, the motors for the LCC dnd CRDM coolers are not load shed on a Phase A signal.
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Observation No. 6.12 - Periodic Test of Component Cooling System Surge Tank for Internal Baffle Integrity Response-The original procurement specification for the CCS Surge Tank did not require i
a test to ensure that there was no leakage through the baffle plate. Also, no preoperational test was conducted after installation to ensure that there was no leakage. Because of the fact that the baffle has a continuous fillet weld-on both sides, it'is unlikely that there is any significant leakage through the weld joints. Calculations HCG-BYC-081586 and NCB-BYC-090586 prove that adequate NPSH for the CCS pumps can be maintained with the Surge Tank empty.
Any loss of water through the baffle, however, reduces the amount of water which can be used to make up outleakage from the system through pump seals, etc., in the event that a postulated condition of one failed train could cause the entire tank to drain during accident mitigation operations. A leaking baffle could therefore shorten the time which the system could operate without makeup. Therefore, a test will be performed by May 16, 1987, to determine the extent of any leakage through the baffle plate. PIR SQNMEB87069 has been.
written to resolve, correct, and track this deficiency.
Observation No. 6.13 - Component Cooling System Pump Discharge Pressure Switch Reset
Response
Functional tests will be performed to verify that the alarm operates at setpoint and resets on decreasing pressure. This test will be performed postrestart (by March 31, 1988) because the circuit is an alarm circuit and nonsafety-related.
Concerning the lack of hard copy documentation to verify proper component setpoint, the vendor was contacted and agreed to supply the necessary data sheets. The information was received on November 24, 1986, and is contained in a file memorandum. TVA concurs with these data sheets.
Observation No. 6.14 - Project Evaluation of Electrical Failure Analysic for Procedure SQEP-12, Design Review Checklist
Response
The DBVP group did not initially consider the effect of imposed voltages when answering question B.3.c of the SQEP-12 checklists. Also, the generic note referenced in the response to this question was misleading.
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l To address the question of imposed voltages when answering question B.3.c of the checklists, the DBVP group wrote PIR SQUEEB86171.
The DBVP electrical group has performed a cereview of the e,lectrical SQEP-12 checklists to ensure that imposed voltages cannot have adverse effects on the ability of safety-related systems to perform their safety functions. The results of this receview will be documented in a report to be issued by-April 30, 1987.
Observation No. 6.15 - Periodic Functional Test of Agastat Timer Relays in Pump Motor Start Circuits l
1 Resoonse.
The subject 0.5-second reset timer has been identified as a Critical l
Structures, Systems, and Components (CSSC) component requiring periodic l
calibration and will be calibrated before restart of unit 2.
Calibration of I
this component will also verify that it performs its function.
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The identified 0.5-second reset timer will be incorporated into Sequoyah l
Standard Practice SQE-8, " Control of Installed Permanent Process Instrumentation," by December 31, 1987, for periodic calibration.
TVA has an effort underway to address the larger but related issue of nonidentification of some CSSC components that require calibration.
(Refer to Nuclear Quality Audit and Evaluation Branch [NQA&EB] Audit Report No.
QSQ-A-86-0007 and Licensee Event Report [LER] SQRO-50-327/87010.)
It should be noted that an integrated systems test to verify ~ function of the 1
timer for station blackout followed by safety injection is not warranted.
The TVA position on the diesel loading sequence for loss of offsite power with a delayed safety injection is that such a scenario does not significantly contribute to the probability of core melt. Thus, TVA plans to revise the FSAR in the.1988 update to identify that only coincident diesel generator i
loading sequences are design basis events (reference:
TVA letter from R. Crldley to Stewart Ebneter dated March 12, 1987).
l Observation No. 7.4 - Project Review of Support Variance
Response
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The concern in item 1 was caused by the failure of the checklist preparer to identify the basis or justification for the statenent that no calculations were required or needed for the support variances on FCR 455.
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The concern in item 2 was caused by the checklist preparer referencing the 2.
wrong Records Information Management System (RIMS) number for calculations that approved the support variances for FCR 565.
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- Because both the deficiencies found were contained in the Civil / Structural Section 6.4 of the checklist, cereviewc were performed to evaluate the extent of the concerns in that section. The results follow:
1.
The possible extent of the item 1 concern is that other judgments exist on the checklists that do not reference a basis or justification. To ansure that the extent of this concern does not reveal inadequacies in the program, the Civil DBVP rereviewed all their checklists to ensure that judgments reference a basis or justification. Only four other unjustified judgments out of over 2000 reviews performed by the DBVP were found by the receview, and they have been corrected. Therefore, the extent of this concern is considered addressed.
2.
The possible extent of the item 2 concern is that more improper references for variances exist.
To ensure that this was an isolated incident, the Civil DBVP Group randomly pulled 10 ECN checklists and receviewed the references to verify accuracy. This represents a sample of more than 50 percent of checklists containing references in section 6.4 to variance calculations.
The ECNs receviewed were:
1.
ECN 2419 System 32 6.
ECN L5371 System 65 2.
ECN 2775 System 90 7.
ECN L5684 System 68 3.
ECN 2777 System 68 8.
ECN L5882 System 30 4.
ECN L5009 System 67 9.
ECN L5887 System 62 5.
ECN L5119 System 90 10.
ECN L5979 System 01 No further incorrect calculation references were found; therefore, the reference for FCR 565 is considered an isolated incident.
1.
Even though the initial reviewer probably had a basin for not requiring calculations for the variance, none was referenced or could be found.
Therefore, the checklist has been revised to indicate that no calculations can be found. This has been placed on PIR SQNCEB8639.
2.
The reference for FCR 565 on support variance calculation was incorrect and no additional calculation can be found. Therefore, the checklist has been revised.
The deficiency was placed on PIR SQNCEB8639.'
The DBVP checklists are prepared, reviewed, and approved in an effort to ensure accuracy of the ECN review.
The deficiencies found are considered to be isolated and were corrected, and no further action to prevent recurrence is required. PIR SQNCEB8639 will be dispositioned for noted deficiencies before restart of unit 2.
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ENCLOSURE 2 The following responses relate to observations, unresolved items, or deficiencies identified during previous inspections that are reported by IE Report Nos. 50-327 -328/86-55 to be open. These observations are identified by number and title, with the corresponding IE report number in parenthesis.
Observation No. 1.1 - Impact of Walkdown Findings on Operating Procedures (86-38)
Response
Corrective Action Report (CAR) SQ-CAR-85-10-016 documents the corrective action taken to address this observation. A survey of System Operating Instructions (SOIs) for systems listed in Administrative Instructiori (AI)-37 was accomplished. This survey compared the valve checklist with the latest "as constructed" drawing for possible omissions of valves.
It also checked for possible mislabeling of valves. The completion of corrective action, verification that corrective action was taken, and closure of this CAR provides the documentation that the corrective action procedures and processes accomplish their objectives.
The operating instructions that may be utilized in support of Surveillance Instructions (sis) receive another verification that walkdown findings are incorporated. As each applicable operating instruction receives its review, as required according to SI-1 Appendix F, a reference list of drawings used in preparation or performance of the procedure is generated. This is entered into a computer base.
The drawings, as noted in the drawing deviations (DDs) generated by the walkdown personnel, are compared against the computer base.
The section responsible for the procedure that has drawings common to the procedure and DDs 2 i transmitted a copy of the DD and requested to see if the deviation affects ti.a procedure. This program is documented in Appendix G of SI-1.
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Observation No. 1.2 - Walkdown Scope Difference From Calculation Boundaries (86-38)
Response
NRC noted that the walkdown boundary did not fully match the Design Baseline boundary calculation, SQN-OSG7-048. There is a requirement in SQEP-16 to justify these differences in the System Evaluation Report (SYSTER). This item has remained open pending a review of a sample of boundary justifications.
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Observation No. 1.3 - System Interfaces on Drawings (86-38)
Response
j TVA acknowledges that the present practice of using "out-of-function" information on drawings may be misleading with regard to accurately depicting system interfaces, as noted by the NRC team.
Although this type of information is used throughout the Sequoyah drawing system, the benefits and liabilities of its use are perhaps most notable in the NSSS interfaces.
It is important that this information be available as a -
reference, but TVA also acknowledges that drawing users should recognize the appropriate use of "out-of-function" infoemation.
i Steps have been and are being taken to address this issue. The DNE project policy concerning "out-of-function" information on drawings is contained in a TVA memorandum dated January 30, 1987, from the Sequoyah Project Engineer. This observation is also addressed in TVA's. comments on the DBVP draft Safety Evaluation Report (SER) (reference: TVA letter from R. L. Gridley to James M. Taylor dated February 27, 1986).
"Out of function" is a term that has been used mostly at Sequoyah by DNE personnel and is generally unknown to personnel outside DNE (e.g.,
operations, Modifications, Maintenance). For this reason, DNE personnel do not require specific training regarding out-of-function information. A memorandum has been written from the Project Engineer to the Plant Manager and Modifications Manager which defines the term "out of function,"
describes DNE policy at Sequoyah concerning out-of-function information, and requestu that personnel using drawings at Sequoyah be trained in the-appropriate interpretation of out-of-function information.
Although it is project policy at Sequoyah to delete most out-of-function information on future drawing revisions, certain drawings are greatly enhanced by out-of-function information. On a case-by-case basis, some out-of-function information may be left on drawings when operations personnel (or other users) indicate that it significantly improves understanding of the drawing.
In these cases, the out-of-function information will accurately reflect the plant configuration to the extent practical. Although all plant drawings are included in this policy, priority may be given to certain drawings (e.g., primary or critical) because of their importance to operations.
Observation No. 2.3 - Status of (NSSS) Vendor Proprietary Information (86-45)
Response
NRC noted that TVA has committed to replace all NSSS vendor proprietary document commitments / requirements (C/Rs) with nonproprietary references j
during the phase of this program following unit 2 restart. This effort is 1
complete and is documented in Westinghouse Electric Corporation letter TVA-86-609.
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Deficiency D3.1 Exhauster Installation (86-27)
Response
TVA's response was provided to NRC by letter from R. L. Gridley to J. Nelson Grace dated July 28, 1986.
This item is open as a confirmatory 4
item.
Deficiency D3.2 USQD Requirement (86-27)
Response
TVA's response was provided to MRC by letter referenced above under D3.1-1.
This item is open as a confirmatory item.
Deficiency D3.2 Piping Flow Diagram (86-27)
Response
This deficiency identified a piping flow diagram with a missing class break indication. As a result of this deficiency, SCRMEB8614R1 was j
written to identify the problem, and ECN L6784 was issued to correct the
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deficiencies. Under this ECN, all flow diagrams and associated piping j
drawings were reviewed for missing and unclear class breaks and revised as required. All required changes to the drawings have been completed. This item is open as a confirmatory item.
Deficiency D3.3 Pipe Support Friction Design (86-27)
Response
l TVA's initial response was provided to NRC by letter from R. L. Cridley to 3
J. Nelson Grace dated July 28, 1986.
A schedule change was provided by j
letter from R. L. Gridley to J. Nelson Grace dated December 31,.1986.
Currently this issue remains open.
Section V of the July 28, 1986 submittal should be amended as follows. This revision provides additional information on our program scope and method of evaluation.
OTHER RELEVANT INFORMATION OR COMMENTS The following areas represent the major systems which are subjected to thermal loads and movements:
- 2. Essential Raw Cooling Water
- 3. Chemical and Volume Control
- 4. Component Cooling Water
- 5. Mainstream and Feedwater
- 6. Reactor Heat Removal
- 7. Fire Protection
- 8. Spent Fuel Cooling Water
- 9. Raw Cooling Water
- 10. Extraction Steam j
- 11. Safety Injection
- 13. Makeup and Purification l
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A minimum sample of 60 randomly-selected supports biased to major systems (as listed above) will be evaluated to represent TVA rigorously analyzed seismic class pipe supports.
The supports selected will be evaluated in two years as follows:
1.
Field investigations will be performed by site design representatives to visually inspect the support's physical condition and identify if there is any physical damage or visible marks because of thermal stress.
2.
An evaluation of the supports will be performed which includes the forces caused by friction to ensure that the weld stresses,.
plate bending stresses, and anchorage pullout loads are within -
the design basis.
.l The above information, in conjunction with the information provided from Watts Bar Nuclear Plant (WBN) friction investigation, will be used to substantiate that Sequoyah pipe supports are adequate as designed.
Additionally, our schedule to completion, defined in section III, should read as follows: " Evaluation will start at the beginning of i
unit 2 fuel cycle 4 and will be completed by the end of unit 2 fuel cycle 4."
This schedule is consistent with a similar alternate analysis commitment documented in TVA's response to NRC Safety Evaluation Report (reference:
Interim Acceptance Criteria For Small Bore Piping Safety Evaluation Report, S. A. White to B. J.
Youngblood, January 28, 1987).
Observation No. 3.4 - Pipe Support Design Criteria (86-45)
Response
l NRC noted on Observation 3.4 that technical discrepancies exist between j
the WBN piping support design criteria (WB-DC-40.31.9) previously used at Sequoyah and the Sequoyah piping support design criteria (SQN-DC-V-24.1) issued June 23, 1986. The two technical issues not addressed by the Sequoyah criteria are:
1.
Consideration of pipe stress induced by the use of stiff pipe clamps.
2.
Prohibition'on the use of pipe sleeves as axial pipe restraints unless specifically identified in the sleeve design (i.e., use of l
lugs attached to the sleeve).
l SQN-DC-V-24.1 applies to the design of all Category I and I(L) piping 1
supports and supplemental steel for that piping located in Sequoyah Category I structures. The omission of specific design requirements-presents the possibility that components of piping supports and supplemental steel logically covered by the missing criteria would lack specific design bases.
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- However, the omission of the items identified by Observation 3.4 from SQN-DC-V-24.1 is appropriate and does not diminish the technical adequacy of the design criteria as demonstrated below:
1.
Pipe stress induced by the use of stiff pipe clamps:
Stiff pipe clamps are specifically addressed in section 2.2 of General Construction Specification G-43, " Support and Installation of Piping Systems in Category I Structures," which is referenced by SQN-DC-V-24.1.
Therefore, the inclusion of criteria related to stiff pipe clamps in SQN-DC-V-24.1 is unnecessary.
2.
Pipe sleeves as axial restraints:
Section 8.3.1.2 of WB-DC-40.31 reads as follows:
Unless specifically called for on the support location isometric, I
pipes passing through vertical (floor) or horizontal (wall)
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sleeves, when axial support is required, will be axially supported independently of the sleeve.
If a support point is required adjacent to the sleeve, it will be obtained with supplementary steel or other suitable means which impose no load on the sleeve itself unless the sleeve has been designed for a pipe load.
Procedures and design standards are in place which ensure proper consideration of axial pipe loads at wall and floor sleeves. If the piping analyst assumes such restralnt, the wall or floor penetration is given a support mark number, and the support load tables reflect the axial load resulting from the analysis. The support designer then provides an adequate method of transferring loads to the floor or wall.
This is generally accomplished independently of the pipe sleeve. However, Mechanical Design Standard (DS)-MS.2.6, " Sleeve-Sizing, Sealing and Anchoring,"
prov. ides specific design guidance if use of the sleeve to transfer axial pipe loads is desired. The concerns identified do not constitute conditions adverse to quality; therefore, no action (s) to correct are necessary.
Observation No. 4.2 - Definition of Reactor Protection System and Neutron Monitoring System Scope (86-38)
Response
NRC noted that the boundary calculation, SQN-OSG7-048, did not explicihly define the boundaries of systems 92 and 99.
Revision 3 of the calculation included drawings that explicitly define the system boundaries and is available for NRC audit.
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' Observation No. 4.4 - Spray Shield Commitment / Requirement for Certain Hydrogen Igniters in Upper Compartment (86-45) l
Response
NRC noted that design criteria SQN-DC-V-26.1 needed to be revised to include the requirement for enlarged spray shields for the hydrogen igniters. This has been incorporated in the latest issued version of SQN-DC-V-26.1, Rev. 1, and is available for NRC audit.
Deficiency DS.3 Temporary Alterations Using Temporary Alteration Control Forms (TACFs) (86-27)
Response
i A Sequoyah engineering procedure, SQEP-44, was issued on January 16, 1987,
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for the handling of site nonengineered safety-related temporary changes for USQD evaluation. The procedure provides guidelines for DNE to perform a safety evaluation on all safety-related TACFs before their implementation.
Therefore, this item should be closed.
Unresolved Item US.3 Motor-operated Valve Thermal Overload Trip Setting (86-27)
Response
TVA has taken the following actions:
1.
A policy memorandum was issued on November 4, 1986, prompting the following additional actions.
2.
Design criteria SQN-DC-V-11.9 was issued January 15, 1987, which establishes the requirements for setting the thermal overload relays.
3.
Calculation SQN-APS-033 was completed, which verified the settings of l
all Class 1E motor-operated valve (MOV) trip settings.
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NEB has prepared an active valve list (AVL) which was issued under i
DNE calculation SQN-SQS2-0028, Rev. 1.
Although the effort on the AVL is considered as a part of the corrective action plan to address the concern on the MOV thermal overload protection system (ECTG reports 237'.1 and 237.4), the intent is to identify valves that are required to perform a mechanical motion in the event of a design basis event (in order to shut down the plant, raintain it in a safe shutdown condition, or mitigate the consequences).
This effort does not include the determination of the setpoint for the MOV thermal overload protection system which is not in the scope of the AVL.
EEB will use the AVL as a basis for addressing the concern of thermal overload protection for electric motors on MOVs to comply with l
Regulatory Guide 1.106, Rev. 1.
This effort is encompassed by the corrective action to disposition SCR SQNEEB86167.
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- Observation No. 5.1 - Walkdown Scope (86-38) 4
Response
TVA responded to this item by letter from R. L. Gridley to James M. Taylor dated February 3, 1987. Additional-information was provided to NRC during an NRC inspection on February 2-13, 1987, regarding specific instruments included in the walkdowns of dormant safety-related sense lines.
Observation No. 5.4 - Design Criteria (86-45)
Response
A response from TVA was provided to NRC by letter from R. L. Gridley to g
James M. Taylor dated February 3,1987.
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Observation No. S.5 - Connitment/ Requirement Inclusion in Design Criteria for the Auxiliary Power System (86-45)
Response
1.
Commitments / Requirements (C/Rs) 1072, 1073, 1077, and 1078 were originally determined to not belong in the Auxiliary Power System (APS) Design Criteria SQN-DC-V-11.4.1.
It was later determined that C/R number SQNEEBDRW1077 should be, and now is, addressed in design guide documents to be issued after unit 2 restart.
These documents will be issued by September 1, 1987, and will be used by design and procurement engineers as uesign input documents.
2.
C/R number SQNEEBDRW1092 was not included in the original data base sort of the C/R evaluation sheets but was reviewed later for possible inclusion into SQN-DC-V-11.4.1.
It was determined at that time that it did not belong in SQN-DC-V-11.4.1.
This C/R will be addressed in a design gulie document to be issued af ter unit 2 restart and by September 1, 1987.
3.
C/Rs SQNEEBDRW1017, 1018, and 1019 were not incluied in the original data base sort of the C/R evaluation sheet but were reviewed later for possible inclusion into SQN-DC-V-11-4.1.
It was determined that these C/Rs do not belong in SQN-DC-V-11.4.1, but that they do belong in the 125-V Vital Power System Design Criteria (SQN-DC-V-11.6).
SQN-DC-V-11.6 now includes those C/Rs.
Deficiency D6.1 Auxiliary Feedwater (AFW) Pump Discharge Pressure Switch Ratings (86-27)
Response
As stated in the original response (letter from R. L. Gridley to J. Nelson Grace dated July 28, 1986), the documentation to verify the technical differences in the AFW Pump Discharge Pressure Switch ratings listed on three instrument data sheets was not available for NRC review.
Design Calculations SQN-CA-D053, Rev. 1, and SQN-CA-D053, Rev. 2, provide l
the technical justification necessary and are available for NRC verification of setpoints listed in the data sheets.
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' Deficiency D6.1 Feedwater Bypass Control Valve Solenoid Replacement (86-27)
Response
Instead of a procedure change, as mentioned in the previous response, a new procedure, SQEP-44, was issued to ensure that required safety functions are identified and tracked.
ECN L6692 was issued to correct the specific problem. At the time of the audit, this information was not available for review. The following documents are available for NRC review:
SQEP-44, ECN L6692, FCR 4633 on ECN L6692, and SQEP-12 Checklist for ECN L6692.
Deficiency D6.1 AFW Pump Suction Precaure Switch Setpoint Calculations (86-27)
Response
The improperly referenced calculation (SQN-CA-D053-0-8CG-LCS-110882) was completed in March 1984 and superseded the previous issue of 1979. This calculation provides the technical justification necessary to verify the setpoint changes made in the 1984 ECNs.
Deficiency D6.3 Specification of Hydrostatic Tes' to Demonstrate Instrument Pressure Boundary Integrity Af ter Seismic Qualification Testing (86-27)
Response
TVA provided a response to this item by letter from R. L. Cridley to J. Nelson Grace dated January 30, 1987.
Observation No. 6.2 - Neutron Monitoring Detector Qualification Basis (86-38)
Response
Discussions are taking place between Westinghouse Electric Corporation and 4
TVA regarding the effort to address justification of responses to questions on the Unreviewed Safety Question Determination (USQD) for this issue. Revision of the USQD has not yet been finalized.
Regarding the responses prepared by the Plant Operations Review Staff (PORS) for SCR SQWNEB8609, Revision 0, which according to the NRC team, lacked an adequate justification for the evaluation, the Engineering Report for this SCR was revised to address the concern.
Observation No. 6.3 - Consistency in AFW Turbine Controls Walkdown Scope (86-38)
Response
TVA provided a response to this item by letter from R. L. Gridley to James M. Taylor dated February 3,1987.
(Refer to response to Observation 5.1.)
- Observation No. 6.5 - Replacement Part and Equipment Qualification (86-45)
Response
Design Input Memorandum (DIM) SQN-DC-V-32.0-001 was issued on January 29, 1987, and includes clarifications that replacement parts comply with applicable IEEE Standard 323 Revisions for environmental-qualification.
Observation No. 6.7 - Oil-Free Compressed Air Requirement (86-45)
Response
DIM SQN-DC-V-32.0-001 was issued on January 29, 1987, and includes a definition of " oil free" to be "less than 1 part per million (ppm) by weight."
Deficiency D4.3 Evaluation of Structures for Reinforcing Bar Cuts (86-27)
Response
TVA's initial response was provided to NRC by letter from R. L. Gridley to J. Nelson Grace dated July 28, 1986.
A revised response was provided to NRC by letter from R. L. Gridley to J. Nelson Grace dated December 31, 1986. TVA has completed an evaluation of structures for reinforcing bar cuts in response to Sequoyah Employee Concerns Task Group element 215.02.
This evaluation is documented in the following CEB calculations:
SUBJECT CALC ID NO.
Aux Bldg Slab El 714 SCG-IS123 Aux Bldg Slab El 734 SCG-1S123 Aux Bldg Slab,El 749 SCG-IS125 Struct Wall Sample - U Line SCG-IS128 Aux Bldg Shield Walls SCG-1S129 Reactor Bldg Shield Wall U2 SCG-1S130 SQN/WBN Dwg Comparison SCG-1S131 Rebar Cut Documentation -
SCG-1S132A Reactor Bldgs Ul&2 SCG-1S132B (3 Volumes)
SCG-1S132C Crane Wall Rebar Cuts SCG-IS133 i
. Deficiency D4.3 Steam Generator Access Platform Design (86-27)
Response
TVA's initial response was provided to NRC by letter from R. L. Gridley to J. Nelson Grace dated July 28, 1986. A revised response was provided by letter from R. L. Gridley to J. Nelson Grace dated December'31, 1986.
Observation No. 7.2 - Commitments / Requirements Related to Drilled-in Anchors (86-45)
Response
The NRC review identified a concern during their review of CEB compliance to section 5.6.5 of SQEP-12. SQEP-29 requires that an assessment be made and documented of each C/R against the existing General Design Criteria (GDC) to establish the need to revise an existing GDC or generate a new GDC for Sequoyah restart. NRC's concern was that the TVA review excluded C/Rs concerning drilled-in anchors with the justification that such inclusion was not required for restart.
The specific C/Rs addressed concern drilled-in anchors.
These items are used extensively at Sequoyah and could involve all safety systems. -
The TVA position is that no new or revised GDCs are needed for unit 2 restart because of uncaptured C/Rs, including those mentioned. The justification for that position related to drilled-in anchors is that all C/Rs concerning drilled-in anchors have been captured as follows:
C/R ID Number Design Input Document SQNCEB-CG1122:
Revised DC-Cl.7.1 SQNCEB-CG1123:
Revised DC-C1.7.1 SQNCEB-CG1124:
Issued Specification Revision Notice (SRN) for G-32 &
,, Quality Information Request (QIR) for DS-C.l.7.1 SQNCEB-CG1125:
Issued SRN for G-32 & QIR for DS-C.l.7.1 SQNCEB-CG1137:
Issued SRN for G-32 & QIR for DS-C.l.7.1 SQNCEB-CG1140:
Issued SRN for G-32 & QIR for DS-C.l.7.1 SQNCEB-CG1166:
Commitment met by G-32 SQNCEB-CG1167:
Commitment met by DS-C.l.7.1 SQNCEB-CG1168:
Commitment met by G-32 SQNCEB-CGil81:
Commitment met by response to NRC information request SQNCEB-CGil82:
Commitment met by G-66 The concern identified does not constitute conditions adverse to quality; therefore, no action (s) to correct is necessary.
The project has issued General Civil Design Criteria SQN-DC-V-1.0 to adequately document the capture of all restart C/Rs.
The purpose of SQN-DC-V-1.0 is to establish the requirements for use in the design of all civil features of Sequoyah.
The listed documents comprise the design bases for the plant.
In the criteria's Appendix A is a listing of all current Civil Design Criteria for the design of various features of Sequoyah. These design criteria include any DIMS in force at the time of use.
The criteria's Appendix B lists the references to specific documents or sections / parts of documents in which the C/Rs referencing "All Safety Systems" are captured. These references are to be considered design input documents.
a EA verified that the C/Rs for drilled-in anchors were properly incorporated in the GDC for capture under G-32, G-66, or-DS-C-1.7.1.
I Also, the generation of General Civil Design Criteria, SQN-DC-V-1.0, is sufficient to adequately document the capture of all restart C/Rs
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associated with drilled-in anchors.
t Observation No. 7.3 - Revision of Design Criteria for Restart (86-45)
Response
The NRC review identified a concern during their review of-CEB compliance to section 5.6.5 of SQEP-12.
SQEP-29 requires that an assessment be made j
and documented of each C/R against the existing GDC to establish the need to revise an existing GDC or generate a new GDC for Sequoyah restart. The TVA evaluation of "All Safety-Related Systems" analysis report (999 sort) cited that GDC SQN-DC-V.l.1.1 must be revised to incorporate C/R i
SQWCEB-CG-1170.
The copy of the GDC provided to NRC had not been revised to do so, and the project advised that none of the existing CEB GDCs needed to be revised as a result of the 999 sort review.
The specific C/Rs addressed a concern regarding through-wall bolts for masonry walls. These items are used extensively at Sequoyah and could involve all safety systems.
The TVA position is that no new or revised GDCs are needed for unit 2 restart because of uncaptured C/Rs, including those mentioned. The justification is that C/R SQNCEB-CG 1170, which addresses attachments to masonry walls, has been revised. Wording was eliminated which was construed to require that only through-wall bolts be allowed in masonry walls. A requirement to use only through-wall masonry anchors does not j
exist. The concern identified does not constitute a condition adverse to quality; therefore, no action (s) to correct is necessary.
l The project has issued General Civil Design Criteria SQN-DC-V-1.0 to j
adequately document the capture -f all restart C/Rs. The purpose of j
SQN-DC-V-1.0 is to establish th(
quirements for use in the design of all civil structural' features of Sequoyah.
The listed documents comprise the design bases for the plant.
In the criteria's Appendix A is a listing of all current Civil Design criteria for the design of various features of j
Sequoyah. These design criteria include any DIMS in force at the time'of use. The criteria's Appendix B lists the references to specific documents i
or sections / parts of documents in which the C/Rs referencing "All Safety Systems" are captured. These references are to be considered design input i
documents.
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ENCLOSURE 3 LIST OF CONNITMENTS IN ENCLOSURES 2 AND 3 FOR SEQUOYAH NUCLEAR PLANT 1.
Submit FSAR revision for reduced CCS heat loads in 1987 update.
2.
Submit FSAR revision for ERCW flow rates and heat loads in the 1988 update.
3.
Complete new CCDs for WDS drawings (47W560-series) by October 1,1987.
4.
Revise the SQN Rigorous Analysis Handbook by April 30, 1987, to provide guidance to analysts regarding evaluation of slip joints.
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5.
Submit FSAR revision to address the use of copper tubing for field l
routed instrument lines in the 1988 update.
6.
Notify piping analysts by April 30, 1987, to be quality conscious in preparing isometric drawings.
7.
Conduct section training and revise Rigorous Analysis Handbook checklist by April 30, 1987, to address coordination of pipe rupture evaluations.
8.
Revise AI-19, Part IV, by April 30, 1987, to direct instrument engineers to reference any source documents used to develop accuracies, setpoints, etc., used in workplans.
9.
Perform a test on the CCS Surge Tank by May 16, 1987, to determine the extent of any leakage through the baffle plate.
10.
Perfoam a functional test by March 31, 1988, on the CCS Pump Discharge Pressure Switch to verify that the alarm operates at setpoint and resets on decreasing pressure.
11.
Issue'a report by April 30, 1987, that documents the results of the receview of the electrical SQEP-12 checklists for consideration of imposed voltages.
12.
Agastat timer relay model 7012-PBL, designated in pump motor start-circuits as CCS-A, AFW-A, ERCW-A, and CPAK-1A, will be calibrated before restart of unit 2.
s 13.
Agastat timer relay model 7012-PBL, designated in pump motor start circuits as CCS-A, AFW-A, ERCW-A, and CPAK-1A, will be incorporated into SQE-8 by December 31, 1987.
14.
Disposition PIR SQNCEB8639 before restart of Sequoyah unit 2.
15.
Issue design guide documents by September 1, 1987, incorporating C/R numbers SQNEEBDRW1077 and SQNEEBDRW1092.
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