IR 05000327/1987006
| ML20206B856 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/03/1987 |
| From: | Architezel R, Imbro E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20206B848 | List: |
| References | |
| 50-327-87-06, 50-327-87-6, 50-328-87-06, 50-328-87-6, NUDOCS 8704100066 | |
| Download: ML20206B856 (22) | |
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U.S. NUCLEAR REGULATORY COMISSION l
0FFICE OF INSPECTION AND ENFORCEMENT
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i Division of Quality Assurance, Vendor, and i
Technical Training Center Programs
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Report Nos.:
50-327/87-06, 50-328/87-06 j'
Docket Nos.:
50-327; 50-328
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Licensee:
Tennessee Valley Authority
6N, 38A Lookout Place j
1101 Market Street
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Chattanooga, TN 37402-2801 Facility Name:
Sequoyah Nuclear Plant, Units 1 and 2 Inspection At:
Knoxville, TN
Inspection Conducted:
February 2-13, 1987
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Inspection Team Members:
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i; Team Leader:
R. E. Architzel, Senior Inspection Specialist, IE
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Mechanical Systems:
R. W. Parkhill, Inspection Specialist IE l
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F.Mollerus, Consultant,MollerusEngIneering,Inc.
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J. Nevshemal, Consultant *
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i Mechanical Components:
A. V. duBouchet, Consultant
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Civil / Structural:
A. Unsal, Consultant, Harstead Engineering I
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Electrical Power:
S. V. Athavale, Inspection Specialist, IE
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j Instrumentation and L. Stanley, Consultant, Zytor Inc.*
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i Control:
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H. J. Miller, Deputy Director, DQAVT, OIE**
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i E. V. Imbro, Section Chief, QA Branch, OIE**
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Ralph E. Architzel F '
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Team Leader
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[ Y bA,, /& 4 - 0 -) /
Eugene V. Imbro Date
i Section Chief t
j Quality Assurance Branch
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- Part time
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- Attended exit meeting
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I'DH ADOCK 05000327 a
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SEQUOYAH NUCLEAR POWER PLANT
DESIGN CALCULATION REVIEW PROGRAM INSPECTION REPORT 50-327/87-06 & 50-328/87-06 FEBRUARY 2-13, 1987
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1.
INTRODUCTION AND BACKGROUND The design calculation review program was developed by the Division of Nuclear Engineering (DNE) because past audit findings have shown that the design basis l
for TVA's nuclear power plants have not been adequately documented by support-ing calculations or such calculations, if performed, are no longer retrievable.
This calculation review program was identified by TVA as augmenting the Design Baseline and Verification Program (DBVP) by providing defallec technical reviews of calculations that supported engineered changes to the plant design l
I since the initial license to operate was granted.
The TVA calculation review l
program is, however, broader in scope than the DBVP and also includes a review of initial design calculations.
The NRC conducted an inspection of the calcu-lation review program because it was determined in a previous inspection (50-327/86-55,50-328/86-55) that the DBVP did not perform a detailed technical review of the calculations that supported the engineered changes made to the plant since receipt of the operating license. The DBVP review of calculations only included a check to verify that appropriate calculations existed, and that the proper technical attributes have been considered.
- The Sequoyah Nuclear Plant (SQN) Design Baseline and Verification Program will be used by TVA to provide the required level of confidence that the modifications to selected plant systems, implemented since receipt of the operating license, have not resulted in any violation of the plant's licensing basis.
The DVBP is described in the " Program Plan for the Engineering Assurance Independent Oversight Review for the Sequoyah Nuclear Plant Design Baseline and Verification Program " dated May 9, 1986 and forwarded to the NRC as an enclosure to Mr. R. L.
g Gridley s letter dated June 27, 1986.
The design calculation review program is described in an enclosure to TVA letter from Mr. R. L. Gridley dated January 20, 1987.
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2.
PURPOSE This report summarizes the results of the NRC inspection conducted to assess the adequacy of TVA's design calculation review program and the extent to which it augments the Design Baseline and Verification Program to support restart of
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Sequoyah's Nuclear Plant.
The purpose of this inspection was to:
(1) Assess the adequacy of the design calculation review program as it augments the DBVP.
(2) Assess TVA's identification of essential calculations, i.e., calculations required to support the plant design.
(3) Assess TVA's calculation review scope to ensure it includes a representative sample of the design calculations performed by the specific disciplines l
being evaluated.
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(4) Assess Engineering Assurance's (EA) technical audit of the design calcula-
tion review performed by the Department of Nuclear Engineering.
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l (5) Assess on a sampling basis the technical adequacy of the calculations l
supporting Sequoyah design.
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3.
INSPECTION ACTIVITIES
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To implement the inspection's purpose, the following activities were generally
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performed by all team members as related to their specific design discipline (i.e. MEB - Mechanical Engineering Branch; NEB - Nuclear Engineering Branch;
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CEB - Civil Engineering Branch; EEB - Electrical Engineering Branch) within the Department of Nuclear Engineering (DNE).
(1) Reviewed procedures associated with design calculation review program.
(2) Reviewed the list of TVA identified essential calculations for comprehen-siveness.
(3) Reviewed the scope of the design calculation review program for its f
interface with the DBVP and whether the calculations reviewed were repre-
sentative of the specific DNE discipline.
(4) Performed independent technical reviews of a sample of design calculations
within the DNE scope of review and outside DNE scope of review.
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(5) Performed independent technical reviews of a sample of calculations reviewed by EA.
4.
RESULTS OF NRC INSPECTION The following paragraphs characterize the team's observations and conclusions as they relate to the inspection activities identified in Section 3 with the
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detailed description of the observations in each discipline provided in Attachment A.
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4.1 Review of Procedures
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During review of the procedures associated with the calculation review efforts several concerns were identified relating to proposed changes to the procedures.
The team was informed that Nuclear Engineering Procedure (NEP) 9.1, Corrective Action, was to be revised and reissued by March 30, 1987.
The revised pro-cedure was to be consonant with a recent revision (1-4-87) to Nuclear Quality Assurance Manual (NQAM) Generic Procedure Part I, Section 2.16, Corrective Action.
The team was concerned with several aspects of the proposed revision to NEP 9.1.
One concern related to the new requirement that a Condition Adverse to Quality (CAQ) was needed to constitute a substantiated condition as opposed to identification of a potential safety issue (General Observation No. 1).
Another concern was identified relating to provisions requiring confirmation of
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a system's inability to meet specific Technical Specification functions, versus l
design functions, when assessing impact on operability (General Observation No.
2). Additionally, under the revised procedure TVA indicated that EA will now review significant CAQs.
Previously, only sampling reviews of CAQs were
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performed by EA.
As a result of this change, and because all generic CAQs are classified as significant, EA will review all CAQs that have the Potential
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I Generic Condition Evaluations (PGCE) checked "yes" without performing a similar
review of those checked "no."
However, EA will continue to review on a sampling
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basis those calculations that have the PGCE checked "no." (Mechanical
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Observation No. 5).
For both the Nuclear Engineering Branch and the Mechanical Engineeting Branch all Significant Condition Reports (SCR) and Problem Identification Reports
(PIR) generated as part of the design calculation review program had no PGCE
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performed due an invalid policy statement (Mechanical Observation No. 4).
4.2 Review of Listing of Essential Calculations MEB l
The listing of MEB essential calculations was developed in accordance with policy memorandum B44860625002.
The following systems have the mechanical hydraulic calculations performed by Westinghouse and TVA is only responsible for the piping routing, stress analysis and support:
chemical volume and control system; residual heat removal system; upper head injection system; and safety injection system.
For all other systems excluding primary coolant system, MEB is responsible for the hydraulic calculations.
In general, TVA has responsibility to layout and arrange NSSS vendor designed system equipment and piping while the NSSS vendor has the responsibility for all mechanical l
specifications and calculations.
For this reason, MEB calculations for systems in the NSSS scope are limited primarily to special evaluations.
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MEB has completed its compilation of essential calculations, has identified several essential calculations that were missing and has stated in MEB memorandum B44861217017 that these essential missing calculations would be regenerated prior to *estart.
In its review of the essential calculation listing the team found some minor inconsistencies such as duplication of calculations, calculations which did not state an objective, or the calculation listed was obsolete and superseded by another calculation (see MEB observation No. 7). The team considers that the process and guidelines used to develop the list of essential MEB calculations
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an adequate approach to establishing a complete list of essential calculations.
However, the team notes and agrees with EA's finding that no documentation is available to substantiate that the list of essential calculations contains I
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those calculations necessary to support licensing commitments and operating conditions permitted by Technical Specifications.
NEB
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The Team's inspection of the list of NEB essential calculations found the list tu be complete.
l EEB The team reviewed the list of EEB essential calculations for restart of Unit 2.
This list was the result of combined efforts of TVA and Sargent Lundy.
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defined a minimum set of essential calculations for restart based on system safety classification.
i This list was later reviewed and evaluated by Sargent
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& Lundy for restart of the plant.
With the incorporation of the enhancements recommended by Sargent & Lundy, the team concluded that the list of essential
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calculations in EEB was adequate.
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CEB
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Civil Engineering Branch (CEB) on December 2, 1986 issued policy memorandum
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PM 86-02, Civil Discipline Policy for Design Calculations, to provide requirements
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for classification and indexing of calculations.
This memorandum included
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the Civil Engineering Master Calculation List which defines the calculations j
to support the scope of work performed in CEB.
This list includes a classifi-cation of each calculation to be either essential or desirable.
The memorandum
required the development of project specific calculation lists, since the l
master calculation list was generic to all TVA nuclear plants.
l In accordance with CEB policy memorandum PM 86-02, the Sequoyah project developed an action plan for the civil discipline calculations.
This plan is i
i contained in TVA memorandum, 825870119307}onsintothreecategories:
from C. N. Johnson to R. O. Barrett.
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The plan separated the essential calculat
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All calculations necessary to support ongoing restart ECNs.
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Calculations necessary to substantiate the design within the scope of the l
Sequoyah Design Baseline and Verification Program which includes missing and improperly performed calculations, l
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The remainder of essential calculations in the Sequoyah civil engineering master calculation list which are not addressed by items A and B.
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Prior to the NRC inspection, TVA had already identified the calculations for
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Categories A and B.
During the inspection, the team noted that TVA was still working to compile the list of calculations for Category C.
Instruction
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SQN-C-QIO2,B25870123332, was issued to assure that all essential civil
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discipline calculations in Category C would be retrievable.
In addition, CEB
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has issued supplementary instructions for review of retrievability of calcula-i tions to support requirements contained in design criteria, design standards 2..
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and guides, and construction specifications.
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The team has reviewed the above-mentioned documents and believes that TVA's program to retrieve all essential CEB calculations for the Sequoyah project is
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adequate and, if implemented correctly, will identify all essential calcula-i
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tions as described in the CEB policy memorandum PM 86-02.
The team reviewed
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program implementation in that the list of rigorous analyses problems tabulated i
under Category B was verified to include all of the rigorous analyses for piping systems within the scope of the DBVP.
The team also identified a missing " essential" calculation for the seismic Category 1 essential raw cooling water (ERCW) buried pipe which runs between
the ERCW pumphouse and the auxiliary building not identified by CEB.
CEB was unable to retrieve this calculation and had not tabulated this missing calcula-
tion under Category C of the calculation index, to insure the regeneration of
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the calculation prior to restart of SQN unit 2 (Obs. No. CEB-8),
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4.3 Review of Scope of Design Calculation Review Program
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MEB As delineated in MEB memorandum B44861217017, 55 out of 402 essential calculations were initially selected for review with only one out of 15 DBVP calculations i
selected for review.
These 55 calculations were randomly selected from the following systems:
System 30 - Ventilation, (17 cales)
System 31 - Cooling / Heating, (8 cales)
System 32 - Essential Control Air (1 calc)
System 67 - Essential Raw Water Cooling (16 cales)
i System 70 - Component Cooling Water (12 cales)
System 82 - Standby Diesel Generator (1 calc)
After reviewing this scope of review, the NRC team suggested the enhancements i
identified in MEB Observation No. 1 (e.g. the sample should include all MEB sections, calculations performed later than January 1986 and overlap with DBVP)
MEB memorandum B44870211026 to the Chief Mechanical Engineer recommended that the review scope be expanded.
With these enhancements the sample size will increase to 92 calculations out of 402.
Seven of the 92 calculations reviewed
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are within the scope of the DBVP and represent approximately one half, 7 out of 15, of the DBVP calculations.
This adequately reflected the team's recommenda-tions.
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NEB
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The Nuclear Engineering Branch has six sections:
Thermal Hydraulics - Performs calculations for missiles, pipe rupture, con-tainment temperature and pressure, etc.
Radiation Protection - Performs calculations for shielding, radiation source terms, offsite dose, mild and harsh radiation environment, ALARA, radiation monitor design, etc.
Environmental Control - Performs calculations for 10 CFR 50.49 environment, HVAC FMEA's, safety-related HVAC setpoints, etc.
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Nuclear Waste - Performs calculations for radiation waste system and component design, H2 generation and accumulation, radiation waste safety evaluation, etc.
- i Safety Analysis - Perhrms calculations for safety-related components and systems, fluoding offects, safety related requirements, and regulatory commit-ments, etc.
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Safety Analysis (located at site) - Performs calculations to support equipment qualification, etc.
l These six Sections have performed approximately 350 calculations which have been classified as essential.
The team found that approximately 180 of these calcula-tions had been selected for review under the NEB calculation review program, This includes 100% of the calculations performed by the Knoxville Safety Analysis i
Section, 110 calculations and 70 other calculations that were randomly selected
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from a list of the remaining 240 essential calculations.
The randomly selected calculations covered four of the other five (5) sections.
The site safety
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analysis group was not covered.
The results to date of the NE8 calculation review program identified ten calculations that were " Unacceptable" which is defined as being not supportive of the current plant design.
This indicates a need for a follow-on NE8 effort.
The team recommended that the follow-on effort should be biased towards
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those NEB sections with the largest population of " Unacceptable" calculations.
As part of the inspection of the NES program results, the team found that the program did not distinguish between calculations within the scope of the D8VP and those supporting the initial design.
Therefore, the team was unable to l
draw a conclusion relative to the technical adequacy of calculations supporting engineered changes within the D8VP scope.
The team recommended that the
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follow-on effort be biased towards calculations within the scope of the D8vP.
EEB
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Most of the design calculations required for restart of Unit 2 were done
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recently by TVA using up-to-date methodology and also using computers.
These calculations were checked and reviewed by the in-house staff using guidelines
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similar to guidelines used by Sargent & Lundy.
Many calculations were reviewed I
by Sargent & Lundy.
The team reviewed calculation checklists for two calcula-tions (Aux, power system and control power system) and noted that these check-lists were adequate in that a comprehensive listing of technical attributes was provided.
Since most if not all design calculations have been redone, the team
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concluded that the activities of EEB was fully covered by the program.
CEB Due to the large numbr af essential calculations (>10,000) that CEB has for the Sequoyah project, a different approach from the other disciplines was adopted to determine technical adequacy of the calculations.
Instead of reviewing individual calculations for technical adequacy, CEB decided to review previous verification programs to determine the design areas which had not been reviewed and the areas in which concerns had been raised.
In the civil / structural
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discipline the following areas of information were considered in this review:
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Conditions adverse to quality B.
Employee concern program C.
Employee interviews D.
External reviews E.
Internal reviews This information was used by TVA as indication of potential design problems.
IVA's view is that where these information sources do not give indication of problems, detailed review of calculations for technical adequacy is not The above information was reviewed by TVA relating to the following warranted.
design and analysis features:
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Foundations /geotechnical B.
Concrete structures C.
Steel structures D.
Embedded and surface mounted plates E.
Cable tray supports
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Pipe restraints G.
Seismic analysis of structures
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Analysis of steel containment vessel l
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Site drainage and probable maximum flood calculations J.
Conduit and HVAC duct supports K.
Pipe rupture
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The results of this review are contained in the attachment to CEB memorandum B41870130013 which makes recommendations to DNE management for further actions that should be taken to determine technical adequacy in various features listed above.
The CEB management has adopted some of these recommendations and has
started a program to review miscellaneous steel design calculations, and other l
areas where warranted by the results of the review, as shown in TVA document B41870204002.
The team noted that the specific CEB program description needs to be formally submitted to the NRC in order that NRC can assess the CEB calculation effort scope of review.
One particular concern is the scope and depth of past reviews (beyond the standard quality or design verification calculation check) that TVA is relying upon to justify not examining civil engineering calculations in the current review program.
NRC reviewed two calculations in an area which TVA was not planning to review based on the CEB review of previous verification programs.
In one of these calculations, the NRC found an unjustified assumption for concrete compressive strength of 6300 pst vs. 4000 psi as stated in the FSAR and that allowable stress in concrete and steel had been exceeded by 19% and 16%,
respectively, also without justification (OBS CEB-1).
This observation has
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given NRC cause to question TVA's methodology for determining which areas in CEB require further review.
TVA was requested to provide justification of their rationale for excluding certain areas of CEB design from being reviewed on a sampling basis for technical adequacy.
In this regard TVA was also requested to provide descriptions of the scope and depth of previously con-ducted internal and external reviews that they are using as a basis for not performing current detailed technical reviews of calculations similar to what has been done in the other three technical branches.
4.4 Technical Review Summary
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The specific results of the NRC team's inspection of the DNE design calculation review program are provided in Attachment A.
These observations originated from the team's overview of calculations reviewed by DNE and EA, as well as independent review of calculations outside of the ONE and EA review scope.
Since the calculation review program complied a listing of essential calculations from which a sample was reviewed, both pre-and post-operating Ilcense calculations were included in the review.
As a result of these design calculation reviews, many of the observations identified by the inspection team as well as by TVA involve initial design activities.
While the focus of TVA's effort in the DBVP has been on the design change process after the operating license was issued, TVA must, in some fashion, address the generic implications of these observations on the initial design.
In this connection, it is recognized that, in some areas, TVA has already indicated that it will conduct a complete review of
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l calculations similar to those that have been found to be unacceptable.
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One issue identified by the team is of particular concern as it raises questions about the ef fectiveness of TVA's process for assuring that engineering changes
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are incorporated into plant operating procedures and surveillance program, i
Specifically, the operations staff was not notified promptly of necessary
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Class 1E battery operating restrictions after it was determined by DNE that
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the installed battery capacity was less than that required by new design
calculations (Observation No. EEB-1).
Since the DBVP and calculation review programs are resulting in numerous changes to the design basis, it.is important that TVA take steps to assure that the ramifications of these changes l
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on plant operations be identified and reflected in operating procedures.
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ATTACHMENT A - OBSERVATIONS
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GENERAL OBSERVATIONS Observation No. GEN-1 - Substantiated Condition for a CAQ
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Nuclear Engineering Procedure (NEP) 9.1, Corrective Actions, revised 7-1-86, defines the controlled system within the DNE to document, evaluate and resolve
conditions adverse to quality (CAQs) relating to all engineering work within
DNE.
The NRC had previously reviewed implementation of the corrective action
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program within DNE/0E carried out under the provisions of OEP-17, which pre-
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ceded NEP 9.1.
i NEP-9.1 requires the documentation of "... any condition which renders an item
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unacceptable to perform its required function or creates uncertainty concerning
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its ability to meet design requirements..." The procedure also requires a test for significance; those items determined to be significant are documented via significant condition reports (SCRs), all other CAQs are documented via r
j problem identification reports (PIRs).
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The team was informed that NEP-9.1 was in the process of revision to agree j
with the recently revised corporate QA procedure for corrective actions, NQAM i
Part I, Section 2.16, revision dated 1-4-87.
Although the draft NEP-9.1 was not available for review, TVA informed the team that the draft was the same as the NQAM relative to CAQs.
Under the new procedures, all CAQs are to be
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documented as conditions adverse to quality reports (CAQRs); SCRs and PIRs will
no longer be generated within DNE.
Section 4.2 of NQAM Part I, Section 2.6, l
defines CAQs as (emphasis added):
" Adverse conditions include nonconforming material, parts or components; failures, malfunctions; deficiencies; deviations, hardware problems i
involving noncompliance with licensing commmitments, specifications, I
or drawing requirements; abnormal occurrences; and nonhardware
problems such as failure to comply with the operating license,
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technical specifications, licensing commitments, procedures,
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instructions, or regulations.
Unsubstantiated conditions are not defined as CAQs."
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The team was concerned with the lack of inclusion of unsubstantiated conditions within the set of CAQs.
The licensee does not appear to have an alternative controlled system in place to identify and resolve conditions which create uncertainties regarding an item's ability to perform design functions.
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situations could be interpreted as unsubstantiated during identification and reviews.
The team was concerned that the net effect of the procedure change
would be to eliminate a set of CAQs which had previously been identified and resolved by SCRs and PIRs, without establishing an alternative method for
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identification, tracking and resolution.
Observation No. GEN 2 - CAQ Operability Determinations i
The team was informed that NEP-9.1, Corrective Actions, was in the process of j
revision to agree with the recently revised corporate QA procedure for correc-tive actions, NQAM Part I, Section 2.16, revision dated 1-4-87.
Although the
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draft NEP-9.1 was not available for review, TVA informed the team that the i
draft was the same as the NQAM relative to CAQs.
Under the new procedures, all
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CAQs are to be documented by condition adverse to quality reports (CAQRs); SCRs
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and PIRs will no longer be generated within DNE.
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The team noted that the new procedure requires the management reviewer of the organization initiating a CAQR to assess the condition for potential impact
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on operability (procedure section 5.2.2).
This operability assessment is to be performed in accordance with procedure Attachment 5, Guidelines for Potential Operability Determinations. This requires that the component's operability be determined by its ability to perform its safety related function required by the technical specification rather than its design related function.
The team is concerned that restricting the operability assessment to a confirmed tech-nical specification non-compliance inappropriately disregards design based operability requirements which are included in such documents as the FSAR and design criteria.
For example, on some nuclear power plants, there are no specific technical specification operability requirements for room coolers; however, removing these from service could compromise the environmental quali-fication of safety-related equipment.
MEB OBSERVATIONS Observation No. MEB-1 - MEB Design Calculation Scope of Review Based upon the description of the MEB calculation review sample size provided in memorandum B44870206043 and discussions with DNE personnel, the team recommended the following enhancements to ensure a representative review:
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A.
A 100% review of related essential calculations for all calculations determined by DNE calculation review program to be unacceptable.
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of the inspection two generic areas had been determined to be unacceptable, viz. HVAC heating /coolin condition calculations. g load determination calculations and off-design B.
Calculations produced by all four MEB sections would be sampled.
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Calculations dated later than January 1986 would be sampled since review of one calculation by NRC dated later than January 1986 indicated that even
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new calculations should not be assumed to be correct (see MEB observation No. 6).
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Calculations supporting the DBVP would be expanded to include seven of the fifteen calculations within the DBVP scope.
These enhancements were recommended to the MEB Chief Mechanical Engineer
from his staff memorandum B44870211026.
Observation No. MEB-2 - SI Pump Mini Flow Rate In reviewing the design interface responsibility between MEB and Westinghouse
the team noted that the design criteria for the safety injection (SI) system
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(805860805507) required the SI pump minimum recirculation flow (miniflow) to i
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be 40 GPM.
Whereas the miniflow orifice drawing (No. RMF-46368 Rev. 0) and TVA correspondence with Westinghouse documented the miniflow to be 30 GPM.
MEB acknowledged after verification with Westinghouse that the design criteria was in error and will be updated to reflect a mini flow rate of 30 GPM.
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Observation No. MEB-3 - Water Hammer
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The design interface between the mechanical systems (MEB) and the pipe stress analysis discipline (CEB) was tested by reviewing the containment spray opera-tional modes calculation (B25861223300).
The team noted that no mention was made of water hammer effects.
Later it was discovered that water hammer effects (i.e., forcing functions) are transmitted to CEB by NEB rather than via MEB.
Subsequently, CEB verified that no documentation was available which demonstrated that water hammer effects had been evaluated for the containment spray system which is in TVA's design responsibility.
After further evaluation, CEB performed an informal calculation which demonstrated that the effect of water hammer on the containment spray system is insignificant.
CEB should formally update their stress analysis to document that water hammer effects are insignificant and the basis for that determination.
While reviewing the broader issue of systems that had been evaluated for water hammer, the team questioned why the feedwater system water hammer analysis had been completed but not issued.
CEB needs to justify the reason for not issuing the feedwater water hammer analysis.
Observation No. MEB-4 - Potential Generic Condition Evaluation In reviewing the problem identification reports (PIRs) and significant condition reports (SCRs) generated as a result of the MEB calculation review program, the team noted that none required a potential generic condition evaluation (PGCE).
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MEB's justification for not requiring a PGCE was that all other TVA nuclear facilities were having similar calculational review programs.
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i instructions from the Chief Mechanical Engineer via memorandum B44860325014,
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stated that a PGCE for SCRs was not required for the identical reason.
Since the calculational review programs are not 100% comprehensive, but instead are based upon sampling, the aforementioned reason does not ensure that a specific technical issue raised on Sequoyah will also be addressed at the other TVA nuclear facilities.
Therefore, MEB needs to re-evaluate the need of a PGCE for all PIRs and SCRs generated to date as a part of the calculation review program.
This observation also applies to NEB.
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Observation No. MEB-5 - EA's Review of CAQs In discussing a pending change to TVA's procedure entitled " Corrective Action,"
NEP-9.1, it was stated that EA will review all CAQs that have potential generic condition evaluation (PGCE) checked "yes."
However, a similar review will not be performed for CAQs with PGCE checked "no."
This methodology biases the process to reduce the total member PGCEs generated, rather than verifying if the need for a PGCE has been accurately assessed.
For example, under this system, the MEB SCRs and PIPS that were inappropriately checked "no" for PGCE (See observation MEB-4), would not have had the opportunity to be corrected by EA.
It is recommended that EA be used to verify the adequacy of all CAQs for PGCE, not just CAQs that have been determined to require a PGCE.
Observation No. MEB-6 - Component Cooling Water System Design Pressure In reviewing the calculation which establishes the component cooling water system design pressure (B44861210008), it was noted that the design pressure was established by combining the pump head associated with a LOCA-SI flowrate A-3
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and the static head between the high and low points in the system.
Consideration-
of pump shutoff head and surge tank relief valve setpoint was omitted without justification.
If considered, the resulting maximum system pressure would exceed the design pressure by approximately 20 psi.
Therefore, MEB should assess the impact of the two aforementioned parameters on the design pressure of the component cooling water system as well as assess their impact on testing requirements for the associated piping and other components.
Observation No. MEB-7 - Identification of Controlling Calculations In reviewing the listing of MEB calculations, the team noted that some calculations addressed the same design attribute, others evaluated off-design conditions, some had been superseded in part by other calculations, and some did not clearly state the purpose of the calculations.
Because of these examples, it is suggested that MEB put in place a system that clearly identifies the controlling calculations for each design attribute.
Observation No. MEB-8 - Inconsistent Equipment Qualification Temperature Calculation NEB 811007235 was prepared to analyze a deficiency in the preoperation test results for the turbine driven auxiliary feedwater pump.(TDAFW) pump room ventilation system and to provide suggestions to reduce the temperature rise.
The discussion and methodology in the calculation uses a temperature rise to 125 F, which the calculation states is the equipment qualification temperature.
This is not consistent with the plant's environmental data sheet, 47E235, Sheet
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71, that specifies a design temperature of 104 F and a peak temperature of 110 F for the TDAWP pump location.
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Observation No. MEB-9 - Unverified Heat Load Input Calculation SWP 811202004 was identified by DBVP as a calculation supporting ECN 5415. The calculation was issued to determine reduced ERCW flow rates for ESF coolers when the ERCW is less than the maximum design temperature.
The calculation uses some heat loads that are less than the design or rated capacity of the coolers being evaluated.
These input heat loads are described in the calculation as calculated room loads.
However, these heat loads could
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not be verified by calculations, tests or other references.
The basis for these heat loads should be justified.
NEB OBSERVATIONS Observation No. NEB-1 - ECCS Pump NPSH The calculation for the minimum water level in the containment sump SQN-0SG7-008
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entitled, Determination of Minimum level in Containment Sump at Time of Switch Over to Recirculation Mode (B45-851218-235) resulted in a water level of 13.2 ft which is the top of the reactor vessel shield wall.
The calculation assumed that the following water inventories are introduced into the containment; RWST available volume, RCS, accumulators, and ice melt.
This set of assumptions
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may not be limiting because they result from a large break LOCA.
Since the purpose of the calculation is to determine minimum water level, the initiating event should be one which results in minimizing the water introduced into the containment.
In terms of minimizing water level, the limiting event might be a small break located high on the RCS which would reduce the ice melt contribution to water level.
The calculation also assumed that the reactor cavity volume A-4
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would not be filled during the event even though there are penetrations in the
shield wall.
The team feels that this assumption needs to be justified because it is non-conservative.
The water level determined by this calculation along with the sump fluid temperature design value commitment is used in the evalua-tion of the NPSH available.
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The team found in Section 6.3 of the FSAR that no credit was taken for sub-cooling of the sump fluid.
The team also found that NUREG-0011 Supp. 1 (Sequoyah SER) gave relief from this commitment and established a sump fluid temperature of 190 F with no credit for containment pressure over atmospheric.
The SER (dated February 1980) concluded that the 190 F temperature would result in about 2.8 ft of H,0 excess NPSH available over that required by the ECCS pumps.
The FSAR should be updated to reflect the allowance for subcooling specified in the SER.
The team found that a fibrous insulation (NUK0N) was added to the pressurizer power relief valve loop seal since OL.
Calculation NEB-841127-220 was performed to determine the effect of this insulation on the sump screen pressure drop if it were to become detached during a LOCA.
The effect on pressure drop was determined to be 6.6 ft of H 0 when distributed according to Reg. Guide 1.82.
This calculation post-dated the SER determination that resulted in 2.8 ft of H O
excess NPSH available assuming a 190 F sump fluid temperature.
Included in this calculation was an analysis of the NPSH required for the ECCS pumps.
Additionally, this analysis did not use the potential runout flow to determine the NPSH required for the pumps.
The potential runout flow is the intersection
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of the system resistance curve with the pump vendor certified head curve.
The team feels that the NPSH calculational effort has not addressed the limiting case because of:
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the non-conservative assumptions in the water level calculation, B.
the potential impact of the fibrous insulation on sump screen pressure drop, and C.
the failure to consider pump runout flow to determine the NPSH required for the ECCS pumps.
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m Observation No. NEB-2 - Wide Range Containment Pressure Transmitters TVA calculation SQN-NAL4-002 Rev. 6 (845870109235) stated that wide range containment pressure transmitters PT-30-310 and -311 have a range of -5 to
+60 psig with a 110.98 psi instrument error due to accident environment and seismic effects.
This calculation stated that these instruments were required by NUREG-0737 Section II.F.1, but that they were not required for mitigation of design basis events.
In addition, it was stated that these transmitters and their display indicators were not needed for operator action, and their use
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was not specified in the plant emergency procedures.
On these bases, the calculation determined that the instrument error of 110.98 psi was acceptable.
The team does not agree that the 116% error is appropriate since containment vacuum conditions could not be monitored.
In addition, use of these instru-ments should be prescribed in the plant emergency procedures. The team is also concerned that the containment isolation provisions provided are not in com-pliance with the FSAR commitment (page 6.2-131) of double barriers for con-tainment isolation.
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Observation No. NEB-3 - Essential Setpoint Calculations In HVAC calculation SQN-APS5-005 (B45860908236) there are five criteria pro-vided for determining whether a setpoint calculation is essential or not.
One
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l of the criteria used a redundancy argument in that the existence of a redundant instrument would, by implication, permit the instrument under review to be designated as non-safety related.
The team does not agree that this is a valid criterion for such determinations.
During the team's review of this calculation, the redundancy criterion had not been used as a basis for classification of any HVAC instrument setpoint calculations.
EEB OBSERVATIONS Observation No. EEB-1 - Battery and Charger Sizing The team is concerned about the adequacy of the installed system based on the
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i team's review of calculation CPS-004 for class IE battery and charger sizing.
Specifically, The evaluation of imposed loads on the batteries does not address motor
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in rush currents, worst case current requirements of 6.9 kv class IE switch gear during load shedding, and random loads.
However, connected
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loads evaluated were assumed to be energized for the entire duration of
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the discharge.
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The evaluation of the imposed load does not utilize the name plate rating i
of class IE inverter which is rated for 20 kva, but uses only 15 kva which represents the attached loads.
The team found no control mechanism to limit the loading to 15 kva.
The team noted that submergence calcula-tion SQN-SBMG-1 uses the rated value of 20 kva and a trip setting of the
inverter feeder breaker of 175 amps.
At the minimum voltage of 105 v the inverter could draw an 18.5 kva load which is not consistent with 15 kva r
assumption used in calculation CPS-004.
This could result in overloading of the battery and charger which would reduce the system's ability to supply the required voltage.
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The charging time of 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> was arrived at by using a charging efficiency of 75%.
This value was assumed and needs verification from the vendor, but was not listed as an unverified assumption.
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The plant installed battery system has a lower capacity (14 positive plates)
than the required capacity (16 positive plates) determined by recent TVA
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calculations.
To justify the adequacy of the installed system, TVA has used a lower aging factor, making the end of life 89% instead of 80% as originally calculated.
However, the team noted that information regarding changing end of life (from 80% to 89%) was not passed to Sequoyah operations department when the calculation was completed in June 1986.
This informa-tion was not passed to operations until another issue regarding battery performance was discovered during an INPO visit in November-December 1986.
This demonstrates a breakdown in comnunication between the design organi-zation and operations.
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The battery is approximately 14 years old; 20 years is considered to be the
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normal life.
Considering the age of the battery and also considering the ommissions in the calculation the team is concerned about the adequacy of the i
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following items:
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Momentary in rush loads and random loads, j
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Maximum rating of the connected loads.
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In addition, TVA should verify that all loads added to the Vital DC System as a result of ECNs are included in the battery sizing calculation.
TVA.should j
also verify by test prior to restart that calculated battery capacity can be supplied by the installed battery system.
Observation No. EEB-2 - Breaker Coordination The team's review of calculation APS-003 for the breaker coordination study indicated that TVA found that the feeder breakers for the ERCW 480V boards IA-A, IB-B, 2A-A, 28-B, and the feeder breakers for the diesel generator 480V
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boards 1Al-A, IA2-A, IB1-B, 182-B, 2Al-A, 2A2-A, 2B1-B and 282-B are not coordinated with the individual load breakers.
The team reviewed TVA's correc-tive actions and noted that TVA intends to reset the feeder breakers. Review of the breaker coordination curves for diesel generator 480V boards revealed that resetting the breakers will not resolve the problem since the breakers cannot t
be reset to the appropriate valves.
It would appear that these breakers either have to be replaced or the sensor mechanism has to be replaced.
TVA intends to
perform this corrective action after restart due to the fact that the loss of one board in ERCW and DG system would not degrade the safe shutdown capability
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of plant.
The team found that the calculated instantaneous trip settings (700%) do not agree with the trip settings shown on the drawings (900%) #3591 A12 Rev. 5, 3591A 14 Rev. 4, 3591A 16 Rev. 6, 3591A 18 Rev. 4, 3591 A20 Rev. 4, 3591 A22 Rev. 4, 3591A 24 Rev. 6 and 3591A 27 Rev. 3.
TVA informed the team
that this mistake was due to improper verification of installed breaker data by i
the walkdown team.
Observation No. EEB-3 - 120V AC and DC Solenoid Valve Voltage i
The team's review of calculation SQN-CPS-001 for 120V AC and 125V DC solenoid valve voltage study indicated the following:
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The calculation of cable impedance does not address the effects of junction
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boxes, electric conduit seal assemblies, cable slacks and higher temperature (ambient) of operation during accident.
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In the 120V AC voltage study, assumptions were not listed in a dedicated
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section but were scattered all over in the calculation and were indicated by a word " assume" in parenthesis next to the value assumed.
EEB personnel indicated to the team that all such unverified assumptions will be verified before restart.
The team believes if all assumptions were listed in a dedicated section, the chance of some assumptions not being verified (due to being buried in the calculation) will be reduced.
Observation No. EEB-4 - Setpoint Accuracy Calculation for Replacement of Rosemount with Gould Transmitters The setpoint accuracy calculation for the containment annulus differential pressure transmitters PDT-65-80, -82, -90, and -97 was provided in I&C Calculation File 29 (RIMS 843850830903).
This calculation addressed Rosemount transmitters.
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An ECN had been initiated to replace these transmitters with Gould transmitters for environmental qualification purposes.
The setpoint accuracy calculation
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was not updated to reflect use of Gould transmitters.
Observation No. EEB-5 - Assumed Value Error for Sensor Measurement and Test Equipment Accuracy
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In the setpoint accuracy calculation for RWST level transmitters LT-63-50, -51,
-52, and -53 (RIMS B43860228901), the assumed value for sensor measurement and test equipment accuracy (SMTE) was listed as 10.05 percent of span rather than 10.5 percent of span.
In the numeric portion of the calculation, the SMTE term actually used the correct value of 0.5 percent.
Hence, this appears to be a single random human error that had no impact on the calculation results.
CEB OBSERVATIONS Observation No. CEB-1 - Rigorous Piping Analysis N2-67-8A The geometry for this piping analysis consists of an L-shaped branch line composed of 1-inch and 2-inch diameter pipe with a 1-inch by 2-inch relief valve located at the branch line elbow.
This branch line is supported from 24-inch ERCW run pipe located in the auxiliary building.
The team noted the following:
(1) The piping S mensions and configuration detailed on piping stress isometric
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47K450-115 Rev. 2 are not completely defined on piping physical drawing 47W450-4 Rev. 45.
Only the larger 10-foot 3-inch span dimension for the 1-inch diameter line is detailed, and an elevation for this span is not shown.
In addition, a designation for the valve is not provided.
(2) The piping stress isometric details a span length of 8-feet 9-1/4-inches for the 1-inch diameter pipe, which is less than the 10-foot 3-inch span length detailed on the piping physical drawing.
(3) The 1-inch by 2-inch TVA Class C relief valve was purchased under the
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requirements of TVA procurement specification 72C53-92795-10.
Attachments A and 8 of the specification exempt TVA Class B and Class C valves less i
than 6-inches in diameter from the seismic qualification criteria which are stipulated for the valve vendor in Appendix F, Guide for Qualification of Seismic Class I and Seismic Class II Mechanical and Electrical Equipment, which forms a part of the procurement document.
This exemption is not consistent with the rigidity requirements specified for TVA Class B and Class C valves in FSAR Tables 3.9.2-1 and -3, or FSAR Table 3.2.1-2, which indicates that TVA Class C valves in the ERCW system have been seismically qualified by test, or SQ8-DC-V-7.4, ERCW Design Criteria, Section 3.7.1.2, Seismic Requirement, which requires the ERCW system to be designed to seismic Category I criteria.
Observation No. CEB-2 - Structural Steel Sizing Calculations The review of TVA calculation PWP 840625 808 auxiliary building access and sup-port platforms and stairs, showed that various steel members shown on drawings 48N1210, 48N1211, 48N1213 and 48N1214 do not have any calculations.
There is no design documentation supporting the structural adequacy of these members.
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Observation No. CEB-3 - Structural Steel Details
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During the review of TVA calculation PWP 840625 808 auxiliary building access and support platforms and stairs, the team noted that there was a lack of evaluation for structural steel connections and details.
TVA uses welded and bolted connections for the structural steel in the auxiliary building.
The calculations reviewed for drawings 48N1210, 48N1211, 48N1213, 48N1214, 48N1215, 48N1216, and 48N1216-1 did not show evaluations for welds and bolts.
Also, the stress calculations for the stiffener plates used in drawing 48N1216 Detail A could not be located.
Observation No. CEB-4 - Platform Steel Calculations and Drawings The review of TVA calculation PWP 840625 808, addressing auxiliary building access and support platforms and stairs, showed that for drawing 48N1214 one of the structural members was calculated to be a channel section 9[13.4. The review of drawing 48N1214 Rev. 3, Plan at elevation 692'-6", shows this member to be a channel section 8[11.5. Therefore, a member smaller than what was required by calculations was installed.
A similar situation exists for comp' uter models prepared for the analysis of steel platform at elevation 724'-3.
The calculations performed for drawing 48N1216-1 shows that the two computer models prepared do not match the configuration of the platform as shown on drawing 48N1216-1 Rev. 11.
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Observation No. CEB-5 - Revisions to Steel Platform Calculations The review of TVA calculation PWP 840625 808, addressing auxiliary building access and support platforms and stairs, showed that calculations related to drawing 48N1215 were revised to add additional hanger loads.
The structural steel was analyzed for adequacy to carry the new loads, however, this reanalysis was not checked or verified.
Also, the components which resulted in the additional support loads were not identified.
Observation No. CEB-6 - Seismic Loads for Steel Platforms m..
The review of TVA calculation PWP 840625 808, addressing auxiliary building access and support platforms and stairs, showed that steel platforms were not consistently designed for seismic loads.
In certain cases, seismic loads were not included in the load combination to determine structural adequacy.
This can be seen in the calculations performed for drawings 48N1211, 48N1213 and 48N1214.
In addition, calculations for drawing 48N1216 showed that steel stress allowables were lowered by 1/3 to account for seismic loads.
A later revision used the
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full value of the allowable to evaluate additional conduit loads without con-sideration of seismic loads.
No justification was provided either for lowering the stress allowables or for subsequently raising them.
The same calculation assumed the platforms were rigid to determine seismic
loads.
There were no substantiating frequency calculations to justify the rigidity assumptions.
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i Observation No. CEB-7 - Rigorous Piping Analysis N2-67-3A-4-l
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The piping geometry for this analysis primarily consists of 6-inch diameter i
ERCW pipe which is supported from the containment building at elevation 682-feet i
6-inches and is rigidly attached to the steel containment vessel (SCV) at penetrations X-59 and X-63.
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The team found that:
(1) Rev. O of the piping analysis documented acceptance of accelerations j
greater than 2 and 3 g for valve 2-FCV-67-88 adjacent to penetration X-59.
Rev. 1 of the piping analysis evaluated a valve operator change due to ECN L-5824 (replacement of motor operators due to lack of NUREG-0588 y
i environmental qualification of motor insulation). TVA performed a " study" to evaluate the effects of increased valve operator weight and center of
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gravity dimension on the valve dynamic response, but did not document this evaluation in Rev. 1 of the piping analysis.
l (2) Rev. 2 of the piping analysis references four CEB "as-built" pipe support gap evaluations per ECN L-6706.
The team found errors in two of the four
"as-built" evaluations.
The calculation to determine if the 1-15/16-inch
"as-built" radial gap for pipe support HERCW-10 could accommodate the pipe movement at that location was incorrectly performed.
No calculation was performed to evaluate the 3/4-inch "as-built" radial gap for pipe support
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HERCW-14.
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j (3) The team checked the documents which qualify pipe penetrations X-59 and X-63 and found that the original qualification documents for the Unit 2
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i penetrations had been incorrectly superseded by the qualification documents i
which CEB prepared for the corresponding Unit 1 penetrations.
I Observation No. CEB-8 - Qualification of Seismic Category I Buried Pipe
The team asked CEB to provide the criteria used to qualify the seismic
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Category I buried ERCW pipe which runs between the ERCW pumphouse and the
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auxiliary building, and the associated calculations.
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CEB indicated that buried pipe is seismically qualified in accordance with design criteria SQN-DC-V-13.5, Design Criteria for Seismically Qualifying Buried Piping Systems, which CEB issued on September 5,1972.
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f However, CEB cannot retrieve the calculations which document the seismic
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l qualification of the buried ERCW pipe.
Moreover, CEB did not identify these
" essential" calculations as missing to evaluate and assess the need for regen-
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eration of these calculations before restart of SQN Unit 2.
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Observation No. CEB-9 - Reinforcing Bar Cut Evaluation The review of TVA calculation PWP 840920 705, south main steam valve room
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auxiliary building, showed that two reinforcing bars were cut and documented on FCR 2255. An evaluation of the slab was made for these cuts to determine structural adequacy.
This evaluation failed to consider the seismic load on j
the slab.
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slab is structurally adequate to carry all the loads including seismic.
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j Observation No. CEB-10 - Weld Evaluation for Conduit Support TVA calculation B25 850304 300 was performed to design conduit support MK115
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for FCR 2420.
A detailed computer analysis for the welded structural frame
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showed the steel member stresses were within allowables.
Since all the I
structural members were connected by welds, an evaluation of the welds for i
maximum moment and shear should have been performed.
The team could not locate such weld evaluations.
Observation No. CEB-11 - Pipe Rupture Evaluation for Concrete TVA calculation PWP 840920 705, south main steam valve room auxiliary building, was revised in 1976 to accommodate the changes in pipe rupture loads.
An evaluation of the floor slab was made by assuming that the concrete compressive strength would be 6500 psi compared to the original design value of 4000 psi.
A justification for this assumption was not documented.
i Even with assuming higher strength for concrete, the calculations showed that in certain areas the concrete and ste.el allowable stresses were exceeded by 19 and 16%, respectively.
Stresses above the allowables were accepted by both
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the calculation preparer and checker without any documented technical justification.
Observation No. CEB-12 - Use of Variable Damping for Conduits
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i TVA design criteria SQN-DC-V-13.10, seismically qualifying conduit supports, was revised on 11/20/85 to include the span lengths and the support loads as developed in TVA calculation B41 851105 028.
A review of this calculation
showed that a variable damping ratio was used in determining the seismic H
loads on the conduit supports.
TVA used a damping value of 2% for frequencies
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greater than 10 Hz and 5% for frequencies less than 10 Hz.
TVA's commitment, as shown on Table 3.7.2.4 of the Sequoyah FSAR, is a constant 2% damping value for the safe shutdown earthquake.
The use of a higher damping value would lower the conduit support loads and might yield an unconservative design.
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ATTACHMENT B
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List of Persons Contacted Name TVA Organization / Title Beth Hall SQN Licensing Engineer
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W. Pennell Manager, Engineering and Technical Services D. L. Williams Manager, Knoxville Licencing R. C. Weir NEB Chief Engineer F. A. Koontz NEB Assistant Chief Engineer C. A. Chandley MEB Chief Engineer R. Corbett MEB Assistant Chief Engineer R. Barnett CEB Chief Engineering R. O. Hernandez CEB Assistant Chief Engineer K. S. Seidle CEB Assistant Chief Engineer W. S. Raughley EEB Chief Engineer J. Hutson EEB Assistant Chief Engineer A. Capozzi Manager Engineering Assurance F. E. Denny EA J. S. Colley EA J. P. Little MEB R. McColl MEB F. Carr MEB J. Purkey MEB
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T. B. 016erding MEB i
G. A. Silver MEB W. V. Chen MEB J. W. Warner MEB J. Stellern MEB K. B. Oxner MEB J. B. Hubble MEB L. W. Boyd MEB R. M. Devault NEB R. S. McKeehan NEB K. D. Keith NEB
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W. J. Kagay CEB D. Riffert CEB S. D. Stone CEB K. C. Brime CEB A. D. Sowars CEB J. A. Ellis CEB B. Neely CEB 5. Taylor CEB M. Maxwell CEB R. Alexander CEB J. Peyten CEB N. Perry CEB M. Cones *
CEB K. L. Mcalg CEB J. Rochelle CEB R. C. Widliams EEB i
R. R. RCaves EEB J. Niceiy EEB
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LPDR DQAVT Reading QAB Reading i
RParkhill, IE LSpessard, IE EVImbro, IE j
HJMiller, IE
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BKGrimes, IE JMTaylor, IE REArchitzel, IE JGKeppler, SP SDEbneter, SP GZech, SP BDLiaw, SP
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KBarr, RII
BBHayes, 0I
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Inspection Team (8)
JHolonich, NRR
Resident Inspector NSIC Regional M=inistraterrRI-I WCline, RII CRStaMerNRR~
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04/S /87 04/3/87 04/J/87 04/J/87 04/'7/87
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