ML20134G185
| ML20134G185 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 01/13/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20134G156 | List: |
| References | |
| 50-327-96-16, 50-328-96-16, NUDOCS 9702100310 | |
| Download: ML20134G185 (18) | |
See also: IR 05000327/1996016
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U. S. NUCLEAR REGULATORY COMMISSION
REGION ll
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Docket Nos: 50-327, 50-328
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Report No: 50-327/96-16, 50-328/96-16
Licensee: TVA
Facility: Sequoyah Units 1 & 2 !
Location: Sequoyah Access Road
Hamilton County, TN 37379
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Dates: September 23-27,1996; November 4-19,1996 and
December 16-19,1996
Inspectors: C. Smith, Reactor inspector
N. Merriweather, Reactor Inspector
Approved by: C. Casto, Chief, Engineering Branch
Division of Reactor Safety
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9702100310 970113
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PDR ADOCK 05000327
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EXECUTIVE SUMMARY -
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Sequoyah Nuclear Plant, Units 1 & 2
NRC Inspection Report 50-327,328/96-16
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This special inspection included detail reviews of corrective actions implemented for J
! Problem Evaluation Reports (PERs) No. SQP900372PER, Nuclear Fuel Design '
Changes Not Reconciled / Reflected in Design Basis Documents (DBDs); and
i SO950021PER, Obtain Operability Evaluation for SQNP: Review of WBPER940576. ,
Additionally, a review of the licensee's transition plans for implementing the EQ !
program after Phase 1 site engineering re-organization had been completed was
performed. An Unresolved item involving inadequate safety assessment of Rod
j Control System plant modification was closed and a Violation of 10 CFR 50 Appendix
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B, Criterion lli was cited.
i Results:
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An unresolved item concerning performance of an inadequate 10 CFR
50.59 Safety Evaluation that resulted in an Unreviewed Safety Question.
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An unresolved item concerning untimely revision to the EQ Binders and
EEB calculations.
An unresolved item concerning inadequate design control for
" Nonconforming Plant Conditions.
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An unresolved item concerning the technical acceptability of reducing
the calculated free field beta dose both inside containment and the i
annulus by 50 percent.
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An inspector followup item concerning inconsistent FSAR descriptions of
the reactor power level.
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A violation for inadequate design controls for Rod Control System plant
modification.
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Report Details
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{ 111. Enaineerina
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E1 Conduct of Engineering
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E1.1 PER No. SQP900372PER, Nuclear Fuel Design Changes not
Reconciled / Reflected in Design Basis Documents (DBDs)
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a. Insoection Scope
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i The inspector reviewed PER No. SQP900372PER in order to evaluate the
i adequacy of the licensee's root cause analysis, extent of condition evaluation,
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and developed corrective actions for 10 CFR 50.49 identified deficiencies.
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! b. Observations and Findinas
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Condition Adverse to Quality Report (CAOR) SOP 900372PER, dated
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September 18,1990, documented fuel related design changes made by TVA
! which had not been reconciled or reflected in design basis documents. An
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increase in the average core burnup from 650 EFPD to 1000 EFPD resulted in
an increase in the amount of core activity that is assurned at the start of a
design basis LOCA. Because of this there was an increase in the 100 day
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. integrated accident dose that electric equipment important to safety and
qualified to 10 CFR 50.49 must withstand. TVA management prepared a
Justification for Continued Operation (JCO), (TVA-91-293), to demonstrate that
the requirements of 10 CFR 50.49 were still being met by equipment that had
! previously been environmentally qualified based on a source term of 650
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EFPD. The inspector reviewed the JCO and determined that TVA had
,' concluded that the JCO bounded reactor core designs with U235 fuel having
- - average enrichment less than 4.5 percent 1000 EFPD burnup.
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j- The NRC in a letter dated November 30,1993, Subject: Evaluation of
l Increased Fuel Burnup on Equipment Qualification Sequoyah Nuclear Plant
Unit 1 and 2, transmitted the results of the staff's review of the above JCO to ;
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TVA. The staff concluded that the JCO was not appropriate and TVA was l
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requested to perform a reassessment of equipment qualification for 1000 ;
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EFPD burnup using an acceptable source term (TID-14844) and resubmit the ;
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JCO for the staff's review. TVA performed the reassessment and in a letter '
dated March 4,1994, transmitted the JCO for the staff's review. The NRC in a
- letter dated April 8,1994, informed TVA that the staff had reviewed the
- reassessment and determined that it satisfactorily responded to the staff's
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concern.
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l The inspector reviewed the results of the EQ reassessment titled " Review of
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1000 EFPD with 4.5% U235 Enrichment", performed in support of the JCO
submitted to the NRC. Corrective action plans developed and implemented for
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CAQR No. SOF870012, and SQP870165 were also reviewed during this
i inspection. The specific issues reviewed and the results of these reviews are
discussed in the paragraphs below.
Technical Adeauacy of 10 CFR 50.59 Safety Evaluation
l CAOR No. SOF870012 was written on March 19,1987, to document a !
l condition where the core average exposure limit of 26154 MWD /MTU specified
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in FSAR Table 15.1.7-1 would be exceeded in Unit 1 cycle 4 operation. The
suggested corrective action was to calculate the offsite dose using 1000
Effective Full Power Day (EFPD) and revise the FSAR to reflect the results of
the revised calculation. CAQR No. SQP870165 was written to document the
results of EGTS tests which demonstrated slow response of the dampers to
pressure changes and missing design criteria which specified what the
response time should be. The apparent cause of the dampers slow response t
to pressure changes was due to the use of a pressure indicating controller
having only a proportional band with no reset function. The inspectors '
reviewed a 10 CFR 50.59 Safety Evaluation dated December 2,1987, :
prepared by the licensee to make changes to the FSAR for resolution of the
above deficiencies. Based on this review the inspectors determined that the
following tables in the FSAR were being revised; 1) Table 15.1,7-1, Core and
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Gap Activities Based on Full Power Operation for 650 Days Full Power: 3565
MWt; 2) Table 15.5.3-3, Emergency Gas Treatment System Flow Rates; 3)
Table 15.5.3-4, Offsite Doses From Loss of Coolant Accident; 4) Table 15.5.3-
7, Control Room Personnel Doses for DBA Post Accident Period. Additionally,
changes were being made to selected portions of the narrative descriptions in
the FSAR to facilitate resolution of CAQR Nos. SOF870012 and SQP870165. ;
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FSAR Chapter 15, Table 15.1.7-1 was revised to show new source terms
based on 1000 EFPD operation. The results of offsite dose calculations
performed by the NRC in support of licensing actions were documented in
Safety Evaluation Report (SER) Supplement No.1, dated February 1980. The
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inspectors reviewed section 15.4, of the SER to confirm if the FSAR changes
and offsite dose analysis were acceptable and complied with the current
licensing basis. One discrepancy was identified during this review. Offsite
radiation doses contained in the SER Supplement No.1, Table 15-1,
Radiological Consequences of Design Basis Accidents, was calculated by the
NRC based on the assumption that Unit 1 reactor will be operated at a power
level not in excess of 5% of the rated power of 3582 MWt. Table 15-2 of the
SER, Assumptions Used in the Cr.culation of Loss of Coolant Accident Doses, ,
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also showed the reactor power level as 3582 MW thermal. This value of i
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I reactor power level used in the offsite dose calculation was different from that
used by TVA which was 3565 MW thermal. The guidance delineated in TID-
14844, Calculation of Distance Factors For Power and Test Reactor sites,
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' dated March 23,1962, requires the use of the reactor rated power level
(megawatts) in the calculation which determines the radio nuclide inventory of
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specific isotopes. Numerous inconsistencies concerning the reactor rated
power were identified in FSAR Tables 15.1.2-1; 15.1.7-1; and all the tables in
FSAR section 15.5. The guaranteed core thermal power in Table 15.1.2-1 was
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listed as 3411MW thermal. In Table 15.1.7-1 it was listed as 3565 MW
thermal and in all the tables of FSAR Section 15.5 it was listed as 3582 MW
l thermal. The maximum power level authorized in the facility operating license
i power level is identified as IFl 50-327,328/96-19-05, FSAR Inconsistent
- Description of Reactor Power Level.
1 The results of the above reviews demonstrated that the licensee had
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conslered the consequences of offsite radiation doses to the health and
! safety of ti;e public based on 1000 EFPD operation. Additional reviews of the
! 10 CFR 50.69 Safety Evaluation, however, revealed that the licensee had not
l evaluated whether the increase from 650 to 1000 EFPD operation affected the
! qualification status of equipment that had previously been qualified to a source
term that was based on 650 EFPD criterion. The increase in EFPD from 650
i to 1000 because of fuel related design changes had created an increase in the
amount of core activity that was assumed at the start of a design basis LOCA.
The increase in the core activity resulted in an increase in the 100 day
integrated accident dose that environmentally qualified equipment must
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withstand. The licensing basis for the 10 CFR 50.59 EQ Program was 650
{ EFPD burnup and this requirement was exceeded by Unit 1 cycle 4 operation
i on December 29,1989 and Unit 2 cycle 3 on December 30,1988. This
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"Unreviewed Safety Question" involving failure of the 10 CFR 50.59 Safety i
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Evaluation to address the requirements of environmentally qualified equipment
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resulted in nonconforming and unanalysed plant conditions from December 30,
1988 until July 30,1990, when design basis Calculation TI-RPS-48, Integrated
!- Accident Dose inside Containment and Annulus, Revision 3, was prepared to
! calculate the 100 day integrated accident dose based on the 1000 EFPD
burnup criterion. This item is identified as unresolved item URI 50-327,328/96-
l 16-01, inadequate Safety Evaluation Resulted in Unreviewed Safety Question.
! Corrective Actions implemented for Nonconformina Plant Conditions
Problem Evaluation Report PER No. SOP 900372PER was prepared on
December 18,1990, to document a condition where Nuclear Fuels (NF)
] made core design changes which had not been reconciled or reflected in
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current Nuclear Engineering (NE) design basis documents. A "Cause
Analysis" was performed for this deficiency and the apparent cause was ;
determined to be lack of procedural controls to ensure adequate interface
reviews and appropriate funding for those reviews. Corrective action plans '
developed and implemented for recurrence control included:
1. Revising Corporate Standard 9.2 for core alterations and core
hardware changes to ensure adequate interfaca reviews and
appropriate funding for these reviews.
2. Establishing requirements for NE to provide NF a list of fuel and
i core related parameters which affect engineering calculations and
require review on a cycle specific basis.
l 3. Revising NF Instruction 3.0 to ensure that other design basis
l documents impacted by core component design changes were
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l Other corrective actions which required revising the EQ Binders and Electrical
l Engineering Branch (EEB) calculations to incorporate updated environmental
l . conditions were delayed and transferred to PER No, SQ940040ll, TROI action
- Item No. 36. The inspectors reviewed a copy of TROI Action item No. 36 ;
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dated September 12,1996, and verified that this item was still open.
The licensee prepared a JCO dated September 4,1991, which was applicable
to both Units and would permit continued operation until TVA revised the
design documents to incorporate the 100 day integrated accident doses that
were caused by the 1000 EFPD burnup criterion. TVA's JCO was based on
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the conclusions contained in a document titled " Tennessee Valley Authority,
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Sequoyah Nuclear Plants Units 1 and 2, increase in the 100 Day Integrated
Dose to Equipment in Containment Associated with increased Fuel Burnup,
l Justification for Continued Operation." The JCO stated that TVA will
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reevaluate SONP design basis following the NRC's finalissuance of the new
TID-14844 values in order to eliminate repetitive efforts of revising the EQ
Binders.
On July 18,1992, TVA management prepared JCO No. SQJCO92-013,
Revision 0, and extended the time for implementing corrective actions related -
to TROI Action item No. 36. This extension request was approved by the Site
Vice-President on August 6,1992. On September 17,1993, JCO for
SOP 900372PER was extended by a corrective action request. The corrective
j action request was approved by TVA management on September 20,1993.
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Prior to preparation of the JCOs and during the intervals of time when TVA
management postponed implementing the corrective action to revise the EQ l
Binders and EEB calculations, the core average exposure for both Units
exceeded 650 EFPD operation on the dates listed:
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Unit NO. Cycle No. Date EFPD Exceeded l
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1 4 12-29-89 '
1 5 06-09-91
1 6 11-29-92
1 7 04-02-95
2 3 12-30-88
2 4 05-24-90
2 5 09-28-91
2 6 01-03-94
2 7 10-05-95
On November 30,1993, the NRC transmitted the results of their review of the
' Westinghouse Technical JCO for SONP" to TVA. WA was informed that the
JCO was technically inadequate and that it should be prepared in accordance
with the guidelines of TID-14844. TVA was also requested to perform a
reassessment of equipment qualification based on 1000 EFPD criterion using ,
an acceptable source term and submit it to the NRC for their review. In I
response to this request on February 11,1994, WA prepared "JCO for PER
No. SQP900372PER" which bounds reactor core designs with U235 average
enrichment of less than 4.5% and 1000 EFPD. This JCO included Unit 2 cycle
6, Unit 1 cycle 7, and Unit 2 cycle 7 fuel cycle operation. The NRC reviewed :
"JCO for PER No. SQP900372PER"and concluded that TVA's equipment
qualification reassessment satisfactorily responded to their concern. The
results of this review was transmitted to TVA on March 4,1994.
The inspectors determined that the licensee had continued plant operations
under the JCO without revising the EQ Binders and EEB calculations. This
untimely corrective action for revising the EQ Binders and EEB calculations is
a concern and is identified as unresolved item URI 50-327,328/96-16-02,
Untimely corrective action for nonconforming plant conditions.
Desian Control implemented for Nonconformina Plant Conditions
On July 30,1990 TVA management approved design basis calculation TI-RPS-
48, Integrated Accident Dose inside Primary Containment and Annulus,
Revision 3. This analysis was performed to determine the integrated accident
doses inside the primary containment for equipment qualification based on the
EFPD for calculating the equilibrium reactor core activity being increased from
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-650 EFPD to 1000 EFPD. The analysis was based on the assumption that
core activity is instantaneously released (at t=0) within the primary containment
! in the following fractions of the core inventory.
100 % Noble Gases
50 % lodines
50 % Cesium
l 1% Other Fission Products
Revision 2 of this calculation used a burnup of 650 EFPD and was previously
the calculation of record for demonstrating compliance with the requirements of
10 CFR 50.59. The results of Revision 3 of the calculation when compared to
the 100 day integrated accident doses in revision 2 were as follows:
Location TI-RPS-48. R2 TI-RPS-48-R3
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Upper Containment
Gamma 3.8 E7 3.0 E7
Beta
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4.7 E8
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8.3 E8
Instrument Rooms
Gamma 1.048 E7 1.6 E7 i
Beta 4.7 E8 8.3 E8 i
Lower Containment
Gamma 2.8 E7 2.5 E7 I
Beta 4.7 E8 8.3 E8
Accumulator and Fan Rooms
Gamma 1.048 E7 1.6 E7
Beta 4.7 E8 8.3 E8 'I
Raceway
Gamma 1.048 E7 2.4 E7
Beta 4.7 E8 8.3 E8
Ice Condenser Bed
Gamma 1.34 E7 2.3 E7
Beta 4.7 E8 8.3 E8
Gamma 1.3 E7 5.9 E6
! Beta 5.0 E5 1.38 E6
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l A significant increase in free field Beta radiation resulted from the 1000 EFPD
bumup criteria. The results of these calculations were never incorporated in
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Calculation TI-ECS-55, Summary of Harsh Environment Conditions for
Sequoyah Nuclear Plant. As a consequence the environmental data drawings
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series Number 47E235 were never revised to reflect the integrated accident
doses caused by the new source terms based on 1000 EFPD operation.
l Additionally, FSAR Figures 3.11.2-1 and 3.11.2-2 were never revised to reGect
! the new 100 day integrated dose based on 1000 EFPD operation. The
accident doses on the FSAR Figures were not consistent with the design basis
of 1000 EFPD delineated in FSAR Table 15.1.7-1. This failure to control plant
configuration and ensure that actual plant configuration is accurately depicted
on drawings and has been reconciled with design basis is of concern and is
identified as one example of unresolved item URI 50-327,328/96-16-03
Inadequate design control for " Nonconforming Plant" conditions.
On December 12,1991, TVA management approved design basis calculation
TI-RPS-48, Revision 5, " Integrated Accident Dose inside of Primary
Containment and Annulus," to document the 100 day integrated accident dose
based on 650 EFPD burnup criteria. The calculation was prepared to
implement TVA's management decision to temporarily reduce the 1000 EFPD
burnup criterion. Calculation TI-ECS-55, Revision 16 was prepared to
incorporate and clarify usage of the Containment Buildings design basis post
accident radiation doses determined from calculation TI-RPS-48, Revision 5.
Additionally, plant modification DCN No. 508114A, Revision 16, revised
environmental drawing sheets 45,47, and 48 to replace radiation values that
were no longer conservative. The inspectors concluded that these drawing
revisions were not an accurate representation of actual plant configuration
based on FSAR Amendment 5 to table 15.1.7-1 which delineated 1000 EFPD.
On June 9,1991, Unit 1 cycle 5 operation exceeded the 650 EFPD burnup
criterion that was being used as the basis for the 100 day integrated accident
doses shown on the environmental drawings. This event was preceded by
Unit 1 cycle 4 and Unit 2 cycles 3,4 and 5 having average core exposure in
excess of 650 EFPD. TVA's management failure to control plant configuration
and ensure that actual plant configuration is accurately depicted on drawings
and has been reconciled with design basis is of concern and will be identified
as another example of unresolved item URI 50-327,328/96-16-03.
On March 4,1994, TVA transmitted "JCO for PER No. SQP900372," dated
February 11,1994, to the NRC for their review. One hundred day integrated
gamma, and beta accident doses for the 1) the upper containment; 2) lower
containment; 3) Accumulator Fan Instrument Rooms; 4) Raceway; 5) Ice Bed
Condenser and 6) Annulus were listed in the JCO. The inspectors reviewed
the JCO and determined that the radiation values delineated in the JCO were
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not supported by an approved analysis. A formal calculation had never been
prepared, reviewed and approved to determine the 100 day integrated accident
dose inside the containment and the annulus. The inspectors expressed
concern to TVA management concerning the apparent non-compliance with the
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requirements of the design control program which requires that design
{ analyses shall be performed in a planned, controlled, and correct manner. In
l response to the inspector's concern TVA attempted to reconstitute the analysis
l via computer runs on November 7,1996. The raw computer data that resulted
from this effort was not comprehensible to the inspectors. Calculation No.
SBNSOS2-0163, Dose in Containment and Annulus with 1000 EFPD Burnup
and 4.5 percent U235 Enrichment, was finally prepared and approved on
l November 15,1996 to address the inspector's concern. The results of this
l calculation were reviewed by the inspectors and were determined to be
comparable to the 100 day integrated accident doses for 1000 EFPD at 4.5
percent U235 listed in the JCO. TVA's failure to comply with the requirements
. of the design control program concerning engineering analyses is of concern
and will be identified as one example of URI 50-327,328/96-16-03. Inadequate
design control for Nonconforming Plant Condition.
Technical Acceptability of Roducina Calculated Free Field Beta Dose by 50
Percent
Design Calculation SON-TI-RPS-048, Revision 6 issued October 1994, is the
design basis calculation for the maximum 100-day integrated doses inside
containment and the annulus with source terms for power levels of 3565 MWt.
with average core burnups of 1000 EFPD and enrichments of 5 percent weight
U235. The maximum free field Beta dose in air inside containment was
calculated to be 6.311E+8 rads over 100 days. The licensee then made the
assumption that the maximum calculated free field Beta dose could be reduced
by a factor of 1/2 to account for a semi-infinite source geometry due to
component self-shielding effects. The 50 percent reduction resulted in a
surface Beta dose of 3.156E+8 rads that was below the previously analyzed
Beta Dose given in Revision 2 of the calculation at 650 EFPD and 3565 MWt.
using TlD 14844 source terms. NUREG 0588, For Comment Version and
Revision 1, Section 1 contains positions related to the establishment of the
service conditions for areas inside and outside containment to which
equipment should be qualified. It includes guidance for determining the
radiation environments expected to occur during a design basis event
condition. In Section 1.4(7), Radiation Conditions inside and Outside
Containment, it requires that the maximum Beta dose at the surface of
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unshielded equipment be taken as the free field Beta dose calculated for a
point at the containment center. The licensee did not follow this guidance
when they took the 50 percent reduction fnr self-shielding. The licensee
indicated that this 50 percent reduction is standard industry practice and has
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l' been previously accepted by NRC. The inspector acknowledged the licensee's ;
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position on this concern and indicated that this issue was unresolved pending l
! further review by NRC.
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l The acceptability of the licensee reducing the calculated free field Beta Dose l
! both inside containment and the annulus by 50 percent is unresolved and will
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be identified as URI 50-327,328/96-16-04.
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c. Conclusion
! The inspectors concluded that the licensee failed to implement adequate
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design controls for reactor core design changes and failed to take prompt and 1
effective corrective action for nonconforming plant conditions identified since ;
y September 18,1990. Three violations were identified. Additional review by ;
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the NRC has resulted in these violations being changed to URis pending
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additional NRC reviews. One unresolved item and ons inspector followup item '
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was also identified.
- E2 Engineering Support of Facilities and Equipment
E2.1 'PER No. SQ950021PER, Obtain Operability Evaluation for SONP: Review cf
WBPER940576 '
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j a. Inspection Scope
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! The inspector reviewed PER No. SO950021PER in order to evaluate the
adequacy of the licensee's root cause analysis, extent of condition evaluation,
, and developed corrective actions for 10 CFR 50.49 identified deficiencies.
l b. Obsentations and Findinas
! Watts Bar Adverse Condition Report WBPER940576 identified a problem with
): the pressurizer PORVs where the energized times did not agree with
limitations imposed by the EQ program. The PORVs had been energized in
l excess of 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per year via 56 cycles which exceeded energized times
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specified in the EQ binder. This issue was reviewed for applicability to
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Sequoyah. EQ binder SONEQ-SOL-002 documents that the pressurizer
PORVs are energized for a maximum of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per year. Investigation
revealed, however, that the 40 year energization time documented in the EQ
binder was nonconservative in that the Target Rock solenoid valves had been
in use since 1983.
The root cause analysis performed by the licensee was reviewed by the
inspector and was determined to have been adequately performed. Interim
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corrective actions taken to address this issue involved completing an
- Operability Determination where it was concluded that the PORVs could
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perform satisfactorily until the cycle 7 outage. The pressurizer PORV solenoid
- valves were subsequently replaced during the cycle 7 refueling outage of each ,
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unit. Corrective action plans developed for final resolution of this issue
involved a review of the SONP EQ binders to determine if revisions were
required for any EQ binder, and supporting Qualified Life, or Accident
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Degradation Equivalency Calculations. The results of this review identified 12
EQ binders that required revision. The inspector reviewed the status of ;
corrective action C.9.8 and C.9.9 and determined that the Qualified Life and '
- Accident Degradation Calculations had not been revised to reflect identified
duty cycle / operational time changes. Additionally, revisions to EQ binders
based on' the results of the above calculations have been restrained because !
j of failure to promptly complete the calculations.
l c. Conclusions
The inspector concluded that the Operability Determination performed for PER
No. SQ950021PER was technically adequate. Interim corrective actions of
replacing the pressurizer PORV solenoid coils during cycle 7 refueling outage
of each unit also demonstrated TVA's implementation of prompt corrective
action. TVA's management failure, however, to complete corrective actions -
C.9.8 and C.9.9 for an issue identified on January 13,1993 was considered
less than timely.
E6 Engineering Organization and Administration
a. Insoection Scope
The inspector reviewed the licensee's program documents that control the
environmental qualification program to verify 1) that responsibilities had been
defined and 2) requirements had been specified for establishing and
maintaining the auditable documentation demonstrating qualification of
equipment in compliance with 10 CFR 50.49. The licensee's transition plans
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ier implementing the EQ program after Phase 1 site engineering re- j
organization was also reviewed. '
b. Observation and Findinas
Procedure SSP-6.5, Electrical Equipment Environmental Qualification (EQ) 1
Program, Revision 7, is the controlling procedure for implementing the EQ
program at Sequoyah. Based on review of this procedure the inspector
determined that the program controls clearly identified functional J
responsibilities and levels of authority for adequate implementation of the EQ
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program. Training requirements for personnel engaged in EQ work activities
were also clearly identified on Appendix I of this procedure. No deficiencies
- were identified with the procedural controls in SSP-6.5.
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The inspector reviewed the licensee's transition plans for implementing the ,
10 CFR 50.49 program after Phase 1 reorganization of the site engineering i
l and material section. The following documents were reviewed during this
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Procedure SPP-9.2, Equipment Environmental Qualification (EQ)
Program, Revision 0.
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Procedure NEP-5.12, Program Requirements For Equipment
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Qualification of Electrical Equipment in Harsh Environments, Revision 1.
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Mechanical Design Standard No. DS-M18.14.1, Design Standard for
i Environmental Qualification of Electrical Equipment in Harsh
Environment, Revision 0.
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, The inspector also conducted interviews with personnel engaged in EQ work
activities from the EE/NE discipline and Maintenance Planning and Technical
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(MP/T) section. The interviews were intended to assess the level of the
worker's understanding of the EQ program requirements and to verify that EQ
i
training requirements had been met. All personnel interviewed were
j knowledgeable of the EQ program requirements and had completed EQ
j training. No deficiencies were identified with the licensee's staff involved with
EQ program activities,
l
j At the time of the inspection procedure SPP-9.2 was in the process of being
reviewed for approval by NE management for replacing SSP-6.5 upon
i
completion of the Phase 1 site engineering reorganization. This is an upper
- tier program document that delineate EQ program controls to be implemented
>
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at Sequoyah, Browns Ferry and Watts Bar. Based on this review the inspector
1
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determined that SPP-9.2 failed to adequately establish program controls for l
successfulimplementation of the EQ program at Sequoyah. Ownership of the l
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EQ program was not identified; functional responsibilities and levels of
'
, authority for implementing the program was not described; and the
implementing instructions lacked clarity and specificity because of the upper
tier nature of the procedure. The procedure also failed to identify training
i. requirements for personnel involved with EQ work activities.
'
TVA management was informed of this inspection finding. In response TVA
management told the inspector that they concurred with the findings and
] procedure SPP-9.2 would not be approved for replacing SSP-6.5 in its present
s
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form. The inspector was also advised of personnel changes that would be
implemented on October 1,1996, for Phase 1 reorganization of the site
engineering and materials section. On this date TVA management will have '
l
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only one person who have completed EQ training in the l&C section which now
has ownership of the EQ program. Similarly, one EQ trained person will be in
the MP/T section to perform EQ duties. The licensee has essentially
- de-centralized the EQ program, disbanded the dedicated staff who performed
.. EQ activities, and has now included in the position descriptions of engineering
j and MP/T personnel requirements for performing EQ duties.
i
c. Conclusion
i
The inspector concluded that the transition plan for implementing the EQ
program after Phase 1 reorganization of the site engineering section was
inadequate based on procedure SPP-9.2. Additionally, the number of
trained personnel required for performing EQ duties after October 1,1996,
does not appear to be adequate based on the numerous large scale ongoing
corrective actions presently being implemented for identified EQ deficiencies.
E.8 Miscellaneous Engineering issues
E.8.1 Employee's Concern Program
1
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a. Insoection Scooe
The inspector reviewed implementation of the licensee's Employee Concern
Program to verify that employee's concerns related to inadequacies in the 10
CFR 50.49 Environmental Qualification Program are promptly and adequately
addressed by TVA management.
b. Observations and Findinas
Numerous concerns have been expressed by TVA personnel during exit
interviews concerning the adequacy of the 10 CFR 50.49 Environmental
Qualification Program. The inspector reviewed the employee's concerns
documented in the following Concerns Resolution Program, (CRP) Files and
conducted discussion with the Concerns Resolution Staff Manager concerning
implementation of the program.
File No. ECP-96-SQ-903
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File No. ECP-96-SQ-918
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File No. ECP-96-SQ-922
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File No. ECP-96-SQ-927
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File No. ECP-96-SO-928
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File No. ECP-96-SO-991
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Based on these discussion's the inspector determined that File No. ECP-96-
i
SQ-992-F1 was prepared as a collector file for issues raised by employees
during exit interviews concerning the adequacy of SONP programs. The scope
' of the employee's concerns included inadequacies involving the 10 CFR 50.49
EQ Program; Leak Rate Testing; Appendix R; Q List; Vendor Manuals; and
i
Technical Specification Testings. TVA management had already taken actions
i
' to address these concerns. The inspector reviewed Engineering
Reorganization Assessment Report, NASQ 96023-phase 1, and verified that
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EQ concerns were addressed in this investigation. The report concluded that
although there has been a significant reduction in SON Engineering personnel,
contingency plans and tasks reassignmer,ts have been developed to ensure
responsibilities are adequately assumed by remaining site and/or contract
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personnel.
Additional EQ concerns raised by employees have been documented in File
No. ECP-96-SO-A07-F1. These issues among others have been identified as
action items to be included for review in upcoming audits. The inspector wn
informed that the results of the Engineering Reorganization Assessment-Phase
2, scheduled for January,1997, will also provide additional indepth
investigation of EQ concerns raised by TVA employees.
c. Conclusion
The inspector concluded that employee's concerns are promptly addressed by
TVA management. Concerns involving inadequacies in implementing the 10
CFR 50.49 EQ program have not yet been fully investigated to validate the
employees specific concerns. It is the inspectors understanding that the
investigations to be performed during phase two of the engineering
reorganization assessment will satisfy this requirement.
E.8.2 (Closed) Unresolved item (URI) 50-327,328/96-02-04, Omission of
Surveillance Tests for Rod Control System.
URI 50-327,328/96-02-04, was identified in connection with plant modification
DCN No. M11445A, Revision 0, that was developed and implemented for Unit
1 during cycle 7, refueling outage. The plant modification was intended to
address safety concerns described in NRC Generic Letter (GL) 93-04, Rod
Control System Failure and Withdrawal of Rod Control Cluster Assemblies,
The safety assessment performed for this plant modification was determined to
be technically inadequate. Specifically, the Safety Assessment Checklist,
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Appendix G, item 22, incorrectly stated that there were no new credible failure '
l- modes associated with the hardware change. This error led to omission of
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requirements from the DCN for development and implementation of
recommend surveillances described in WCAP-13864, Revision 1. TVA
l
management in their letter dated June 10,1990, committed to the corrective
l
action delineated in PER No. SQ960677 PER for developing a new procedure
to comply with GL 93-04 and the WOG recommendations. The action due
date for this corrective action is February 15,1997. Additionally, plant
modification DCN No. M11730A, has been revised to address the new failure
modes introduced by the hardware changes. Based on the corrective actions
completed by the licensee this URI is closed.
An apparent violation of 10 CFR 50 Appendix B, Criterion lil, will be identified
for failure to implement adequate design controls for " Rod Control System"
plant modification.
1
C. Exit j
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The inspection scope and results were summarized with those persons
indicated in paragraph D on November 22,1996 and December 19,1996. The
inspector described the areas inspected and discussed in detail the inspection
results. One unresolved item related to the technical acceptability of reducing
the free field beta dose inside the containment and annulus by 50 percent was l
identified; and one inspector followup item concerning inconsistent FSAR
description of the Reactor power was also identified. An Unresolved item in
connection with inadequate safety assessment of Rod Control System plant
modification was closed, and a violation of 10 CFR 50 Appendix B, Criterion lli
was opened.
On January 8,1997, in a telephone conversation, the licensee was informed
that three unresolved items related to PER No. SQP900372PER were made
unresolved items pending the results of a meeting with TVA. A date for the
meeting was not yet determined.
Licensee Emplovees
R. Adney, site Vice President
- B. Alsup, Quality Assessment Supervisor
J. Beasley, Site Quality Manager
- L. Bryant, Assistant Plant Manager
- G. Buchanan, Component Engineering Manager
C. Butcher, Electrical Design Manager
M. Burzynski, Engineering and Materials Manager
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l *R. Driscoll, Site Training Manager
M. Fecht, Nuclear Assurance and Licensing Manager
T. Flippo, Site Support Manger
- J. Herron, Plant Manager '
- C. Kent, Radchem Manager
- B. Lagergren, Operations Manager
- P. Leahy, Shift Manager, Operations
M. Lorek, Mechanical Engineering Manager
R. Newby, Concerns Resolution Staff, Manger
R. Norton, SON Assessment Supervisor
R. Profitt, Licensing Engineer
J. Rupert, Engineering and Service Support Manager
- R. Shell, Licensing and Industry Affairs Manager
J. Smith, Site Licensing Supervisor
- Attended exit interview on December 19,1996 only.
Insoection Procedures Used
IP 37550 Engineering
IP 37551 Onsite Engineering
items Ooened/ Closed / Discussed
Opened
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50-327,328/96-16-01 URI Inadequate safety evaluation resulted
in Unreviewed Safety Question.
(Paragraph E1)
50-327,328/96-16-02 URI Untimely corrective action for
nonconforming plant conditions.
(Paragraph E1)
50-327,328/96-16-03 URI Inadequate design control for
nonconforming plant conditions.
(Paragraph E1) l
50-327,328/96-16-04 URI Technical acceptability of reducing the i
calculated free field beta dose inside
containment and annulus. (paragraph
E1) l
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50-327,328/96-16-05 IFl FSAR inconsistent description of
reactor power level. (Paragraph E1)
50-327/96-16-06 VIO Inadequate Design Controls for Rod
Control System plant modification.
(Paragraph E.8)
,
Closed
URI 50-327,328/96-02-04 Omission of Surveillance Tests for Rod
Control System.
Acronyms -
CAQR Condition Adverse to Quality Report !
CFR Code of Federal Regulations l
DBDs Design Basis Documents
1
EEB Electrical Engineering Branch,
EFPD l
Effective Full Power Day -
EGTS Emergency Gas Treatment System
EQ Environmental Qualification !
FSAR Final Safety Analysis Report
JCO Justification for Continued Operation l
LOCA Loss of Coolant Accident
MWt Megawatts Thermal .
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NE Nuclear Engineering !
NF Nuclear Fuels '
NRC Nuclear Regulatory Commission
PER Problem Evaluation Report
PORV Power Operated Relief Valve
SER Safety Evaluation Report
TVA Tennessee Valley Authority
URI Unresolved item -
WOG Westinghouse Owners Group i
,