ML20149L064
| ML20149L064 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/23/1997 |
| From: | Jaudon J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Kinglsey O TENNESSEE VALLEY AUTHORITY |
| References | |
| 50-327-97-03, 50-327-97-3, 50-328-97-03, 50-328-97-3, NUDOCS 9707310087 | |
| Download: ML20149L064 (8) | |
See also: IR 05000327/1997003
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July 23. 1997
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. Tennessee Valley Authority
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ATTN: Mr. Oliver D. Kingsley, Jr.
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President, TVA Nuclear and
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Chief Nuclear Officer
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6A Lookout Place
1101 Market Street
Chattanooga. TN 37402-2801
Dear Mr. Kingsley:
SUBJECT:
NRC INSPECTION REPORT NOS, 50-327/97-03 AND 50-328/97-03
Thank you for your response of June 11, 1997, to our Notice of Violation
issued on May 12. 1997, concerning activities conducted at your Sequoyah
facility.
In your response you stated that although design input Calculation
TI-RPS-048 was revised. -the design basis was maintained and a justification
-for continued operation (JCO) was established.
You therefore denied that
Violation G existed.
,
,
After careful consideration of the basis for your denial of Violation G, we
have concluded, for the reasons presented in the enclosure to this letter.
that the violation occurred as stated in the Notice of Violation with the
exce) tion of Example 1 of the Violation. Therefore in accordance with
10 C:R 2.201(a), please submit.tg this office within'30 days of the date of
this letter a written statement describing steps which have been taken to
correct Violation G. Examples 2 and 3 and the results achieved, corrective
steps which will be taken to avoid further violations, and the date when full
compliance will be achieved.
.We appreciate your cooperation in this matter.
Sincerely,
Original signed by
Johns P. Jaudon
Johns P. Jaudon, Director
Division of Reactor Safety
Docket Nos. 50-327. 50-328
Enclosure:
Evaluations and Conclusions
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cc w/ enc 1:
(See page 2)
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9707310087 970723
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ADOCK 05000327
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cc w/ encl:
O. J. Zeringue. Senior Vice President
Nuclear Operations
Tennessee Valley Authority
6A Lookout Place
.
1101 Market Street
!
Chattanooga, TN 37402-2801
Jack A. Bailey. Vice President
Engineering & Technical Services
Tennessee Valley Authority
6A Lookout Place
1101 Market Street
Chattanooga. TN 37402-2801
R. J. Adney
Site Vice President
Sequoyah Nuclear Plant
Tennessee Valley Authority
P. O. Box 2000
Soddy-Daisy. TN 37379
General Counsel
Tennessee Valley Authority
ET 10H
400 West Summit Hill Drive
Knoxville. TN 37902
Raul R. Baron General. Manager
Nuclear Assurance and Licensing
Tennessee Valley Authority
-4J Blue Ridge
1101 Market Street
!
Chattanooga. TN
37402-2801
Pedro Salas. Manager
Licensing and Industry Affairs
Tennessee Valley Authority
4J Blue Ridge
1101 Market Street
Chattanooga. TN 37402-2801
Ralph H. Shell. Manager
Licensing and Industry Affairs
l
Sequoyah Nuclear Plant
l
P
O. Box 2000
Soddy-Daisy. TN 37379
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(cc w/ encl cont'd - See page 3)
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(cc w/ enc 1 cont'd)
J. T. Herron. Plant Manager
,
Sequoyah Nuclear Plant
Tennessee Valley Authority
!
P. O. Box 2000-
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Soddy Daisy, TN 37379
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Michael H. Mobley. Director
.
Division of Radiological Health
.
3rd Floor. L and C Annex
401 Church Street
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Nashville. TN 37243-1532
County Executive
Hamilton County Courthouse
Chattanooga. TN 37402
Distribution w/ encl:
J. R. Johnson. RII
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M. S. Lesser. RII
S. E. Sparks. RII
R. W.,Hernan. NRR
F Hebdon. NRR
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H. L. Whitener. RII
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C. F. Smith, RII
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E. D. Testa. RII
-D. H. Thompson. RII
J. H. Moorman.-RII
PUBLIC
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NRC Resident Inspector. Operations
U. S. Nuclear Regulatory Commission
1260 Nuclear Plant Road
<
Spring City. TN 37381
,
NRC Resident Inspector. Operations
Sequoyah Nuclear Plant
~U. S. Nuclear Regulatory Commission
2600 Igou Ferry Road
Soddy-Daisy. TN 37379
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. EVALUATIONS AND CONCLUSIONS
On May 12. 1997. A Notice of Violation (Notice) was issued for a violation
identified during a routine NRC inspection.
TVA responded to the Notice on
June 11, 1997.
In the response. TVA denied that the violation occurred as
stated in the NDV.
The NRC1s evaluation and conclusion regarding the
licensee's responses are as follows:
Restatement of Violation
G.
10 CFR 50. Appendix B. Criterion III. Design Control, requires in part
that measures be established to assure that applicable regulatory
requirements are correctly translated into drawings and procedures.
The
measures shall include provisions to assure that appropriate quality
standards are specified and included in design documents.
The design
control measures shall also provide for verifying or checking the
adequacy of design.
Tennessee Valley Authority Nuclear Quality Assurance Plan TVA-NOA-
PLN89-A. Revision 6. Section 7.0. Design Control, requires that measures
be established to ensure that the performance of design analysis shall
be planned and controlled. Additionally, it requires that measures to
control plant configuration and ensure that the actual
)lant
configuration is accurately depicted on drawings and otler appropriate
design output documents and reconciled with the applicable design basis
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shall be established, documented, and implemented.
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TVA-NOA-PLN89-A. through Section 7.0 and Appendix B. endorses the
requirements of ANSI N45.2.11-1974. Quality Assurance Requirements for
the Design of Nuclear Power Plants.
Section 4.0 of this standard
requires that design analyses shall be performed in a planned,
controlled. and correct manner.
Design analyses shall also be in a form
suitable for reproduction, filing and retrieving.
Contrary to the above the established design control measures were
deficient in that the following deficiencies were identified:
1.
As of July 30, 1990. radiation dose values contained in design
basis Calculation TI-RPS-48. Integrated Accident Dose Inside
Primary Containment and Annulus. Revision 3. were never
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incorporated in Calculation TI-ECS-55. Summary of Harsh
Environment Conditions for Sequoyah Nuclear plant.'to ensure
revision of environmental data drawing series number 47E235.
Additionally. FSAR Figures 3.11.2-1. and 3.11.2-2 were never
revised to reflect the new 100-day integrated accident doses based
on a source term of 1000 EFPD.
This failure to control plant
configuration and ensure that actual plant configuration was
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accurately depicted on drawings resulted in discrepancies in
design basis information listed in FSAR Table 15.1.7-1 and FSAR
Figures 3.11.2-1 and 3.11.2-2.
2.
On December 12. 1991. TVA management approved design basis
Calculation TI-RPS-48. Revision 5. " Integrated Accident Dose
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Inside of Primary Containment and Annulus." to document the
100-day integrated beta and gamma radiation doses based on a
source term of 650 EFPD.
Radiation dose values contained in this
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calculation were incorporated into Calculation TI-ECS-55. " Summary
of Harsh Environment Conditions for Sequoyah Nuclear Plant.
Additionally, plant modification DCN No. 508114A. Revision 16.
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revised Environmental Drawings Number 1.2-47E235 Sheets 45, 47.
and 48 to replace radiation values that were no longer
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conservative.
These drawing revisions did not accurately depict
actual plant configuration in that on the following dates listed
the core average exposure for both units exceeded 650 EFPD
operation.
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Unit No.
Cycle No.
Date EFPD Exceeded
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1
4
12-29-89
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5
06-09-91
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3
12-30-88
2
4
05-24-90
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2
5
09-28-91
This failure to control plant configuration and ensure that actual
plant configuration was accurately depicted on drawings resulted
in discrepancies between the units' current licensing basis of
1000 EFPD burnup criterion and approved design basis information
depicted on the environmental drawings.
3.
From February ll. 1994. to November 15, 1996, the licensee failed
to perform a calculation to determine the integrated maximum
hypothetical accident gamma and beta doses inside the primary
containment to support a justification for continued operation for
50 PER-900372 PER
This is a Severity Level IV Violation (Supplement I).
Summary of Licensee's Violation Resoonse
The licensee denied that a violation of 10 CFR 50 Appendix B. Criterion III
existed.
The licensee considered that design control was maintained following issuance
of Calculation TI-RPS-48. Integrated Accident Dose Inside Containment and
The licensee did not identify. however, the revision level of the
calculation for which they were taking credit.
The licensee stated that plant
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design basis was controlled by design output documents such as drawings.
setpoint and scaling documents, and general construction specifications.
The licensee also stated that calculations and design input documents such as
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design criteria and design standards support the design output documents. The
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licensee believed that design basis changes did not occur until the design
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output documents are changed. regardless of changes to design input documents
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or calculations.
It is the licensee's position that although the design input
document (Calculation TI-RPS-48) was revised. .the design basis was maintained
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'and a justification for continued operation (JCO) was established.
The licensee provided background information 'concerning reduction of fifty
percent in the free field beta dose. The licensee considered the action of
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not incorporating the results of calculation TI-RPS-48. Revision 3. into the
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' design basis as prudent and within NRC regulations. The licensee's rationale
for preparing Revision 5 of Calculation TI-RPS-48 was to ensure consistency
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between the design basis of 650 Effective Full Power Day (EFPD) average core
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exposure and approved design output documents. The licensee discussed the
circumstances surrounding preparation of the JC0 using TID-14844 methodology
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and stated that the calculation to support the JC0 was performed using a OA
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controlled computer program. The licensee took credit for the revised JC0
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which was submitted to the NRC on March 4. 1994, and which was approved by the
NRC in their letter to TVA dated April 8.1994.
NRC Evaluation
-The NRC staff has carefully reviewed the licensee's response and has concluded
that the licensee did not provide any additional information that was not
already considered in determining the significance of the violation.
10 CFR 50.2 defines " Design Bases" to mean that information which identifies
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the specific functions to be performed by a structure, system or component of
a facility, and the specific values or ranges of values chosen for controlling
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parameters as reference bounds for design.
These values may be (1) restraints
derived from generally accepted " state of the art" practices for achieving
functional goals or (2) requirements derived from analysis (based on
calculations and/or experiments) of the effects of a postulated accident for
which a structure, system, or component must meet its functional goal.
The
licensee's Nuclear Quality Assurance Plan. TVA-NQA-PLN89-A. Revision 6.
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requires design controls to be established and implemented to ensure that
applicable specified design recuirements, such as design bases. are correctly
translated into design output crawings. Additionally, the design change
process requires that actual plant configuration shall be accurately depicted
on drawings and shall be reconciled with the applicable design bases.
The concept of configuration management and adequate design controls involve
first establishing a baseline where )lant design output documents are
consistent with well defined design ]ases and where design output documents
represent actual as-built plant configuration.
Having established that
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baseline. the design char-e process shall assure consistency between design
bases, design output documents, and as-built plant configuration for all plant
modi fications.
On September 12. 1990, the licensee discovered that nuclear fuel related
design changes had been made which had not been reconciled with nor reflected
in Nuclear Engineering design basis documents.
Calculation TI-RPS-48.
Revision 3. calculated the integrated maximum hypothetical accident gamma and
beta doses inside the primary containment based on average core exposure of
1000 EFPD. The licensee stated that the result of this calculation was
unrealistic and was never incorporated into the design output documents. The
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licensee's concern was that the calculation did not account for a reduction of
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fifty percent in the free field beta dose.
This concern was reviewed by NRR.
and it was determined that a reduction of fifty percent in the free field beta
dose was technically acceptable.
On this basis we agree with the licensee's
actions of not incorporating Revision 3 of the calculation within the design
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basis-, and Example 1 of the Violation is withdrawn.
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Calculation TI-RPS-48. Revision 5, was issued on December 6. 1991, to return
the calculation to the original design basis condition of 650 EFPD average
core exposure.
Prior to this Unit 1 cycles 4 and 5. and Unit 2. cycles 3, 4
and 5 had already exceeded the licensing basis of 650 EFPD and the approved
design output documents no longer accurately reflected as-built plant
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configuration. On September 6. 1991, a JC0 was prepared to document the
effects of an increase in fuel burnup which had not been addressed in the E0
prograa. This JC0 was determined to be technically inadequate by the NRC and
was revised based on NRC's review and comments.
From September 12, 1990.
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until April 8.1994, the licensee's configuration control of the Environmental
Qualification program failed to meet the requirements of their Nuclear Quality
Assurance Plan. Additionally, during this time a technically adequate JC0 was
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not in effect, and this was contrary to the guidelines of Generic Letter 91-
18.
The licensee's statement that design basis changes do not occur until the
design output documents are changed, regardless of changes to design input
documents or calculations, is incorrect.
The definition of design basis in
10 CFR 50.2 does not support this argument. Additionally. the concept of good
configuration management and adequate design controls was not evident from the
licensee's actions for the Violation. Example 2.
The licensee is required to
implement adequate design controls for plant modificaticns which ensure that
actual plant configuration is accurately depicted on drawings and is
reconciled with design basis information as defined in 10 CFR 50.2.
The licensee stated that calculations to support the JC0 were performed using
a OA controlled computer program.
The licensee was unable. however, to
present objective evidence which corroborated this claim and which
demonstrated that the analysis had been performed in accordance with the
requirements of ANSI N45.2.11-1974.
On November 22, 1996. Calculation
SONSOS2-0163 was prepared with the specific purpose of validating the JC0 that
was submitted to the NRC on March 4. 1994.
This action was taken when
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computer runs performed by the licensee did not provide useful data in support
of the radiation values delineated in this JCO.
NRC Conclusion
For the above stated reasons for Examples 2 and 3 the NRC staff concludes that
the Violation occurred as stated and Violation Example 1 is withdrawn.
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