ML20149L064

Forwards Evaluations & Conclusions Re 970512 NOV from Insp Repts 50-327/97-03 & 50-328/97-03 Conducted at Facility.Nrc Completed Review of NOV Responses & Concluded That Violations Occurred as Stated

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.

President, TVA Nuclear and

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Chief Nuclear Officer

6A Lookout Place

1101 Market Street

Chattanooga. TN 37402-2801

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Dear Mr. Kingsley:

SUBJECT: NOTICE OF VIOLATION

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. ,

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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

License Nos. DPR-77, DPR-79

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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PDR ADOCK 05000327

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cc w/ encl:

O. J. Zeringue. Senior Vice President

Nuclear Operations

Tennessee Valley Authority

6A Lookout Place

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!

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-

Soddy Daisy, TN 37379 i

Michael H. Mobley. Director .

Division of Radiological Health  !

3rd Floor. L and C Annex .l

401 Church Street i

Nashville. TN 37243-1532

County Executive

Hamilton County Courthouse

Chattanooga. TN 37402

Distribution w/ encl:

J. R. Johnson. RII l

M. S. Lesser. RII l

S. E. Sparks. RII l

R. W.,Hernan. NRR  !

F Hebdon. NRR 1

H. L. Whitener. RII l

C. F. Smith, RII i

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

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NRC Resident Inspector. Operations

Sequoyah Nuclear Plant

~U. S. Nuclear Regulatory Commission

2600 Igou Ferry Road

Soddy-Daisy. TN 37379

/

OFFirF Pff 0P9 Rif OP9 Pff-ORD Rff-

SIGNATURE

NAMF C' th:pd HChristensen TLesser UryC

DATE 07 / / 3 / 97 01/ l3 / 97 07 / M i s7 02/$/97 07 / / 97 07 / / 97

copy? YES [ NO [Y[h NO YES NO Ifd$) NO YES NO YES NO

OFFICIAL KLLUKU LUPY MUMLNi NAM. 5:\Dkb\LB\ThA. W

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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 l

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 j

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 I

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 i

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

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Calculation TI-RPS-48. Revision 5. " Integrated Accident Dose

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.

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Additionally, plant modification DCN No. 508114A. Revision 16.

revised Environmental Drawings Number 1.2-47E235 Sheets 45, 47.

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and 48 to replace radiation values that were no longer

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

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operation. 1

Unit No. Cycle No. Date EFPD Exceeded

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1 4 12-29-89

1 5 06-09-91

2 3 12-30-88 l

2 4 05-24-90 l

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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

Annulus. 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 '

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 l

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

'and a justification for continued operation (JCO) was established. l

The licensee provided background information 'concerning reduction of fifty

percent in the free field beta dose. The licensee considered the action of j

not incorporating the results of calculation TI-RPS-48. Revision 3. into the  !

! ' design basis as prudent and within NRC regulations. The licensee's rationale  !

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for preparing Revision 5 of Calculation TI-RPS-48 was to ensure consistency .

between the design basis of 650 Effective Full Power Day (EFPD) average core i

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exposure and approved design output documents. The licensee discussed the

circumstances surrounding preparation of the JC0 using TID-14844 methodology 1

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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

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 i

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 i

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. i

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. i

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 i

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

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 ]

basis-, and Example 1 of the Violation is withdrawn. '

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. l

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 i

not in effect, and this was contrary to the guidelines of Generic Letter 91-  !

18. 1

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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