ML20211K563

Amend 220 to License DPR-77,authorizing Operation at Reactor Core Power Levels Not in Excess of 3,411 Megawatts Thermal

ML20211K563
Person / Time
Site:	Sequoyah
Issue date:	03/04/1997
From:	NRC
To:
Shared Package
ML20198J149	List: IA-97-261, Amend 220 to License DPR-77,authorizing Operation at Reactor Core Power Levels Not in Excess of 3,411 Megawatts Thermal ML20198J145 ML20211K533 ML20211K563
References
FOIA-97-261 NUDOCS 9710090277
Download: ML20211K563 (2)

Text

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• NUCLE AR REGULATORY COMMISSION wAs.usec toes, p. c. poses t

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' TEkNESSEE V4LLEY AUTHORITY DOCKET NO. 50 327 -

SEQUDYAll NUCLEAR PLANT, UNIT 1 _

FACILITY OPERATING LICENSE l .

License No. OPR-77 1., The Nuclear Regulatory Comission (the Comission) having found that:

A.

The application for licenses filed by the Tennessee Valley Authority complies with the standards and requirements of the Atomic Energy Act '

of 1954. as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter 1. and all required notifications to other agencies or bodies have been duly made; B.

Construction of the Sequoyah Nuclear Plant. Unit 1 (the facility

~has been substantially completed in conformity with Provisional ).

Construction Pemit No. CPPR.72 and the application, as amended, the provisions of the Act and the regylations of the Comission; C.

The facility will operate in conformity with'the application, as '

amended, the provisions of the Act, and the regulations of the Commission; .

D.

There is reasonable assurance: (1) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in conpliance with the regulations of the Comission set forth in 10 CFR Chapter i; E.

The Tennessee Valley Authority is technically and financially qualified to engage in the activities authorized by this operating license in accordance with the Comission's regulations set forth in 10 CFR Chapter 1; F.

The Tennessee Valley Authority has satisfied the applicable provisions of 10 CFR Part 140.

• Financial Protection Requirements and Indemnity Agreements", of the Comission's regulations; G.

The issuance of this license will not be inimical to the comon defense and security or to the health and safety of the public; 9710090277 971007 '

• PDR FOIA EUBANKS97-261 PDR

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(4) Pursuant to the Act and 10 CFR Parts 30,40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuc! car material without restriction to chemical or physical form, for sample analysis or instrument cahbration or associated with radioactive apparatus or components; and ,

(5) Pursuant to the Act and 10 CFR Parts 30,40 and 70, to possess, but not I separate, such byproduct and special nuclear materials as may be produced by the operation of the facility,  ;

1 C. This license shall be deemed to contain and is subject to the conditions  !

specified in the Commission's regulations set forth in 10 CFR Chapter I and is l subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the add,tional conditions specified or incorporated below:

(1) Mag nJm Power Level The Tennessee Valley Authonty is authorized to operate the facility at reactor core power love!s not in excess of 3411 megawatts thermal,  ;

(2) Technical SDecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 220 are hereby incorporated into the license. The R224 licensee shall operate the facility in accordance with the Technical Specifications.

(3) initiel %t Prooram The Tennessee Valley Authonty shall conduct the post fuel-loading initial test program (set forth in Section 14 of Tennessee Valley Authority's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received pnor NRC approval. Major modifications are defined as:

a. Elimination of any test identified in Section 14 of TVA's Final Safety Analysis Report as amended as being essential; I b. Modificatien of test objectives, methods or acceptance entena for any test identrfied in Section 14 of TVA's Final Safety Analysis Report as amended as being essential;

c. Performance of any test at a power level different from there desenbed, and March 4,1996 Amendment No 220

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Evaluation Iteport n.,oi..,Ji%"l't;;-

"*d * *'"'"" ' n.in':;d,"Ji: lit; Sequoyah Nuclear Plant oocu.,so Q .

so 32s Units 1 and 2 Tennessee Valiay Authority t#

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stainless steels, constitutes an acceptable balls for steting the requirements of General Design Criterion 26 of Appendis A to 10 Cf R Part 50.

The design precedures and criteria that the applicant has used for the reactor i

internals are in conformance with tit #blished technical peorecures, positions, 9' standards, and criteria as cited above which are acceptable to the staf f.

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4. ) Nuclear Desion 1 ej 6 l 1he nuclear design of Sequoyah Units 1 and 2 is the same as that of frnjan and Salem i Unit 1. These reactors have been previously reviewed and approved. Sequoyah has

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rated power of 3411 thermal megawatts and consists of 193 assemb!les containing Cl the Westinghouse 17:17 rod fuel assembly array. Our review was based on informa-tien supplied by the applicant in the final Safety Analysis Report and es.endments h I.' '

thereto, and referenced topical reports. Ou review was conducted within the f guidelines provided by the Standard Review ttan, Section 4.3.

I 4.3.1 i

yO f in Bases Design bases are presented which comply with the applicable General Design criterf s. I i fuel cenign limits are specified which meet the requirements of Coeral Design i Criterton 10. A negative prompt feedt.ack coef ficient is required which satisfies

., il General Design Criterion ll, and po er oscillation is required either to be not

,) l possible or to be detected and supprelled by the control system, which satisfies '

,i General Design Criterion 12. A monitoring and control system is provided which i!

i aulcmatically initiates a rapid reactivity insertion to prevent eacteding fuel .

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design limits in normal operation and anticipated transients. This satisfies i!

Ceneral Design Criteria 13 and 20.

The control systee is designed so that no

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single failure or single operator error will cause a vlotation of fuel design limits and 50 that shutdown is assured even when the single rod cluster control h i i4 assembly (control rod) of highest worth is assumed to be stuck out of the core.

Fyrther a criemical shim system is provided unich is capable of controlling normal '

power changes and bringing the reactor to cold shutdown, the control system, when combined with the engineered safety features, is required to control reactivity '

changes during accident conditions, ce activity insertion rates and amovats are $ I )

I f controlled 50 that lietted damage occurs to the pressure boundary and the core

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stays in coolable geometry. The reactivity control system meets the requirements '

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of General Oesign Criteria 25, 26, ?? and 28 On the basis of the above. we find 6

the design bases presented in the final Safety Analysis Report to be acceptable II ii Ceston Description N

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The final $afety Analysis Report contales the description of the first cycle f uel ,*

F loading which consists of three dif ferent enrichments and nas a first cycle of  ;.,

appronmately one year. The enrichment distribution, buenable poison distribution.

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1. j asae9 e, UNITED STATES f NUCLEAR REGULATORY COMMISSION t nEoioN H

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(**"* / A January 13,1997 Tennessee Valley Authority ATTN: Mr. Oliver D. Kingsley, Jr.

President, TVA Nuc! car and Chief Nuclear Officer SA Lookout Place 1101 Market Street Chattanooga, TN 37402 2801

SUBJECT:

NOTICE OF VIOLATION (NttC INSPECTION REPORT NO. 50 327/9616 AND 50-328/96-16)

Dear Mr. kingsley:

An NRC inspection was conducted on September 23 27,1996, November 4 22,1996, and December 16-19,1996, at your Sequoyah facility. The purpose of the inspection was to determine whether activities authorized by the licenso were conducted safely and in accordance with NRC requirements. At the conclusion of the inspection the findings were discussed with those members of your staff identified in the enclosed report.

Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel,and observation of activities in progress.

Based on the results of this inspection, certain of your activities appeared to be in '

violation of NRC requirements, as specified in the enclosed Notice of Violation (Notice). The violation is of concern because it is indicative of inadequate implementation of your design control program. Four unresolved items were also identified in connection with the use of high burnup fuel having average core exposure of 1000 Effective Full Power Days (EFPD). We,are requesting a meeting with TVA to obtain additional information for resolution of these unresolved items.

The responses directed by this letter and the enclosed Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Pub. L. No. 96 511.

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TVA 2 In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice " a copy of this letter and its enclosures will be placed in the NRC Public Document Room.

1 Sincerely.

Original signed by Charles A. Casto l Charles A. Casto, Chief Engineering Branch Division of Reactor Safety Docket Nos. 50 327, 50 328 Licenso Nos. OPR 77, DPR-79

Enclosures:

1. Notice of Violation

2. NRC Inspection Report f cc w/encls:

O. J. Zeringue, Senior VP General Counsel Nuclear Operations Tennessee Valley Authority Tennessee Valley Authority ET 10H 6A Lookout Place 400 West Summit Hill Drive 1101 Market Street Knoxville, TN 37902 Chattanooga, TN 37402 2801 Raul R. Baron, General Manager Dr. Mark O. Medford, VP Nuclear Assurance and Licensing Technical Services Tennessee Valley Authority Tennessee Valley Authority 4J Blue Ridge 6A Lookout Place 1101 Market Street 1101 Market Street Chattanooga, TN 37402 2801 Chattanooga, TN 37402 2801 Pedro Salas, Manager R.J.Adney Licensing and Industry Affairs Site Vice President Tennessee Valley Authority Sequo/ab Nuclear Plant 4J Blue Ridge Tennessee Valley Authority 1101 Market Street P. O. Box 2000 Chattanooga, TN 37402-2801 Soddy Daisy, TN 37379 (cc w/encls cont'd - See page 3)

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O TVA 3  ;

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(ce w/encls cont'd)

Flalph H. Shell, Manager l Licensing and industry Affairs Sequoyah Nuclear Plant ,

P. O. Box 2000 '

Soddy Daisy TN 37379 J. T. Herron, Plant Manager Sequoyah Nuclear Plant Tennessee Valley Authority P. O. Box 2000 -

Soddy Dalsy, TN 37379 Michael H. Mobley, Director Division of Radiological Health 3rd Floor, L and C Annex 401 Church Street Nashville, TN 37243 1532 County Executive Hamilton County Courthouse Chattanooga, TN '37402

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• TVA 4 Distribution w/encls:

E. W. Merschoff, Rll M. S. Lesser, Ril S. E. Sparks, Ril R. W. Hernan, NRR F. Hebdon, NRR H. L. Whitener, Ril C. F. Smith, Ril E. D. Testa, Ril D. H. Thompson, Ril J. H Moorman, Rll P. Steiner, Ril PUBLIC NRC Resident inspector, Operations U. S. Nuclear Regulatory Commission

1. 1260 Nuclear Plant Road Spring City, TN 37381 NRC Resident inspector Sequoyah Nuclear Plant U. S. Nuclear Regulatory Cornmission

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NOTICE OF VIOLATION Tennessee Valley Authority Docket Nos. 50 327 and 50 328 Sequoyah Nuclear Plant License Nos. DPR 77 and .

DPR 79 During NRC inspections conducted on September 23 27. 1996. November 4 22.

1996. and December 16 19. 1996. a violatton of NRC requirement was identified.

In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions." NUREG 1600, the violation 15 listed below:

10 CFR 50. Appendix B. Criterion ill requires that measures shall be established to ensure that desiga activities shall be prescribed and i

accomplished in accordance with procedures of a type sufficient to  !

assure that applicable design inputs are correctly translated into specifications, drawings, procedures, or instructions. Applicable design arcu'.! such as design bases, regulatory requirements, codes and ,

standards shall e identified. documented. and their selection reviewed i and approvad. The design input shall be 5)ecified on a timely basis and i to a level 7f deta.1 necessary to permit tie design activity to be '

carried out in a :.orrect manner and to provide a consistent basis for making design decisions. 3ccomplishing design verification measures. and evaluating design changes.

Contrary to the above plant modtfication DCN No. M11730A. Revision 0, was approved on November 30. 1995, for modtfying Unit 1 Rod Control System without incorporating applicable design inputs concerning new failure modes introduced by the hardware modification described in Westinghouse Topical Report WCAP 13864. Section 3.5. This failure ,

resulted in the plant modification package omitting requirements for  !

development and implementation of new surveillance tests required to j determine any component failure which 1s undetectable during normal operation.

This is a Severity Level IV Violation (Supplement 1) i Pursuant to the previsions of 10 CFR 2.201. Tennessee Valley Authority is I hereby required to >ubmit a written statement or explanation to the U.S.

Nuclear Regulatory Commission. ATTN- Document Control Desk. Washington. D.C.

20555 with a copy to the Regional Administrator. Region !!. and a copy to the NRC Resident inspector. Sequoyah Nuclear Plant, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a " Reply to a Notice of Viola. tion" and should include for each violation: (1) the reason for the violation, or. If contested, the basis for disputing the violation. (2) the ccrrective steps that have been taken and the results achieved. (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. i Your response mcy reference or include previous docketed correspondence. If the correspondence adequately 6ddresses the required response. If an adequate i'

Enclosure 1

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1 Notice of Violation 2 reply is not received within the time specified in this Notice, an order or l Demand for Information may be issued as to why the license should not be modifica. suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown. consideration will se given to extending the response time.

Because your response will be placed in the NRC Pubic Document Room (PDR), to the extent possi)le, it should not include any personal privacy. 3roprietary, or safeguards information so that it can be placed in the POR wit 1out redaction. However, if you' find it necessary to include such information, you should clearly indicate the specific information that you desire not to be I placed in the PDR. and provide the legal basis to support your request for withholding the information from the public.

Dated at Atlanta, Georgia this 13th day of January 1997 l

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U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket Nos: 50 327. 50 328 License Nos: DPR 77. DPR 79 Report No: 50 327/96 16, 50 328/96 16  !

Licensee: TVA Facility: Sequoyah Units 1 & 2 Location: Sequoyah Access Road Hamtiton County. TN 37379 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 Enclosure 2 ogr y ,- nd IrnY

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EXECUTIVE

SUMMARY

Sequoyah Nuclear Plant. Units 1 & 2 NRC Inspection Report 50 327.328/96 16 This special ins)ection included detail reviews of corrective actions implemented for 3roblem Evaluation Reports (PERs) No. 50P900372PER. Nuclear fuel Design Changes Not Reconciled / Reflected in Design Basis Documents (DBDs):

and 50950021PER. Obtain Operability Evaluation for 50NP: Review of WBPER940576. Additionally, a review of the licensee's transition plans for implementing the E0 program after Phase 1 site engineering re organization had been completed was performed. An Unresolved item involving inadecuate safety assessment of Rod Control System plant modification was closed anc a Violation of 10 CFR 50 Appendix B. Criterton 111 was cited.

Results:

e An unresolved item concerrang performance of an inadecuate 10 CFR

' 50.59 Safety Evaluation that resulted in an Unreviewec Safety Ouestion, o An unresolved item concerning untimely revision to the E0 Binders and EEB calculations.

o An unresolved item concerning inadequate design control for

" Nonconforming Plant Conditions."

e An unresolved item concerning the technical acce)tability of reducing the calculated fr ee field beta dose bot 1 inside containment and the annulus by 50 percent.

e An inspector followup item concerning inconsistent FSAR descriptions of the reactor power level, e

A violation for inadequate design controls for Rod Control System plant modification.

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Report Details l

l 111. Enaineerina 1

El Conduct of Engineering El.1 PER No. 50P900372PER. Nuclear Fuel Design Changes not Reconciled' Reflected in Design Basis Documents (DBDs)

a. Inspection Scong l

The inspector reviewed PER No. 50P900372PER 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.

b. Observations and Find 1,n.si Condition Adverse to Quality Report (CAOR) SOP 900372PER, dated Seatember 18, 1990, documented fuel related design changes made by TVA whic1 had not been reconciled or reflected in design basis documents. An increase in the average core burnup from 650 EFPD to 1000 EFP0 resulted in an increase in the amount of core activity that is assumed at the start of a design basis LOCA.

Because of tnis there was an increase in the 100

{ day 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 91293) to I

demonstrate that the requirements of 10 CFR 50.49 were still being met by equipment that had )reviously been environmentally qualified based on a source term of 650 E:PO. The ins)ector reviewed the JC0 and determined that TVA had concluded t1at the JC0 bounded reactor core designs with U235 fuel having average enrichment less than 4.5 percent

, 1000 EFP0 burnup, l

l The NRC in a letter dated November 30. 1993.

Subject:

Evaluation of

' Increased Fuel Burnup on Equipment Qualification Sequoyah Nuclear Plant Unit I and 2. transmitted the results of the staff's review of the above JC0 to TVA. The staff concluded that the JC0 was not appropriate and TVA was requested to perform a reassessment of equipment qualification for 1000 EFPD burnup using an acceptable source term (T10 14844) and resubmit the JC0 for the staff's review. TVA performed the reassessment and in a letter dated March 4.1994. transmitted the JC0 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 concern The inspector reviewed the results of the E0 reassessment titled " Review of 1000 EFPD with 4.5% U235 Enrichment" performed in support of the JC0 submitted to the NRC. Corrective action plans developed and implemented for CAOR No. 50F8700l?. and 50P870165 were also reviewed during this inspection. The specific issues reviewed and the results of these reviews are discussed in the paragraphs below

2 j

1 Technical Adeuuacy of 10 CFR 50 !9 Safety Evaluation CAOR No. 50F870012 was written on March 19. 1987. to document a condition where the core average exposure limit of 26154 MAD /HTV specified 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 (EFPO) and revise the FSAR to reflect the results of the revised calculation. CA0R No. 50P870165 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 to pressure changes was due to the use of a pressure indicating controller having only a pro)ortional band with no reset function. The inspectors reviewed a 10 CFl 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 .evised: 1) Table 15.1.7 1. Core and 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 Of fsite Doses from 1.oss 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 CA0R Nos.

SOF870012 and 50P870165.

FSAR Chapter 15. Table 15.1.71 was revised to show new source terms based on 1000 EFPD operation. The results of offsite dose calculations performed by the NRC in sup) ort of licensing actions were documented in Safety Evaluation Report (SER) Supplement No.1. dated February 1980.

The 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.l.

Table 15 1. Radiological Consequences of Design Basis Accidents, was calculated by the NRC based on the assumption that Unit I 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, Assum)tions Used in the Calculation of

, Loss of Coolant Accident Doses, also slowed the reactor power level as 3582 M4 thermal. This value of reactor power level used in the offsitel dosecalculationwasdifferentfromthatusedbyTVAwhichwas3565MWg thermal. The guidance delineated in T10 14844 Calculation of Distance ,

Factors for Pow r and Test Reactor sites, dated March 23.1962. requires the use of the reactor rated power level (megawatts) in the calculation l which determines the radio nuclide inventory of 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 listed as 3411MW thermal. In Table 15.1.7 1 it was listed as 3565 MW thermal and in all the tables of * ' Section 15.5 it was listed as 3582

'MW thermal. The maximum power le authorized in the facility operating license is 3411 MW the..nal. This inconsistency in FSAR t

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V description of the reactor power level is identified as IFl 50-327.328/9610i05.FSARInconsistentDescriptionofReactorPowerLevel.

The results of the above reviews demonstrated that the licensee had i considered the consequences of offsite radiation doses to the health and safety of the public based on 1000 EFFD operation. Additional reviews  :

of the 10 CFR 50.59 Safety Evaluation, however, revealed that the

- licensee had not evaluated whether the incNose 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 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 qualifled equipment must withstand, The licensing basis for the 10 CFR 50.59 E0 Program was 650

• EFPD burnup and this requirement was exceeded by Unit 1 cycle 4 o>eration on December 29, 1989 and Unit 2 cycle 3 on December 30. 1988.

111s "Unreviewed Safety Question" involving failure of the 10 CFR 50.59 Safety Evaluation to address the requirements of environmentally qualified equipment 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 16 01. Inadequate Safety Evaluation Resulted in Unreviewed Safety Question.

Corrective Action, Imolemented for Nonconformina Plant Canditions Problem Evaluation Report PER No. 50P900372PER 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 current Nuclear Engineering (NE) design basis documents. A "Cause Analysis" was )erformed for this deficiency and the apparent cause was determined to >e lack of procedural controls to ensure adequate inter' 'e reviews and appropriate funding for those reviews. Corrective actic ' ms developed and 1mpiemented for recurrence control included:

.. Revising Cornorate Standard 9.2 for core alterations and core hardware changes to ensure adequate interface reviews

- and appropriate funding for these reviews.

2. Establishing requirements for NE to provide NF a list of fuel and core related parameters which affect engineering calculations and require review on a cycle specific basis.

3, Revising NF instruction 3.0 to ensure that other design basis documents impacted by core component design changes were addressed.

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4 Other corrective actions which required revising tha.: E0 Binders and Electrical Engineering Branch (EEB) calculations to incor) orate updated environmental conditions were delayed and transferred to ?ER No.

SO9400*'Jll. TROI action item No. 30. The inspectors reviewed a copy of e TROI Action item No. 36 dated September 12 1996, and verified that this item was still open. ,

The licensee prepared a JC0 dated September 4. 1991, which was applicable to Doth 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, g TVA's JC0 was based on the conclusions contained in a document titled

" Tennessee Valley Authority. Sequoyah Nuclear Plants Units 1 and 2, increaseinthe100DayIntegratedDosetoEguipmentinContainment Associated with increased Fuel Burnup, Justi.ication for Continued Operation." The JC0 stated that TVA will reevaluate SONP design basis following the NRC's 'inal issuance of the new TID 14844 values in order to eliminate repetitive efforts of revising the E0 Binders.

On July 18. 1992. TVA management prepared JC0 No. 50JC092 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 en August 6. 1992. On September 17. 1993. JC0 for 50P900372PER was extended by a corrective action request. The corrective action request was approved by TVA management on September 20. 1993.

Prior to preparation of the JCOs and during the intervals of time when TVA management postponed implementing the corrective action to revise the E0 Binders and EEB cdiculations, the core average exposure for both Units exceeded 650 EFPD operation on the dates listed:

Unit NO. Cycle No. Date EFPD Exceeded 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 t,0vember 30. 1993, the NRC transmitted the results of their review of the " Westinghouse Technical JC0 for SONP" to TVA. TVA was informed that the JC0 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 response to this request on February 11. 1994. TVA prepared "JC0 for PER No. SOP 900372PER" which bounds reactor core l

5 designs with U235 average enrichment of less than 4.5% and 1000 EFPD.

This JC0 included Unit 2 cycle 6. Unit I cycle 7. and Unit 2 cycle 7 fuel cycle operation. The NRC reviewed "JC0 for PER No.

SOP 900372PER'and concluded that TVA's equi)me.nt qualification reassessment satisfactorily responded to t1eir 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 JCD without revising the E0 Binders and EE8 calculations. This untimely corrective action for revising the EQ l g 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.

Desion Control imolemented for Nonconformino 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 analysi:; 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 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 It 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 wt.en compared to the 100 day integrated accident doses in revision 2 were as follows:

Location TI-RPS-48. R2 TI RPS-48-R3 Upper Containment Gamma 3.8 E7 3.0 E7 Beta 4.7 E8 8.3 E8 Instrument Rooms Gama 1.048 E7 1.6 E7 Beta 4.7 E8 8.3 E8 Lower Containment Gamma 2.8 E7 2 5 E7 Beta 4.7 E8 8.3 E8

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6 l

Accumulator and Fan Rooms Gama 1.W 8 E7 1.6 E7 Beta 4.7 E8 8.3 EB Raceway Gama 1.048 E7 2.4 E7 l Beta 4.7 E8 8.3 E8 Ice Condenser Bed Gama 1.34 E7 2.3 E7 8 eta 4.7 E8 8.3 E8 Annulus Ga m 1.3 E7 5.9 E6 Beta 5.0 E5 1.38 E6 i

A significant increase in free field Beta radiation resulted from the 1000 EFPD burnup criteria. The results of these calculations sere never incorporated in Calculation TI ECS 55. Sumary of Harsh Environment Conditions for Sequoyah Nuclear Plant. As a consequence the environmental data drawings series Number 47E235 were never revised to reflect the integrated accident doses caused by the new source terms based on 1000 EFPD operation.

Additionally. FSAR Figures 3.11.2-1 and 3.11.2-2 were never revised to reflect 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 T1 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 I

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 l environmental drawings. This event was preceded by Unit I cycle 4 and Unit 2 cycles 3. 4 and 5 having averag? core exposure in excees of 650

7 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 "JC0 for PER No. 50P900372." dated February 11. 1994, to the NRC for their review. One hundred day integrated gammc. 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.

3 The inspectors reviewed the JC0 and determined that the radiation values delineated in the JC0 were not supported by an approved analysis. A g formal calculation had never been prepared reviewed and approved to determ le 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 requirements of the design control program which requires that design analyses shall be performed in a planned. controlled, and correct manner. In response to the inspector's concern TVA attempted to reconstitute the analysis 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 finall November 15. 1996 to address the inspector'y sprepared concern. and approved The results on of this calculation were reviewed by the inspcctors 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 AcceDtability of Reducino Calculated Free Field Beta Dos,q by 50 percent Design Calculation SON TI RPS 048. Revision 6 1ssued 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 burr.ups 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 Rev!sion 2 of the calculation at 650 EFPD and 3565 MWt. using TID 14844 snurce 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 quali fied it includes guidance for determining the radiation i

~'

~,.

8 environments expected to occur during a design basis event condition.

In Se: tion 1.4(7). Radiation Conditions inside and Outside Containment, it requires that the maximum Beta dose at the surface of 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 for self shielding. The licensee indicated that this 50 percent reduction is standard industry practice and has been previously accepted by NRC. The inspector acknowled9ed the licensee's position on this concern and indicated that this issue was unresolved pending further review by NRC.

The acceptability of the licensee reducing the calculated free field Beta Dose both inside containment and the annulus by 50 perccnt is unresolved and will be identified as URI 50 327.328/96 16 04.

c. Conclusion The inspectors concluded that the licensee failed to implement adequate design controls for reactor core design changes and failed to take prompt and effective correcdve action for nonconform.19 plant conditions identified since September 18. 1990. Three violations were identi fied. Additional review by the NRC has resulted in these violations being changed to URIs pending additional NRC reviews. One unresolved item and one inspector followup item was also identified.

E2 Engineering Support of Facilities and Equipment E2.1 PER No. SO950021PER. Obtain Operability Evaluation for SONP: Review of WBPER940576

a. Insoection Scoce The inspector reviewed PER No. SO950021PER in order to evaluate the adequacy of the licensee's root cause analysis, extent of condition de ficiencies,and developed corrective actions for 10 CFR 50.49 identified evaluation,

b. Observations and Findinos Watts Bar Adverse Condition Report WBPER940576 1dentified a problem with the pressurizer PORVs where the energized times did not agree with limitations imposed by the E0 program. The PORVs had been energized in 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 specified in the E0 binder. This issue was reviewed for applicability to Sequoya.h.

E0 binder SONE 0-50L 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 E0 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 n

9 corrective actions taken to address this issue involved completing an Operability Determination where it was concluded that the PORVs could perform satisfactorily until the cycle 7 outage. The pressurizer PORV solenoidofvalves outage each unit. were subsequently replaced during the cycle 7 refueling I 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 E0 binder. and supporting '

Oualified Life. or Accident Degradation Equivalency Calculations. The results of this review identified 12 E0 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 E0 binders based on the results of the above calculations have been restrained l because of failure to promptly complete the calculations.

c. Conclusions  !

l The inspector concluded that the Operability Determination performed for  !

PER Ho. 50950021PER 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 Scone The inspector reviewed the licensee's program documents that control the i

environmental qualification program to verify 1) that responsibilities i had been defined and 2) requirements had been specified for establishing I and maintaining the auditable documentation demonstrating qualification of equipment in compliance with 10 CFR 50.49. The licensee's transition 1 plans for implementing the E0 program after Phase 1 site engineering re-organization was also reviewed.

b. Observation and Findings Procedure SSP-6 5. Electrical Equipment Environmental Qualification (E0)

Program. Revision 7. 15 the controlling procedure for implementing the E0 program at Sequoyah, Based on review of this procedure the inspector determined that the program controls clearly identified functional responsibilities and levels of authority for adequate implementation of the E0 program. Training requirements for personnel engaged in E0 work aCliVities were also clearly identified on Appendix ! of this procedure.

No deficiencies were identifled with the procedural cnntrols in SSP 6.5.

10 j

The inspector reviewed the licensee's transition plans for implementing the 10 CFR 50.49 program after Phase 1 reorganization of the site engineering and material section. The following documents were reviewed during this effort. j e Procedure SPP 9.2. Eouipment Environmental Qualification (E0) ' i Program, Revision 0 o Procedure NEP 5.12. Program Requirements for Equipment Qualification of Electrical Equipment in Harsh Environments.

Revision 1.

e Mechanical Design Standard No. DS-M18.14.1. Design Standard for Environmental Qualification of Electrical Equipment in Harsh  !

l Environment. Revision 0.

i The inspector also conducted interviews with personnel engaged in E0 work activities from the EE/NE discipline and Maintenance Planning and Technical (MP/T) section. The interviews were intended to assess the level of the worker's understanding of the E0 program requirements and i to verify that EQ training requirements had been met. All personnel

, interviewed were knowledgeable of the E0 program requirements and had l

comple'.ed E0 training. No deficiencies were identified with the licensee's staff involved with E0 program activities.

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 completion of the Phase 1 site engineering reorganization. This is an upper tier program docunnnt that delineate E0 program controls to be i I

implemented at Sequoyah. Browns Ferry and Watts Bar. Based on this i review the inspector determined that SPP-9.2 failed to adequately  !

establish program controls for successful implementation of the E0 program at Sequoyah. Owr.ership of the E0 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 requirements for personnel involved with E0 work activities.

TVA management was info mod of this inspection finding, in response TVA management told the insptactor that they concurred with the findir1s and procedure SPP 9.2 would not be approved for replacing SSP 6.5 in its present form.

The inspector wds also advised of personnel changes that would be implemented on October 1.1936. for Fnase 1 reorganization of the site engineering and materials section. On this date TVA management will have only one person who have completed E0 training in the I&C section which now has owrership of the E0 program. Similarly, one E0 trained person will be in the PD/T section to perform E0 duties. The licensee has essentially de-rentralized the E0 program, disbanded the dedicated staff who performed E0 artivities, and has now included in the position descriptions of engineering and MP/I personnel requirements for performing E0 duties.

e 0

11

c. Conclusion The inspector concluded that the transition plan for implementing the E0 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 E0 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 E0 deficiencies.

E.8 Hiscellaneous Engineering Issues E.8.1 Employee's Concern Program

a. Insnection Scom 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 {

i Qualification Program. The inspector reviewed the emplo documented in the following Concerns Resolution Program.yee's concerns (CRP) files and j

conducted discussion with the Concerns Resolution Staff Manager concerning implementation of the proga am, e File No. ECP-96 50 903 e File No. ECP 96 50 918 e File No. ECP-96-S0 922 e File No. ECP 96-50 927 e File No. ECP 96 50 928 e File No. ECP-96 50-991 Based on these discussion's the inspector determined that File No. ECP-96 50 992-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 E0 Program: Leak Rate Testing: Appendix R: 0 List: Vendor Manuals: and Technical Specification Testings. TVA management had already taken actions to address these concerns. The inspector reviewed Engineering Reorganization Assessment Report. NA50 96023 phase 1. and verified that E0 concerns were addressed in this

• Investigation. The report concluded that althuugh there has been a '

significant reduction sn SON Engineering personnel, contingency plans I and tasks reassignments have been developed to ensure responsibilities are adequately assumed by remaining site and/or cc.t ract personnel.  !

Additional E0 concerns raised by employees have been documented in File

.. j o

i i

12 No. ECP 96 SQ A07 F1. These issues among others have been identified as action items to be included for review in upcoming audits. The inspector was informed that the results of the Engineerin Reorganization Assessment Phase 2. scheduled for January.g 1997 will also provide additional indepth investigation of E0 concerns raised by TVA employees,

c. Conclusion  ; {

l The inspector concluded that employee's concerns are promptly addressed by TVA management. Concerns involving inadequacies in implev nting the  !

10 CFR 50.49 E0 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. i E.8.2 (Closed) Unresolved item (URI) 50 327.328/96 02 04. Omission of  !

Surveillance Tests for Rod Control System.

l URI 50 327.328/96-02 04 was identified in connection with plant  !

modification DCN No. Hil445A, 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 i Rod Control Cluster Assemblies.

10 CFR 50.54 (f). ,

The safety assessment performed for this plant modification was determined to be technically inadequate. Specifically, the Safety Assessment Checklist. Appendix G. Item 22. Incorrectly stated that there were no new credible failure modes associated with the hardware change. '

This error led to omission of requirements from the DCN for development and implementation of recommend surveillances described in WCAP 13864 Revision 1. TVA management in their letter dated June 10. 1990, i comitted to the corrective action delineated in PER No. 50960677 PER i for developing a new procedure to comply with GL 93-04 and the WOG recomiendations. The action due date for this corrective action is ,

February 15, 1997. Additionally, plant modification DCN No. Mll730A. L 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 111 will be identified for failure to implement adequate design controls for " Rod Control System" plant modification.

C. Un 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 detall the inspection results.

One unresolved item related to the technical

.E 13 acceptability of reducing the free field beta dose inside the containment and annulus by 50 percent was identificd: 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 e violation of 10 CFR 50 Appendix B. Criterion III was opened.

On January 8.1997, in a telephone conversation, the licensee was informed that three unresolved items related to PER No. SOP 900372PER were made unresolved items pending the results of a meeting with TVA. A date for the meeting was not yet determined.

Licensee Emoloyees 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

• 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  !

Ooened 50-327.328/96 16-01 URI Inadequate safety evaluation resulted in Unreviewed Safety Question. (Paragraph El)

1 14 50 327,328/96 16 02 URI Untimely corrective ection for nonconforming plant conditions.

(Paragraph El) 50 327.328/96 16 03 URI Inadequate design control for nonconforming plant conditions.

(Paragraph El) 50 327,328/96 16 04 URI Technical acceptability of reducing the calculated free field beta dose inside containment and annulus.

(paragraph El) 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 CAOR Condition Adverse to Quality Report CFR Code of Federal Regulations DBDs Design Basis Documents EEB Electrical Engineering Branch EFPD Effective Full Power Day EGTS Emergency Gas Treatment System E0 Environmental Qualification FSAR Final Safety Analysis Report JC0 Justification for Continued Operation LOCA Loss of Coolant Accident MWt Megawatts Thermal 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 Daners Group

e l

Safety  : 4"W" oy Regulatory o ms on B *?i. 4 .; q: v, h.'g rel51ted... pp'.t%O U

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) a 15.0 ACCID (NT ANALY$l$

1$ 2 Normal. 0peration and Anticipated Operational Transients y Boron Ollution U

j In the safety Evaluation Report we stated the reliance upon an audible rate count

• to alert the operator of postulated baron dilution events during refueling was not i justified.

The applicant provided justification for maintaining the alars setpoint within one-half decade of the source flux level, Based on this margin and on the maximum possible rates of Jilution, the applicant'. analysis showed that the event would be detected and announced by the high flux at shutdown alars within a time period that lef t sufficient margin for the operator to correct the situation before criticality occurred. Fifteen minutes is the required minimum time margin at these conditions in accordance with our Standard Review Plan.

The appitcant has committed to a schedule for setting and monitoring the gap between the high flux at the shutdown alare level and th shutdown source flux level that is consistent with the analysis presented. The setting is to be no higher than 1/2 decade above the count rate, and the margin is to be verified (or reset if necessary) every 30 minutes for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and once per shif t thereaf ter untti the flux level has stabilized.

The required procedures and schedule for verification of the setpoint are to be incorporated in the operator's Surveillance instructions.

f The staf f finds that the analysis, the reactivity changes in the boron dilution event are accounted for satisfactorily. The applicant's analysis cefines a region of reactor conditions for the event tnat are considered safe, according to NRC criteria as described in SRP Section 15.4.6. The procedures adopted by the j 2 appitcant will assure that the reactor remains within the boundaries of the safe l

conditions. The staff, therefore, regstds the question of the boron dilution event immediate'y following shutdown as having been satisfactority resolved.

ATV5 We have reviewed the TVA submittal of October 17, 1979, on Emerger.:.y operating Procedures for the postulated anticipated transients without scram (ATVS) events.

We provided our comments on the proposed procedures and made recosamendations for changes. The proposed procedures must be modified in accordance with our comments However, the 5eoucyah and instructions to be acceptable for full power operation.

15-1 (L b

[* .- l 1

plant *

• be operated at 10-s power (less than or equal to five percent of full

, power prior to completion of procedures modifications without undue risk to the health and safety of the pubile. Our conclusion that low power operation is acceptable is based on our understanding of the expected plant response to the relevant ATV5 events to occur under these operating conditions. 1 I

Normal Operation and Anticipated Operational Transients Section 15.2 of the Sequoyah SER referred ta cur aeneri: review of the Westinghouse Topical Reports WCAP 9226, WCAP-9236, and WCAP 9230 as the licensing bases for the analysis methods and sensitivity studies for postulated main steamline and feedline l breaks. The steamline break information is contained in WCAP-9226. The feedwater line break information was provided in WCAP 9230 and in WCAP 9236, which discusses the NOTRtMP coeputer program used in the analyses. At that time, our review was scheduled for completion in late 1979.

For review of the steamline break topical, the staff requested additional informa-

• tion froe Westinghouse in September 1978. Westinghouse responded with answers to some of our questions in May 1979. In response to staff inquiries Westinghouse has attributed their failure to answer the balance of our questions to higher priority THI 2 analyses requirements.

The staff has previously accepted steaaline and feedline break analyses described in plant applications for PWRs designed by Westinghouse and other reactor vendors.

It has beel our position that a sure detailed account of analytical methods for steaaline and feedline break is required from the vendors for generic review t that the outcome of this review would be applied to licensed reactors. Our g' revis., includes the performance of tr house audit calculations and calculations by technical assistance contractors.

Based on our preliminary review, there is sufficient evidence to conclude that substantial thermal margin exists under postulated steamline and feedline break accident conditions to preclude core damage leading to unacceptable consequences.

Therefore, we conclude that the steamline and feedline break accident analyses for Sequoyah are acceptable wnile our scre detailed review continues. However, our approval is predicated on the assumption that our generic review can proceed on a rencnable schedule. To assure that this assumption is valid, we will recuire a response to our outstanding questions on the topical reports discussed above and a aew commitment for prompt response to any additional information reautrements prior to approval of a full power operating license.

15.3 Accidents and Infrequent Transients 15.3.) Steam tine Break tong-Terr Ef fects of Steam Line Break Because the primary system pressure may have an effect on pressure vessel integrity following a steamline break or a small break loss-of coolant accident, the staff 15-2

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requested additional information regarding the long term scenarios, and effects of

. these events. Using techniques siellar to those reviewed and approved for the D, C. Cook, Unit 2, plant, the applicant has conservatively calculated pressure and temperature conditions for a bounding spectrue of steamline break and small break LOCA events.

Using fracture mechanics techniques the applicant has estimated that, for those accident conditions, reactor vessel integrity can be assured for 17 ef fective full-power years, The fracture mechanics anal y s performed by the applicant is similar to those that we have reviewed for other plants. Although we have not formally accepted these analyses, we do believe they are reasonable and provide assurance that the Sequoyah Units 1 and 2 reactor vessels have adequate margin against failure under postulated accident conditions for a substantial number of years of operation.

As described in Appendix C, Generic Task A 11 is expected to result in an engineer-ing method and safety criteria that will provide the basis for assessing the acceptability of operation over the life of the plant, for both normal transient and accident conditions including consideration of M5LB and small break LOCA. The results of Task A-11 are expected to te available long before they are needed to provide this assessment for the Sequoyah Units 1 and 2 reactor vessels.

Based on the foregoing we have concluded that there is reasonable assurance that the ir.tegrity of the Sequoyah Units 1 and 2 reactor vessels will be maintained during postulated accidents.

Auxiliary Feedwater Runout Flow Followino a $ team Line Break The applicant was requested to address the potential for containment overpressuri-Zation due to the anticipated continuous addition, at pump runout flow, of auxiliary feedwater to the affected steam generator following a postulated main steam line break (MSLB) accident, Our interest in this issue resulted from the 10 CFR Part 21 deficiency report filed by the Virginia Electric and Power Company (VEPCO) dated September 4,1979. In that report, the NRC was informed by VEPC0 that overpressurization of the continment at North Anna Units 3 and 4, could occur in the event of a postulated MSLB inside containment. VEPCO indicated that, due to the anticipated continuous addition of auxiliary feedwater to the broken loop steam generator, at the pump runout flow condition, following a MSLB accident, the containment pressure will reach the containment design pressure in about 10 minutes.

3 To determine if the issue under consideration was generic for all pressurized water reactors (PWRs), we initiated a review of all "near-term" operating iteense appli-cations for PVR plants. The object of the review was to determine if auxiliary feedwater flow was considered in the M5tB analyses and, 1" so, whether pump runout flow conditions were used.

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The applicant indicated that the auxiliary feedwater syste] utilizes runout flow

. control equipment to limit the flow. Therefore, in the original MSLB analysis, the sumiliary feedwater flow to the f aulted stone generator was assumed to exist at maximum capacity from the time of the rupture until realignment of the system is completed by the operator, 10 minutes after the onset of the postulated accident.

The appilcant's original submittal, stated that in one of the pottulated analyses  ;

performed, a failure of the auxiliary feedwater runout protection system was assumed.

In this analysis, it was assumed that flow to the broken loop steam generator at pump runout flow conditions continued frote onset of the accident until the operator manually terstnates flow 10 minutes later. It was concluded by the applicant, and the staff concurs, that the peak containment pressure will reeafn below the containment design pressure. The applicant also indicated that information for use in deciding to terminate the auxiliary feedwater flow to the affected steam generato will be available to the operator tonediately after onset of the accident.

Based on our review of the applicant's evaluation, we find that the applicant's analyses have correctly accounted for the auxiliary feedwater flow and that no o

further analysis is required.

Normal Operation and Anticipated Operational Transients We have reviewed the TVA subelttal of November 9,1979 responding to IE Informa-tion Notice 79-22 on qualification of control systems for Sequoyah Units 1 and 2.

The subatttal identifies plant systems required for safety and states for each safety function that adequate instrumentation would alert the operator to an event, adequate time is available for operator action, and control systes design permits operator action. Eased on the information provided by the applicant, our review of the Sequoyah Final Safety Analysis Report, our related reviews of equipment qualifi-t cation, and sinflar reviews for operating reactors, we have found no event sequence that leads to an unacceptable consequence.

We have concluded that the Sequoyah applicant has satisfied the standards set for operating reactors and that this issue presents no concerns which would restrict operation of the plant.

15.4 Radiolooical Consequences of Accidents 15.4.1 Less-of-conlant Accident This section of the supplement revises in its entirety the material that was present in he Safety Evaluation Report. The Sequoyah Nuclear Plant includes a double cantsinment design to collect and filter the leakage of fission products from a postulated design basis loss-of-coolant accident. The double containment consists of a free standing steel primary containment vessel sur.ounded by a reinforced concrete shleid building. The reinforced concrete auxiliary building is also a par $ of the seconCary containment barrier. Leakage which enters the secondary containment is treated by either the emergency gas treatment system or the auxiliary l building gas treatment system prior to release to the atmosphere. Both of these 15 4

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1 systems are engineered safety features. Another engineered safety feature is the

. Ice condenser with a sodium tetraborate additive to the ice to enhance the removal of lodine in the containment following a loss-of-coolant accident. The dose model and dose conversion parameters are consistent with those given in Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of Coolant Accident for Pressurl ed Water Reactors."

1 In the analysis of the design eatis loss-of coolant accident, the primary contain-ment vis assumed to leak at the design leak rate of 0.25 percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident and at 0.125 percent per day thereaf ter.

The applicant established to the staff's satisfaction that the shield building annulus pressure would not exceed -0.25 inch water gauge pressure and that no

• 1eakage would bypass the gas treatment system throughout the course of the accident (see Section 6.2 of this report for further discussion of these items). The applicant has increased the amount of leakage which enters the auxiliary building following the accident froe 10 percent to 25 percent of the primary containment leakage, assuming that this leakage was exhausted directly to the atmosphere during the first 10 minutes of the accident. Af ter 10 minutes the leakage is processed through the auxiliary building gas treatment system without credit for holdup or sizing.

Seventy-five percent of t,he leakage from the primary containment enters the shleid building annulus where we assumed that it vent directly to the intake of the shleid building annulus recirculttion/ exhaust system. Following passage through the eeergency gas treatment system filters, a fraction of this leakage was assumed in our analysis to be exhausted to the atmosphere with the remainder recirculated to the shleid butiding annulus whsre credit was given for atxing in 50 percent of the annulus free volume. The split between the exhaust and recirculition fractions

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was assumed to be proportional to the air flow rates in th. exhaust and recircula-tion paths of the systems.

The applicant assumed in his dose analysis that it takes 10 minutes to isolate the ,

auxiliary building rather than the previous assumption of 5 minutes (the applicant's analysis of the auxiliary building gas treatment system indicated that the system is desig ed to draw down the building to a -0.25 inch water gauge pressure within 170 seconds). Therefore, our analysis assumes that all leakage into the auxiliary building for the first 10 minutes into the accident is immediately released to the environment. For all times af ter the first 10 minutes into the accident we assume the leatage is exhausted through the gas treatment system.

The doses we calculate for the postulated design basis loss of coolant accident for the Sequoyah Nuclear Plant, shown in Table 15-1, are within d.e exposure guidelines of 10 CFR Part 100.

As part of the loss-of-coolant accident, we have also evaluated the consequences of leakage of containment sump water which is circulated by the emergency core cooling system after that postulated accident. We have assumed the sump water 15-5 l

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contains a mixture of ladine fission products in agreement with Regulatory Culde 1.7, " Control of Combur v ble Gas Concentrations in Containment Following a Loss of Coolant Accident." During the recirculation mode of operatica the sump water is circulated outside of the containment to the auxiliary building. If a source of leakage should develop, such as fram a pump seal f atture, a fraction of the iodine in the water could become airborne in the auxiliary butiding and exit to the etmosphere. Since the emergency core cooling system area in the auxiliary building is served by an engineered safety features air flitration system (the auxiliary building gas treatment system), we conclude that the doses resulting from the postulated leakage of recirculation water would be low and, when added to the direct leakage less of coolant accident doses, wculd result in total doses that are within the guideline values of 10 CFR Part 100.

As discussed in Section 6.2.3 of this report, the applicant recently informed us that during the ongoing Unit 2 construction activities, the minimum pressure that can be achieved in some of the ESF pump rooms will be approximately -0.04 inches water gauge as compared to the -0.25 inches required by the Technical Specifications.

We determined that this pressure is not sufficiently low to assure the removal of airborne iodine activity by the auxiliary building gas treatment system following a postulated accident. We, therefore, have evaluated the 30-day dose at the LPZ distance for a postulated ESF pump seal failure following a loss of coolant accident.

We conservatively assumed no holdup, mixing or removal of the associated airborne lodine activity in the auxiliary building. We also assumed that the Unit I reactor will be operated du*ing this interim period of unit construction at a power level not in excess of 5 percent of the rated power of 3582 n. thermal. Other assumptions of our analyses are listed in Table 15.3.

• Based on our evaluation we conclude that the radiological consequences associated with an E5F pues seal f ailure in conjunction with the doses resulting from a design basis accident are within the guidelines of 10 CFA Part 100. We also conclude that the Unit I reactor shall not be operated at a power level in excess of 5 percent of the rated power level unless the applicant can demonstrate, by * , a' the I

E5F pump room can achieve and maintain a pressure not higher than the -0.25 inch water gauge identified in the Technical Specifications.

The applicant may purge the containment periodically during reactor operatien.

Should a loss of-coolant accident occur when the purge lines are open, a portion of the containment atmosonere plus a portion of any flashed reactor coolant con-talning radioactive iodine fission products would be released to the environment in the short interval oefore the purge isolation valves close and isolate the containment, be have estimated the radiological consequences of this event using conservative assumptions regarding the radioactive iodine concentration in the primary coolant, the amount of reactor coclant inventory released, and the flow rate through the valves. We conclude that the consecuences are such that, even when added to the Calculated doses from Containment leakage, the total is within the guideline values of 10 CFR Part 100.

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4 The applicant has provided redundant hydrogen recorbiners for the purpose of con-l trolling any accumulation of hydregeh within the primary containe.ent following a

, loss of coolant accident. In the event of failure of both recombiners, tht appilcant A

h has provided a backup system. The purged containment effluent would flow to the shield building annulus where it would be subsequently discharged to the atmosphere through the emergency gas treatment system filters. We find the combination of ,

redundant recombiners plus a backup purge capability to be an acceptable method ~

for controlling the potential contribution to the offsite doses from hydrogen purging following a loss of coolant accident.

l While Unit 2 is under construction the equipment hatch of the Unit 2 containment ,

building will be closed off from the interim auxiliary building by two stoel roll-

. up dst.'s. l These doors must be closed in the case of an accident in order to draw  ! l l

down the interim auxiliary building to a negative pressure of 0.25 inch water gauge. .;

These doors will be locked shut or alarmed in the Unit I control room under normal

conditions and plant personnel will be statiened at the doors when they are in use in order to initiate their immediate closing in the case of an accident. Tha staff j , concludes that this control will provide adequate assurance that the interim auxiliary building can be crawn down to the required negative pressure.

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TABLE 15-1 RADIOLOGICAL CONS [QU[NC[5 0F +

DESIGN BASIS ACCIDENTS Exclusion Area" Lew Population Zone **

2 Hour Dose, Rea 30-0sy Dose, Ree Accident Thyroid whole Body Thyroid Whole Body Loss of Coolant 194 9 28 1 i

Fuel Handling 20 1 <1 <1 Stese Line Break i 1) 1-131 at 1 alcrocurie per gram 13 < 0.1 <1 <0.1 1

2) 1-131 at 60 microcuries per gram 26 < 0.1 1 <0.)

• 5tese Generator Tode Rupture

1) 1-131 at 1 micr6:urie per gram 19 < 0.1 1 <0.1 l 2) 1 131 at 60 alcrocuries per gram 214 <0.1 10 <0.1 Control Rod Ejection

1) Leakage thruugh secondary side 42 <0.1 2 < 0.1

2) Leakage through containment 97 <0.1 4 <0.1 Part 100 guideline dose values are: 300 rea thyroid 25 rea whole body

• Exclusion area 7inieurn boundary distance = 556 meters

• • Low population Zone distance = 4828 meters 15 8 "h

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TABLE 15 2 l t

A$$UMPT!ONS U$fD IN THE CALCULATION OF j LO55-0F-C00LANT ACCIDENT 005E 5  ;

Power Level 3542 Megawatts tht*41

= - . . . . . . . . . . . .p berating flee 3 years

, Fraction of Core Inventory Available for Leakage lodines 25 percent Noule Cases 100 percent Initial lodine Composition in Containment Elemental 91 percent Organic 4 percent Particulate 5 percent Primary Containment Volumes Upper containment 7.16 x 10

• Lower compartment (including ice condenser) 5.25x10lcubicfeet cubic feet Shield Butiding Annulus Volume 3.75 x 105 Cung, y,,g Mixing Fraction in Annulus 50 percent Annulus Ventilation Flow Olstribution i 4 e

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TABLE 15 2 (Con't)

Recirculation Flow Exhaust Flow, Time Step Cubic Feet Per Minute . Cubic Feet Per Minute 0 46 seconds 0 0 46 200 seconds 500 3500 200-400 seconds 1500 2500 410-1000 seconds 3000 1000 1000 seconds - 30 days 3900 100 Filter Efficiencies

. Elemental lodine 95 percent Organic lodine 95 percent Particulate lodine 95 percent Ice' Condenser Removal Efficiency Elemental lodine 30 percent Flow Rate through ice Condenser 40,000 cubic feet per minute Period of Ice Condenser Effectiveness 10-60 minutes Primary Containment Leak Rates 0 - 24 Hours 0.25 percent per day

> 24 Hours 0.125 percent per day Eypassing Leakage Fraction (Auxiliary Building Pathway) 0-10 Minutes 25 percent

>10 Minutes O percent Minimum Exclusion Area Boundary Distanca 556 meters low Population Zone 01 stance 4828 meters Atmospheric Diffusion (X/Q) values 02 Hours 1.4 x 10', see per cubic meter 0-8 Hours 6.4 x 10, see per cubic meter 8 24 Hours 4.5 x 10, see per cubic meter 1-4 Days 2.1 x 10, sec per cubic meter 4-30 Days 6.9 r 10 see per cubic meter Q

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e TAett 15 3 ASSUMPTIONS U$ED IN THE CALCULATION OF E5F PUMP SEAL FAILUR[

Power Level 180 Megawatt thermal (5 percent of rat'ed)

Atmospheric Diffusion Values See Table 15-2 Liquid Volume in Primary Containment 500,000 gallons flee of Pump Seal Failure Af ter LOCA 24 hrs.

Pump Seal Failure Flowrate 60 gallons / minute Isolation of Pump Seal Failure 30 minutes Evaporation Fraction 0.1

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