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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20212J6311999-10-0101 October 1999 SER Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plant,Unit 1 ML20212F4761999-09-23023 September 1999 Safety Evaluation Supporting Amends 246 & 237 to Licenses DPR-77 & DPR-79,respectively ML20212F0831999-09-23023 September 1999 Safety Evaluation Granting Relief from Certain Weld Insp at Sequoyah Nuclear Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) for Second 10-year ISI Interval ML20196J8521999-06-28028 June 1999 Safety Evaluation Authorizing Proposed Alternative to Use Iqis for Radiography Examinations as Provided for in ASME Section III,1992 Edition with 1993 Addenda,Pursuant to 10CFR50.55a(a)(3)(i) ML20239A0631998-08-27027 August 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Sequoyah Nuclear Plant,Units 1 & 2 ML20236Y2091998-08-0707 August 1998 Safety Evaluation Accepting Relief Requests RP-03,RP-05, RP-07,RV-05 & RV-06 & Denying RV-07 & RV-08 ML20217K4471998-04-27027 April 1998 Safety Evaluation Supporting Requests for Relief 1-ISI-2 (Part 1),2-ISI-2 (Part 2),1-ISI-5,2-ISI-5,1-ISI-6,1-ISI-7, 2-ISI-7,ISPT-02,ISPT-04,ISPT-06,ISPT-07,ISPT-8,ISPT-01 & ISPT-05 ML20138D2581997-04-28028 April 1997 Safety Evaluation Authorizing Licensee Proposed Alternative to Use 1989 Edition of ASME Boiler & Pressure Vessel Code, Section XI for Performance of Containment Repair & Replacement Activities Until 970909 ML20057F8441993-10-14014 October 1993 SER Granting Relief Giving Due Consideration to Burden Upon Licensee That Could Result If Requirements Imposed on Facility ML20057D5321993-09-28028 September 1993 SER Granting Licensee 921117 Relief Requests ISPT-2 & ISPT-3 Re Inservice Pressure Test Program ML20057D6351993-09-28028 September 1993 SER Granting Relief as Requested for Both ISPT-2 & ISPT-3 Per 10CFR50.55a(a)(3)(i) & 10CFR50.55a(g)(6)(i) ML20128K0221993-02-11011 February 1993 SE Accepting Util Justification for Break Exclusion of Main Steam Lines in Valve Vaults Provisionally Until End of Refueling Outages ML20128E9161993-01-0606 January 1993 SE Approving Request for Relief from ASME Requirements Re First 10-yr Interval ISI Plan ML20247K3321989-09-14014 September 1989 Safety Evaluation Accepting ATWS Mitigation Sys,Pending Tech Spec Issue Resolution ML20245E6951989-08-0303 August 1989 Safety Evaluation Supporting Inclusion of Alternate Repair Method to Detect microbiologically-induced Corrosion in Previously Granted Request for Relief from ASME Section XI Code Repair Requirements ML20247G8661989-07-21021 July 1989 Safety Evaluation Re Silicone Rubber Insulated Cables. Anaconda & Rockbestos Cables at Plant Environmentally Qualified for Intended Function at Plant & Use Acceptable for 40 Yrs ML20247B4891989-07-19019 July 1989 Safety Evaluation Supporting Util 890330 Request to Eliminate Dynamic Effects of Postulated Primary Loop Pipe Ruptures from Design Basis of Plant,Using leak-before- Break Technology as Permitted by Revised GDC 4 ML20246N0321989-07-11011 July 1989 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 1.2, Post-Trip Review,Data & Info Capability ML20244D1771989-06-0909 June 1989 Safety Evaluation Re Generic Ltr 83-28,Items 2.1.1 & 2.1.2 NUREG-0612, Safety Evaluation Supporting Util Request to Delete Three Commitments in Response to NUREG-0612 Re Heavy Load Control on 5-ton Electric Monorail Hoist W/Integral Trolley & 4-ton Monorail Chain Hoist W/Geared Trolley1989-05-26026 May 1989 Safety Evaluation Supporting Util Request to Delete Three Commitments in Response to NUREG-0612 Re Heavy Load Control on 5-ton Electric Monorail Hoist W/Integral Trolley & 4-ton Monorail Chain Hoist W/Geared Trolley ML20245A1301989-04-14014 April 1989 Safety Evaluation Re Shutdown Margin.Procedural,Hardware & Training Enhancements Implemented & Committed to by Util Will Provide Reasonable Assurance That Adequate Shutdown Margin Will Be Maintained at Plant ML20244D8821989-03-14014 March 1989 Safety Evaluation Supporting Procedural,Hardware & Training Enhancements Implemented & Committed to by Util to Provide Reasonable Assurance That Adequate Shutdown Margin Will Be Maintained at Plants ML20195J0891988-11-28028 November 1988 Safety Evaluation Accepting Program for Plant in Response to Items 4.2.1 & 4.2.2 of Generic Ltr 83-28 Re Reactor Trip Sys Reliability ML20205T1621988-11-0707 November 1988 Safety Evaluation Supporting Improvement Plan for Emergency Diesel Generators Transient Voltage Response ML20206G4531988-11-0404 November 1988 SER Supporting Employee Concern Element Rept Co 15101, Floor Drains ML20206G3961988-11-0404 November 1988 SER Supporting Util Investigation of Employee Concerns as Described in Element Rept 308.03 ML20206G5341988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 30114, Malfunction of Doors ML20206G4621988-11-0404 November 1988 SER Supporting Employee Concern Element Rept 204.8(B), Communication & Interface Control ML20206G5291988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 301112, Sys 31 Not Operated Properly ML20206G5241988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 30111, Valve Closure ML20206G5191988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 30105, Questionable Design & Const Practices ML20206G5091988-11-0404 November 1988 SER Supporting Employee Concern Element Rept 23706, Gassing of Current Transformers ML20206G5021988-11-0404 November 1988 SER Supporting Employee Concern Element Rept 23504, Exposed HV Cable Routed W/O Raceway - Personnel Hazard ML20206G4571988-11-0404 November 1988 SER Supporting Employee Concern Element Rept Co 15105-SQN, Flex Hose Connections ML20206G4971988-11-0404 November 1988 SER Supporting Employee Concern Element Rept 23501, 480 Volt Power Receptacles Unsafe ML20206G4861988-11-0404 November 1988 SER Supporting Employee Concern Element Rept EN 232.9(B), Freezing of Condensate Lines ML20206G4591988-11-0404 November 1988 SER Supporting Employee Concern Element Rept 204.7(B), Vendor Documents Legibility & Dissemination Sys ML20206G4801988-11-0404 November 1988 SER Supporting Element Rept EN 232.2, Carbon Steel Vs Stainless Steel Drain Pipes ML20206G4721988-11-0404 November 1988 SER Supporting Employee Concern Element Rept 22912, Panel- to-Equipment Distances ML20206G4661988-11-0404 November 1988 SER Supporting Employee Concern Element Rept EN 229.6(B), Lack of Valves in Sampling & Water Quality Sys ML20206G5431988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 30301, Difficulty in Obtaining Obsolete Equipment ML20206G6111988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 31105, Alara ML20206G6161988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 31106, Health Physics Facilities,Clothing & Protective Equipment ML20206G6211988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 31204-SQN, Mgt & Personnel Issues ML20206G6321988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 31208-SQN, Security at Plant Entrances ML20206G6371988-11-0404 November 1988 SER Supporting Employee Concern Element Rept OP 31201-SQN, Adequacy of Public Safety Svc (Pss) Officer Uniforms in Nuclear Plant Environ ML20206G4351988-11-0404 November 1988 SER Supporting Employee Concern Element Rept Co 11101-SQN, Contact Between Dissimilar Metals ML20206G4381988-11-0404 November 1988 SER Supporting Employee Concern Element Rept Co 11202-SQN, Craft-Designed Hangers as Related to Const ML20206G4081988-11-0404 November 1988 SER Supporting Employee Concern Element Rept Co 10307-SQN, Uncoated Welds as Related to Const ML20206G3661988-11-0404 November 1988 SER Supporting Employee Concern Element Rept EN 21002, Inadequate Environ Qualification of Electrical & Instrumentation Control 1999-09-23
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20212J6311999-10-0101 October 1999 SER Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plant,Unit 1 ML20217G3721999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Sequoyah Nuclear Plant.With ML20212F0831999-09-23023 September 1999 Safety Evaluation Granting Relief from Certain Weld Insp at Sequoyah Nuclear Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) for Second 10-year ISI Interval ML20212F4761999-09-23023 September 1999 Safety Evaluation Supporting Amends 246 & 237 to Licenses DPR-77 & DPR-79,respectively ML20212C4761999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Sequoyah Nuclear Plant.With ML20210L4361999-08-0202 August 1999 Cycle 9 12-Month SG Insp Rept ML20216E3781999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20210L4451999-07-31031 July 1999 Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept ML20210G6631999-07-28028 July 1999 Cycle 9 90-Day ISI Summary Rept ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20209H3831999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Sequoyah Nuclear Plant.With ML20211F9031999-06-30030 June 1999 Cycle 9 Refueling Outage ML20196J8521999-06-28028 June 1999 Safety Evaluation Authorizing Proposed Alternative to Use Iqis for Radiography Examinations as Provided for in ASME Section III,1992 Edition with 1993 Addenda,Pursuant to 10CFR50.55a(a)(3)(i) ML20195K2951999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20206Q8951999-05-0505 May 1999 Rev 0 to L36 990415 802, COLR for Sequoyah Unit 2 Cycle 10 ML20206R5031999-04-30030 April 1999 Monthly Operating Repts for April 1999 for Sequoyah Units 1 & 2.With ML20205P9811999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20204C3111999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20205B6631999-02-28028 February 1999 Underground Storage Tank (Ust) Permanent Closure Rept, Sequoyah Nuclear Plant Security Backup DG Ust Sys ML20203H7381999-02-18018 February 1999 Safety Evaluation of Topical Rept BAW-2328, Blended U Lead Test Assembly Design Rept. Rept Acceptable Subj to Listed Conditions ML20211A2021999-01-31031 January 1999 Non-proprietary TR WCAP-15129, Depth-Based SG Tube Repair Criteria for Axial PWSCC Dented TSP Intersections ML20198S7301998-12-31031 December 1998 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept ML20199G3641998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20197J5621998-12-0303 December 1998 Unit 1 Cycle 9 90-Day ISI Summary Rept ML20197K1161998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20195F8061998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Sequoyah Nuclear Plant.With ML20154H6091998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20154H6251998-09-17017 September 1998 Rev 0 to Sequoyah Nuclear Plant Unit 1 Cycle 10 Colr ML20153B0881998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Sequoyah Nuclear Plant.With ML20239A0631998-08-27027 August 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Sequoyah Nuclear Plant,Units 1 & 2 ML20236Y2091998-08-0707 August 1998 Safety Evaluation Accepting Relief Requests RP-03,RP-05, RP-07,RV-05 & RV-06 & Denying RV-07 & RV-08 ML20237B5221998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Snp ML20237A4411998-07-31031 July 1998 Blended Uranium Lead Test Assembly Design Rept ML20236P6441998-07-10010 July 1998 LER 98-S01-00:on 980610,failure of Safeguard Sys Occurred for Which Compensatory Measures Were Not Satisfied within Required Time Period.Caused by Inadequate Security Procedure.Licensee Revised Procedure MI-134 ML20236R0051998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Sequoyah Nuclear Plant ML20249A8981998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Sequoyah Nuclear Plant,Units 1 & 2 ML20247L5141998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Sequoyah Nuclear Plant ML20217K4471998-04-27027 April 1998 Safety Evaluation Supporting Requests for Relief 1-ISI-2 (Part 1),2-ISI-2 (Part 2),1-ISI-5,2-ISI-5,1-ISI-6,1-ISI-7, 2-ISI-7,ISPT-02,ISPT-04,ISPT-06,ISPT-07,ISPT-8,ISPT-01 & ISPT-05 ML20217E2221998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Sequoyah Nuclear Plant ML20248L2611998-02-28028 February 1998 Monthly Operating Repts for Sequoyah Nuclear Plant,Units 1 & 2 ML20199J2571998-01-31031 January 1998 Cycle 9 Voltage-Based Repair Criteria 90-Day Rept ML20202J7911998-01-31031 January 1998 Monthly Operating Repts for Jan 1997 for Sequoyah Nuclear Plant,Units 1 & 2 ML20199J2441998-01-29029 January 1998 Snp Unit 2 Cycle Refueling Outage Oct 1997 ML20199F8531998-01-13013 January 1998 ASME Section XI Inservice Insp Summary Rept for Snp Unit 2 Refueling Outage Cycle 8 ML20199A2931997-12-31031 December 1997 Revised Monthly Operating Rept for Dec 1997 for Sequoyah Nuclear Plant,Units 1 & 2 ML20198M1481997-12-31031 December 1997 Monthly Operating Repts for Dec 1997 for Sequoyah Nuclear Plant,Units 1 & 2 ML20197J1011997-11-30030 November 1997 Monthly Operating Repts for Nov 1997 for Sequoyah Nuclear Plant,Units 1 & 2 ML20199C2951997-11-13013 November 1997 LER 97-S01-00:on 971017,vandalism of Electrical Cables Was Observed.Caused by Vandalism.Repaired Damaged Cables, Interviewed Personnel Having Potential for Being in Area at Time Damage Occurred & Walkdowns ML20199C7201997-10-31031 October 1997 Monthly Operating Repts for Oct 1997 for Sequoyah Nuclear Plant L-97-215, SG Secondary Side Loose Object Safety Evaluation1997-10-23023 October 1997 SG Secondary Side Loose Object Safety Evaluation 1999-09-30
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NUCLEAR REGULATORY COMMISSION wAsWNGTON, O. C. 20M6 [
%,*...? j ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF SPEC!AL PROJECTS l i
SHUTDOWN MARGIN j TENNESSEE VALLEY AUTHORITY SEQUOYAHNL!CLEARPOWERPLANTUNIT91AND2 DOCKET NOS. 50-327 AND 50-328 I
- 1. INTRODUCTION During scrams at Sequoyah Unit 2 on May 19, 1980 and June 6, 1988, NRC ,
inspectors noted and brought to the attention of TVA operational nanagement an i anomaly in the behavior of the core average temperature (Tav) immediately after a scram. Imediately after both of these scrams, Tav dropped substantially (at ,
i least 25'F) below the no load Tav of Sa7*F. This drop was much greater than the values the inspectors were familiar with at other Westinghouse plants. ,
Sequoyah operators told the inspectors that these drops were typical at !
Sequoyah.
)
- Sequoyah, like all Americtn PWRs, exhibits a markedly negative moderator temp-erature coefficient at the end of core life. A temperature drop of this ,
magnitude adds significant reactivity to the core and may compromisa the i i minfrum reactivity shutdown margin requirement of the Technical Specifications (TS). Therefore, the NRC conducted an inspection at Sequoyah of the core i
shutdown margin on July 11-14, 1988. The results of that inspection are .
1 documented in Inspection Report 50-327,3?8/88-33 dated Septenher 12, 1988.
! TVA also addressed the shntdown margin issue in a Licensee Event Report
! (LER 50-328/88-030) dated July la,1989 and in a submittal dated August 31, 1988.
- f. EVALUATION ,
During the July 11-14, 1988 inspection, an NRC team determired for Sequoyah -
Unit 2 Core 3 that after a reactor trip Tav decreased to en average value of 519'F. This is a drop of 28'F. The minin'm temperature reached was 506*F. i There was no significant difference ir, tne average minimum post-trip temperature before and after the 1985-PP shutdown.
During Sequoyah's initial test program, a post-trin plant performance test was run. During that test, minimun Tay only dropped to 537*F. Durin, the July, l
1988 inspection. TVA attributed the current larger drop to increased leakage
! in the auxiliary steam systems. While this undoubtedly could be a factor, the ;
' team observed that changes in the rain feedwater pump trip functions to prevent r water hammer which had the consequence of starting the steam driven auxiliary feedwater pump earlier in the post-trip transient could also be a factor.
Similarly, changes in control loop tuning for main and auxiliary feedwater l 8810040189 880930 PDR ppOCK C5000327
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- t. ,
i i
.. t 1 -2 :
) and for steam dump could influence core Tav behavior. Therefore, the staff ;
felt that no single factor could be judged a priori to be the cause of the {
4 current greater post-trip drop in Tav and teat only a comprehensive examina- '
l tion of all possible causes would be sufficient to address the problem, in ar August 31, 1988 letter, TVA described a study they were undertaking to ad&ess these various factors. The results of that study and a program for i
- long-term corracthe action will be submitted by TVA by October 14, 1988.
! During the inspection. TVA provided an analysis by Westinghouse that concluded 4 that for all trips to date the TS 3/4.1.1 shutdown margin of 1600 pcm for !
! Unit 2 had been maintained. The inspection' team indepr.ndently calculated the ,
shutdown margins for the five scrams that occurred after restart on May 13, 1988 and examined the input data for the earlier pre-shutdown scrams, Based on this independent review, the staff concurs in Westinghouse's conclusion i that, in terms of the actual physics of the core, the TS minirium shutdown j margin has been maintained for Unit 2. (
l However, for the scram occurring on June 6, 1988, calculations by the team j using the approved TVA shutdown margin procedure then in place, showed that the ;
- TS minimum allowable shutdown margin requirement had been violated. TVA's L j recalculation, which was done at the team's request, confirmed this finding.
4 The team noted, during its inspection, that the TVA shutdown margin instruc- i
- tion (TI-22) was in error in the conservative direction in that the procedure t i required use of a 600 pcm reactivity penalty to account for power Xenon worth l effects. According to Westinghouse's June 17, 1987 1etter to TVA, this penalty .
j needed only to be taken when the reactor core was not at equilibrium prior to l the scram. The procedure in effect on June 6,1988 did not reflect this I caveat. The procedure has since been ccrrected. For the particular scran in I l question, the staff agrees with Westinghouse that the penalty need not have !
been applied and, had it not been applied, the TVA calculation would have shown ,
! adequate TS shutdown margin. The staff notes that had this scram occurred l
- before the xenon had stabilized or if the temperature had, during this trip, i l
dropped as low as the lowest valve in the past (i.e., 506*F), the TS margin [
! reqttrement would have been violated. i I L I As part of its review, the staff examined the Sequoyah Final Safety Analysis l l
Report (FSAR) and the underlying Westinghouse calculations. It discussed the !
calculations with Westinghouse on July 13 and 14,19RS to determine the details of the basis on which the shutdown limit was basad. The steam line break I
l accident controls the mininum core shutdown margin. Vestinghouse calculatinns 1 indicate that a steam line break will cause localized recriticality due to [
l excess cooling and that without adequate shutdown margin this can lead to l 4 departure from nucleate boiling (DNPi in the core ati consecuent fuel clad ;
j failure in the accident. The accident is most severe at nc-load Tav at 0* j
. power and at end of core life. According to Festinghouse, its calculations ,
l assure a power defect reactivity enuivalent to that 'cr 2*F above full load [
l Tav to 2 F below no load, and, therefore, do rot include margin for large !
excursions telcw no inad Tav. j j
L l !
2 ,
f i l '
) .
i .
The team, during the inspection, examined the process by which accident input data was developed by TVA for Westinghouse and determined that the system provided formal methods for plant personnel including the plant reactor engineer to concur in the data. Nonetheless, even though the post-trip Tav undershoot was well-known, it was not identified in the reload checklist provided to Westinghouse.
The consequences of Tay undershoot are more severe during Cycle 3 operation since this was the first low leakage core for Sequoysh Unit 2. The fuel loading pattern used for the low leakage core lowers total rod worth. There-fore, the end-of-life calculated shutdown reactivity at 547'F decreases from 2120 pcm in core 1 to 1600 pcm in core 3. This elimination of excess shutdown reactivity eliminates margin between required and actual minimum shutdown reactivity. However, even though this conbination of Tav depression and reduction in margin might tend to increase the probability or consequences of a design basis accident, it was not addressed in the analysis by TVA for the Core 3 reload. Under 10 CFR 50.59, TVA must justify not submitting an application for a license amendment for the Core 3 reload.
When TVA began to address this issue in June 1988, there were two optione.
available to assure that Unit 2 remained within the design envelope assumed in the accident analysis. Either the operators could control auxiliary feedwater
( AFW) pump flow in manual to maintain Tav as is done at most other PWR plants or they could inject boron to assure that the TS minimum shutdown margin is naintained. Since TVA had installed an automatic AFW control system to address what they perceived were deficiencies in relying on manual action to control AFW, TVA elected to require the operators to add a specified amount of borated water to the plant depending on the post-trip Tav. This alternative is accep-table to the staf f as a short-tern measure until the restart from the Unit ?
Cycle 3 refueling outage. The same problen exists for Unit 1 Core 4 as the core physics characteristics are virtually identical. Therefore, TVA has proposed to use the same corrective action, manual boration after trip, to address the' problem for Unit I restart. Early in core life, the calculated shutdown reactivity is much higher and, for temperature undershoots of the range Seouoyah has experienced, the post-trip Tav excursion would not lead to a violation of the TS shutdown margin limits. Therefore, the staff considers manual boration an acceptable interin neasure to justify restert of Sequoyah Unit 1. However, the staff will review TVA's corrective action program, to be submitted October 14, 198f!, before deciding whether the situation is neceptable for the entire life of Unit 1 core 4.
The team examined the reactor trip procedure which was rodified as discussed above and concludes that the actions specified for the operators provide a reasonable method to assure that adequate shutdown margin will be maintained.
The staff also examined, at TVA's Fuel Performance Franch in Chattanonga, the TVA calculations that determined the arount of borated water to t'e injected.
The calculations were done usine approved rethods and techniques, were appro-priate and conservative, used doct.mented input data and were controlled and verified in a manner consistent witt, their tafety significance, The staff
i ,
l 1 l 7, I notes that the boron additiens were calculated using an assumption of a band to account for errors ',n rod worth determir,ation. The Sequoyah FSAR i
assumed a 10% band. However, Westinghouse in a Topical Report (WCAP 9217 end
- 9218 dated October 1977) provided a basis for showing that the 7% was adequate. ;
This Topical Report was approved by the staff in an SER dated June 15, 1978 and is appropriate for application to Sequoyah.
As stated above, the staff considers the compansatory method described above to assure adequate shutdown margin to be an acceptable interim compensatory action. In its August 31, 1988 letter on shutdown margin. TVA states that it !
i will submit details of its specific program to the staff by October 14, 1988
- to address long-term corrective actions. This program would presumably ;
address reduction in steam system leakage, improved control of steam dumps and r auxiliary feedwater system, modification of core physics parameters or changes ,
to accident analyses.
l
3.0 CONCLUSION
Based on the reviews, inspections and submittals discussed above, the NRC i staff concludes that the procedural changes implemented by TVA for Sequoyah i l
will assure adequate shutdown margin and that, for Unit 2 Core 3 operation and Unit 1 Core 4 restart, the identified excessive undershoot of Tav ;
following a trip does not rcw constitute an unresolved safety question. This r conclusion is limited to short-term operation since it clearly relies on immediatt operator action to compensate for identified deficiencies in either l
]
the design or systen maintenance at Sequoyah. The staff will recuire the I
submission and staff review of the TVA corre'tive action program plan and correction of Unit 2 prior to the restart from Cycle 3 refueling of Unit 2.
l The acceptability of the present situation for the entire Unit 1 Core 4 fuel ,
cycle will be addressed by the staff when the October 14, 1988 submittal
! is reviewed. The need ard schedule for modifi:ations to systems or core i design will be addressed when the TVA corrective action plan is reviewed !
l by the staff.
I 1
Frincipal Contributor: E. Goodwin i
l Dated: September 30, 1988 I
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