ML20083Q803

Proposed Tech Specs Re Conditional Exemption of Eol Mtc Measurement Required in TS SR 4.1.1.3

ML20083Q803
Person / Time
Site:	Sequoyah
Issue date:	05/19/1995
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20083Q790	List: ML20083Q788 ML20083Q803
References
NUDOCS 9505260212
Download: ML20083Q803 (32)

Text

. _ _ _ _ - _ _

ENCLOSURE 1 PROPO5ED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328

(

(TVA-SON-TS-95-07) l LIST OF AFFECTED PAGES Unit 1 3/4 1-5 6-21 6-21 a Unit 2 3/4 1-5 6-22 6-22a l

l 9505260212 950519 PDR ADOCK 05000327 P PDR

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, REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 'The MTC shall be determined to be within its limits during each' fuel cycle.as follows:

a. The MTC shall be measured and compared to the'BOL limit specified in Rif the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel' loading.

b.- The MTC shall be measured at any THERMAL POWER and compared to the

'300 ppm surveillance limit specified in the COLR (all rods withdrawn, .h159 RATED THERMAL POWER condition) within 7 EFPD af ter reaching an equi-librium boron concentration of 300 ppm. In the event this comparison 7F H6 indicate 5]{y3g MTC is more negative than he 300 ppm surveillance R159 limit specified in the COLR, the MTC shall be remeasured and compared to the E0L MTC limit specified in the COLR at least once per 14 EFPD during the remainder of the fuel' cycle.

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l SEQUOYAH - UNIT 1 3/4 1-5 Amendment No.155 00ff 231991 i

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ADRINISTRATIVf CONTROLS HONTHLY REACTOR OPERATING REPORT ,

.'6,9.1.10 Routine reports of operating statistics and shutdown experience including R76 documentation of all challenges to the PORVs or Safety Valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING L1 HITS REPORT 6.9.1.14, Core operating limits shall be established and documented in the CORE OPERATING LlHlTS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1. Moderator Temperature Coefficient BOL and E0L limits and 300 ppm surveillance R159 limit for Specification 3/4.1.1.3

2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5.

3. Control Bank Insertion Limits for Specification 3/4.1.3.6.

4. Axial Flux Difference Limits for Specification 3/4.2.1.

5. Heat Flux Hot Channel Factor, r,(z), and W(z) for Specification 3/4.2.2, and

6. Nuclear Enthalpy Hot Channel Factor and Power Factor Multiplier for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:

i

> 1. WCAP 9272 P A.

• WESTINGHOUSE RELOAD SAFETY EVALUATION HETH000 LOGY'. July 1985 (W Proprietary).

(Hethodology for Specifications 3.1.1.3 Moderator Temperature Coefficient. 3.1.3.5 Shutdown Bank Insertion Limit. 3.1.3.6Heat Control Flux Bank Insertion Limits. 3.2.1 Axial Flux Difference. 3.2.2 Nuclear Enthalpy Hot Channel Factor,)

, Hot Channel Factor, and 3.2.3

2. WCAP 10216 P A.

• RELAXATION OF CONSTANT AX1AL OFFSET CONTROL F, SURVEILLANCE TECHNICAL SPECIFICATION". JUNE 1983 (V Proprietary).

(Methodology for Specification 3.2.1 Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 Heat Flux Hot Channel Factor (W(z) surveillance requirements for F, Methodology).)

3. WCAP 10266 P A Rev. 2. *THE i981 REVISION OF WESTINGHOUSE EVALUATION H00E USING BASH CODE *, March 1987. (W Proprietary).

(Hethodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

4. WCAP 13631 P A.

• SAFETY EVALUATION SUPPORTING A MORE NEGATIVE EOL HOD git TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE.SEQUOYAH NUCLEAR PLANTS

• HARCH 1993 (H Proprietary).

(Hethodology for Specification 3.1.1.3 - Moderator Temperature Coefficient)

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SEQUOYAH UNIT 1 6 21 Amendment Nos. 52, 58. 72. 74 Occober 26, 1993 117. 152, 155. l 156, 171 )

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CORE OPERAllNG LIMITS REPORT (continued) 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g. , fuel thermal-mechanical limits, core thermal-hydraulic limi'.5, g~

ECCS limits, nuclear limits such as shutdown margin, anc transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c The CORE OPERATING LIMITS REPORT shall be provided within 30 days af ter cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision to the NRC Document Control Desk with ccpics to the Regional Administrator and Resident Inspector.

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SPECIAL REPO R76 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4. i ,

6.9.2.2 Diesel Generator Reliability Improvement Prooram As a minimum the Reliability Improvement Program report for NRC audit, required by LCO 3.8.1.1, Table 4.8-1, shall include:

a summary of all tests (valid and invalid) that occurred within the R56 (a) time period over which the last 20/100 valid tests were performed (b) analysis of f ailures and determination of root causes of failures (c) evaluation of each of the recommendations of HUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Reactors,"

with respect to their application to the Plant (d) identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability (e) the schedule for implementation of each action f rom d) above (f) an assessment of the existing reliability of electric power to engineered-saf ety-feature equipment '

l 6-21a Amendment Nos. 52, 58, 72, f SEQUOYAH - UNIT 1 74, 117, 155 y .

1 October 23, 1991 UL.'.

l Insert A -

Measurement of the MTC in accordance with 4.1.1.3.b may be suspended provided the benchmark criteria and the Revised Prediction as documented in the COLR are satisfied. Data required for the calculation of the Revised Prediction is provided in the Most Negative Temperature Coefficient Limit Report per Specification 6.9.1.15.

Insert B

5. WCAP-13749 P-[A], " SAFETY EVALUATION SUPPORTING THE CONDITIONAL EXEMPTION OF THE MOST NEGATIVE EOL MODERATOR TEMPERATURE COEFFICIENT MEASUREMENT," Muy,1993 (Proprietary)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)

insert C 6.9.1.15 The Most Negative MTC Report shall be prepared at least 60 days prior to the date the limit would become effective and maintained on file. This report willinclude the data required for the determination i

of the Revised Prediction of the 300 ppm /ARO/RTP MTC per WC AP-13749-P-[ A].

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. REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL' R146e~

POWER, after each fuel loading.

b. The MTC shall be measured at any THERMAL POWER and compared to the 1

300 PPM surveillance, limit specified in the COLR (all rods withdrawn, l R146 RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negativ than the 300 PPM surveillance limit specified in the COLR, the TC shall be remeasured R146 and compared to the EOL MlC limit specified i the COLR at least once per 14 EFPD during the remainder of the fuel cycle.

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i SEQUOYAH - UNIT 2 3/4 1-5 Amendment No.146 March 30, 1992

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, ADMINISTRATIVE CONTROLS I

HONTHLY REACTOR OPERATING REPORT 6.9.I.10 Routine reports of operating statistics and shutdown experience, including R64 documentation of all challenges to the PORVs or Safety Valves. shall be submitted on a l monthly basis no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

R146

1. Hoderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

3. Control Bank Insertion Limits for Specification 3/4.1.3.6.

4. Axial Flux Difference Limits for Specification 3/4.2.1.

5. Heat Flux Hot Channel Factor, K(z), and W(z) for Specification 3/4.2.2, and

6. Nuclear Enthalpy Hot Channel Factor and Power Factor Multiplier for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the core operating limits shall .be those previously reviewed and approved by NRC in:

1. WCAP 9272 P A. " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July 1985 (W Proprietary) .

(Methodology for Specifications 3.1.1.3 Moderator Temperature Coefficient 3.1.3.5 Shutdown Bank Insertion Limit, 3.1.3.6 Control Bank Insertion Limits, 3.2.1 Axial Flux Difference, 3.2.2 Heat Flux Hot Channel Factor, and 3.2.3 Nuclear Enthalpy Hot Channel Factor.)

2. WCAP 10216 P A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F, SURVEILLANCE TECHNICAL SPECIFICATION", JUNE 1983 (W Proprietary).

(Hethodology for Specification 3!2.1 Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 Heat Flux Hot Channel Factor (W(z) surveillance requirements for F Methodology).)

3. WCAP 10266 P A Rev. 2. *THE 1981 REVISION OF WESTINGHOUSE EVALUATION HODEL USING BASH CODE", March 1987, (W Proprietary).

(Methodology for Specification 3.2.2 Heat Flux Hot Channel Factor).

4. WCAP 13631 P A. " SAFELY EVALUATION SUPPORTING A MORE NEGATIVE EOL MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SEQUOYAH NUCLEAR R161 PLANTS,* MARCH 1993 (W Proprietary).

(Methodology for Specification 3.1.1.3 Moderator Temperature 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits ECCS limits, R146 nuclear limits such as shutdown margin, and transient and accident analysis limits) of i

I the safety analysis are met.

l SEQUOYAH UNIT 2 6 22 Amendment Nos. 44. 50. 64 Ocotber 26, 1993

66. 107, 134. 146, 161

ADMIN 85TRATIVE CONTROLS  !

1 CORE OPERATING LIMITS REPORT (Continued) 6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days Rl-af ter cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision to the NRC Document Control Desk with copies I to the Regional Administrator and Resident Inspector.

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6.9.2.1 Special reports shall be submitted within the time period specified R64 for each report, in accordance with 10 CFR 50.4.

6.9.2.2 Diesel Generator Reliability Improvement Proaram As a minimum the Reliability I'mprovement Program report for NRC audit, required by LC0 3.8.1.1, Table 4.8-1, shall include:

(a) a summary of all tests (valid and invalid) that occurred within the gt4 time period over which the last 20/100 valid tests were performed (b) analysis of f ailures and determination of root causes of fail.ures (c) evaluation of each of the recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Reactors,"

with respect to their application to the Plant (d) identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability (e) the schedule for implementation of each action from d) above (f) an assessment of the existing reliability of electric power to engineered-safety-feature equipment SEQUOYAH - UNIT 2 6-22a Amendment Nos. 44, 50, 64, 66, 107, 134, 146  ;

March 30, 1992

4 a Insert A Measurement of the MTC in accordance with 4.1.1.3.b may be suspended provided the benchmark criteria and the Revised Prediction as documented in the COLR are satisfied. Data required for the calculation of the Revised Prediction is provided in the Most Negative Temperature Coefficient Limit Report per Specification 6.9.1.15.

Insert B

5. WCAP-13749-P-[A], " SAFETY EVALUATION SUPPORTING THE CONDITIONAL EXEMPTION OF THE MOST NEGATIVE EOL MODERATOR TEMPERATURE COEFFICIENT MEASUREMENT," May,1993 (Proprietary)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)

Insert C 6.9.1.15 The Most Negative MTC Report shall be prepared at least 60 days prior to the date the limit would become effective and maintained on file. This report will include the data required for the determination of the Revised Prediction of the 300 ppm /ARO/RTP MTC per WCAP-13749-P-[ Al.

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1 ENCLOSURE 2 i

PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE i

SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-07)

DESCRIPTION AND JUSTIFICATION FOR TS CHANGE 95 07, 1

" CONDITIONAL EXEMPTION TO END-OF-LIFE MODERATOR l

TEMPERATURE COEFFICIENT SURVEILLANCE MEASUREMENT" b

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. . 1 Descriotion of Chanae l

TVA proposes to modify the Sequoyah Nuclear Plant (SON) Units 1 and 2 technical specifications (TSs) to revise Surveillance Requirement (SR) 4.1.1.3 and I associated administrative specifications to conditionally exempt the end-of-life (EOL) moderator temperature coefficient (MTC) measurement. Accordingly, the following changes are proposed:

1. SR 4.1.1.3, MTC SR - This specification is modified to provide for the suspension of the EOL MTC surveillance measurement in the event specified reactor core model benchmark criteria and a revised EOL MTC prediction are satisfied. Reference is also made to a new most negative moderator temperature coefficient limit report. Please note, theie is a grammatical correction made to the Unit 1 SR.

2. TS 6.9.1.14, Core Operating Limits Report (COLR) - WCAP-13749-P-[A),

l " Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," is added to the list of references.

3. TS 6.9.1.15, Most Negative EOL MTC Limit Report - This new specification provides for the preparation of the most negative EOL MTC limit report. The report will be filed and available for NRC audit.

Reason for Chanae The proposed changes to the TSs are requested because it is believed that relaxation of the existing criteria for performing the MTC measurement near EOL is justifiable while still ensuring the EOL MTC is within the safety analysis limits.

Incorporating these changes would eliminate the surveillance measurement when it is certain from other indicators that the EOL MTC limit would not be exceeded.

The proposed change is desired for the following reasons: (1) The measurement of the EOL MTC requires inducing a temperature transient upon the reactor by deviating from the T y program. Any operation that can be avoided in such an abnormal condition is desirable from a reliability and reactivity management standpoint; (2) The current TS surveillance program requires measurements within seven effective full power days (EFPD) of reaching an equilibrium concentration of 300 parts per million (ppm). Potential regulatory issues would be raised if equilibrium conditions could not be reached within this seven EFPD window because of plant equipment problems or a trip situation; and (3) This measurement is very difficult to make and has the potential for large measurement uncertainties, it is possible that due to the large inherent uncertainty the measurement may not meet the acceptance criteria. In this event, the measurement would be required to be repeated every 14 EFPD for the remainder of the fuel cycle.

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Justificadon for Chanaes For Updated Anal Safety Analysis Report (UFSAR) accident analyses, the transient response of the plant is dependent on reactivity feedback effects, in particular, the moderator dencity coefficient (MDC) and the Doppler power coefficient.

Because of the sensidvity of accident analyses results to the MDC value assumed, it is important that the actual core MDC remain within the bounds of the limiting values assumed in the UFGAR accident analyses. While core neutronics analyses will confirm that the MDC is within these bounds during an operating cycle, SON TSs currently place limits on ths MTC during normal operation. MTC measurements are performed at the beginning of cycle (BOC) prior to initial operation above five percent rated thermal power (RTP) and at RTP conditions within seven EFPD after reaching an equilibrium boron concentration of 300 ppm.

In order to ensure a bounding accident analysis, the MDC is assumed to be at its most limiting value for the analysis conditions appropriate to each accident. The most positive MDC limiting value is based on the EOL core conditions corresponding to maximum fuel burnup and minimum boron concentration assuming 100 percent RTP.

Most accident analyses use a constraint MDC designed te bound the MDC at the worst set of initial conditions as well as at the most limiting set of transient conditions. This value of MDC forms the licensing basis for tra UFSAR accident analysis as well as the bases for the current EOL MTC TS requimments.

Converting the MDC used in the accident analysis to a corresponding MTC requires a calculation, which accounts for the rate of change of moderator density with temperature at the conditions of interest.

TSs place both Limiting Conditions for Operation (LCO) 3.1.1.4 and SR 4.1.n?

constraints on the MTC, based on the accident analysis assumptions of the MDC, The most negative MTC LCO limit applies to Modes 1,2, and 3, and requires that' the MTC be less negative than the specified limit value for the all rods withdrawn, .

end-of cycle, rated thermal power condition. To demonstrate compliance with the most negative MTC LCO, the TS surveillance calls for verification of the MTC After 300 ppm equilibrium boron concentration is obtained. Because the hot full power (HFP) MTC value will gradually become more negative with further core burnup and boron concentration reduction, a 300 ppm SR value of the MTC should necessarily be less negative than the EOL LCO limit.

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To account for this effect, the 300 ppm SR value is sufficiently less negative than (

the EOL LCO limit value, which is specified in the COLR, thereby providing )

assurance that the LCO limit will be met as long as the 300 ppm surveillance criterion is met.

Currently, SON Units 1 and 2 TSs require measurements of MTC at beginning of life (BOL) to verify the most positive MTC limit and near EOL to verify the most negative MTC limit. At BOL, the measurement of the isothermal temperature coefficient and subsequent MTC calculation are relatively simply to perform. The measurement is done at hot zero power isothermal conditions and is not complicated by changes in the enthalpy rise or the presence of xenon. The measurement made ncar EOL differs from the BOL measurement as it is performed at or near HFP conditions. MTC measurements at HFP are more difficult to perform because of small variations in soluble boron concentration and changes in xenon concentration and distribution, fuel temperature, and enthalpy rise created by small changes in the core average power during the measurement. Changes in each of these parameters must be accurately accounted for when reducing the me=urement data, or additional measurement uncertainties will be introduced.

Even though theso additional uncertair. ties may be small, the total reactivity change associated with the swing in moderator temperature is also relatively small. The resulting MTC measurement uncertainty created by even a small change in power level can then become significant and, if improperly accounted for, can yield misleading measurement results.

The method to calculate the revised predicted MTC for determining whether the EOL MTC SR is satisfied is described in Reference 1. If the revised predicted MTC meets the SR then the measurement is not required.

In summary, the conditional exemption from the measurement is sought to improve plant availability and minimize disruptions to normal operation of Units 1 and 2 at SON. As documented in References 1 and 2, it has been ccNluded that plant safety will not be compromised by the conditional exemption ot this measurement.

References

1. Westinghouse Commercial Atomic Power (WCAP) - 13749-P, " Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, May 1993."

2. Letter NTD NRC-95-4384 from Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Activities of Westinghouse Electric Corporation to R. C. Jones, Reactor Systems Branch Chief, Division of Engineering and System Technology of U.S. NRC, dated January 16,1995.

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.4 Environmental Impact Evaluation The proposed change does not involve an unreviewed environmental question because operation of SON Units 1 and 2 in accordance with this change would not:  !

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by NRC's -

testimony to the Atomic Safety and Licensing Board, supplements to the FES, i environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

2. Result in a significant change in effluents or power levels. <

3. Result in matters not previously reviewed in the licensing basis for SON that may have a significant environmentalimpact.

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ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-07) i DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION

Significant Hazards Evaluation TVA has evaluated the proposed technical specification (TS) change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SON) in accordance with the proposed amendment will not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The conditional exemption of the most negative moderator temperature coefficient (MTC) measurement does not change the most negative MTC surveillance requirement (SR) and limiting condition of operation (LCO) limits in the TSs. Since these MTC values are unchanged, and since the basis for the derivation of these values from the safety analysis moderator density coefficient (MDC) is unchanged, the constant MDC assumed for the Updated Final Safety Analysis Report (UFSAR) safety analyses will also remain unchanged. Therefore, no change in the modeling (i.e., probabilities) of the accident analysis conditions or response is necessary in order to implement the change to the conditional exemption methodology. In addition, since the constant MDC assumed in the safety analyses is not changed by the conditional exemption of the most negative MTC SR measurement, the consequences of an accident previously evaluated in the UFSAR are not increased. The dose predictions presented in the UFSAR for a steam generator tube rupture remain valid such that more severe consequences will not occur. Additionally, since rnass and energy releases for a loss-of-coolant accident and a steamline break are not increased as a result of the unchanged MDC, the dose predictions for these events presented in the UFSAR also remain bounding.

2. Create the possibility of a new or different kind of accident from any previously analyzed.

Since the end-of-life MTC is not changed by the conditional exemption methodology of WCAP-13749-P, the possibility of an accident, which is different than any already evaluated in the UFSAR, has not been created. No new or different failure modes have been defined for any system or component nor has any new limiting single failure been identified.

Conservative assumptions for the MDC have already been modeled in the l UFSAR analyses. These assumptions will remain valid since the conditional i exemption methodology documented in WCAP-13749-P does not change the safety analysis MDC nor the TS values of the MTC.

3. Involve a significant reduction in a margin of safety.

The conditional exemption methodology is documented in WCAP-13749-P.

This WCAP has been evaluated (

Reference:

SECL 93-117,R1) relative to the design basis, including the TSs, and has been determined to bound the conditions under which the specifications permit operation. The results as presented in the UFSAR remain bounding since the MDC assumed in the safety analyses and the limiting conditions for operation and SR MTCs in the TSs remain unchanged. Therefore, the margin of safety, as defined in the bases to these TSs, is not reduced.

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. '. l ENCLOSURE 4 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-95-07)

CORE OPERATING LIMIT REPORT MARKUPS l

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L3 6 ,,9 4 0 919 802 SEQUOYAH NUCLEAR PLANT UNIT 1, CYCLE 7 CORE OPERATING LIMITS REPORT REVISION 3 September 6, 1994 Prepared:

( . w / ~~ b ~ 9 Y Date Nuclear Fuel)

Reviewed:

Es hs Mk /, 9-9-94 Date React ~or Engineering Supervisor, J $' Y Tectini al Support Manager ' Date

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Date PORC thai an Revision 3_

Pages affected _all_

Reason for Revision Add revised W(z) functions and correspondina reduced axial flux difference limits.

Revised Ficure 1 to recuire fu11v vithdrawn position

> 226 for evele burnuo to 9.000 mwd /MTU.

SEQUOYAH - UNIT 1 Page 1 of 21 Revision 3 kh bODD$W

s COLR FOR SEQUOYAH UNIT 1 CYCLE 7 2.1.2 The 300 ppm surveillance limit is:

The measured 300 ppm /ARO/RTP-MTC should -4 be less F.

Ak k negative than or equal to -3.75 x 10 ~

.D . . -Wa .. -. s ~.

, h/ s.c.ie.__n r (Specification 3/4.1. 3. 5)

_ shutdown Rod Insertion Limit 2.2 2.2.1 The shutdown rods shall be withdrawn to a position as defitwa below:

Steos Withdrawn Cvele Burnuo (MWD /MTUL S 9,000 1 226 to i 231

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> 9,000 to <.14,000 2 222 to S 231 2 14,000 2 226 to i 231 (Specification 3/4.1.3. 6) 2.3 Control Rod Insertion Limits 2.3.1 The control rod banks shall be limited in physical insertion as shown in Figure 1 2.4 Agial Flux Difference - AFD (Specification 3/4.2.1) 2.4.1 The axial flux difference (AFD) limitsfunctions provided in s

Figure 2 shall be used when the W(z) provided in Figures 5-9 are used.

The axial flux difference (AFD) limits provided in functions Figure 3 shall be used when the W(z)

• provided in Figures 10-17 are used.

(Specification 2.5 Heat Flux Hot Channel Factor - Fgi;l;.).

3/4.2.2) .

FqRTP

• K(z) for P > 0.5 Fq (z) $-

pqRTP

• K(z) for P $ O.5 Fo(z) $--0.5 Page 3 of 21 Revision 3 SEQUOYAH - UNIT 1

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'2.1.3 The EOL MTC Revised Prediction shall be calculated from the algorithm defined in Table 3 3 of Reference 5.in Technical Specification 6.9.1.14.a. The MTC data required for this calculation shall be provided in a Most Negative Moderator Temperature Coefficient Limit Report per Technical Specification 6.9.1.15. If the Revised Predicted MTC is less negative than the surveillance requirement of -3.75 x 10* 6k/k/ F, and all benchmark criteria listed in Table 3-2 of Reference 5 are met, then a measurement is not required per Technical Specification 4.1.1.3.b.

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, L3 6,,9 4 0 816 801 SEQUOYAH NUCLEAR PLANT UNIT 2, CYCLE 7 CORE OPERATING LIMITS REPORT REVISION O August 12, 1994 Prepared:

W .

/ $ ~l2. - 9 Y Nuclear Fueu) Date Reviewed:

W A <.& & $~ / 8-/4- 9'/

React $r Engineering Supervisor Date

/ Eh2/Th

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Technica Support Manager Date PORC N rman

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i FA,19Y hatd Revision O Pages affected Reason for Revision 1

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SEQUOYAH - UNIT 2 Page 1 of 16 Revision 0 l l

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. i COLR.FOR SEQUOYAH UNIT 2 CYCLE 7 2.1.2 The 300' ppm surveillance limit is: ,

The measured 300 ppm /ARO/RTP-MTC should -4 be'less Ak/k/*F.

% negative than or equal - v to -3.75w~x 10 '

i ir/ S CotT* }

.- 2-(Specification 3/4.1. 3. 5) l 2 .'2 Shutdown Rod Tnsertion Limit l

2.2.1 The shutdown rods ~ shall be withdrawn to a position -  ;

as defined below:

' Steps Withdrawn ,

cvele Burnuo (MWD /MTQ)_ t d 4,000- 1 226 to $ 231' ,

> 4,000'to < 14,,000 2 222 to 5 231 2 14,000 2 226 to.S'.23'1 ,

t 2.3 control Rod Insertion-Limits (Specification 3/4.1.3.6) 2.3.1 The control rod banks shall be limited in physical ,

insert $ m as shown in Figure 1. f e

A 2.4 Aydal Flux Dif f erence - AFD (Specification 3/4.2.1) ,

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2.4.1 The axial flux difference (AFD) limits are orovided in Figure 2.

Heat Flux Hot Channel Factor - Fg(z) (Specification 2.5 ,

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• K(z) for P > 0.5 P 1 y qRTP Pg(z) 5

• K(z) for P 5 0. 5 THERMAL POWER where P =

RATED THERMAL POWER Page 3 of 16 Revision 0 SEQUOYAH - UNIT 2 I

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2.1.3 The EOL MTC Revised Prediction shall be calculated from the algorithm defined in Table 3 3 of Reference 5 in Technical l

Specification 6.9.1.14.a. The MTC data required for this calculation - "

shall be provided in a Mo'st Negative Moderator Temperature _  ;

Coefficient Limit Report per Technical Specification 6.9.1.15. If the' Revised Predicted MTC is less negative than..the surveillance.

requirement of -3.75 x 10'* 6k/k/*F, and all benchmark criteria listed in Table 3-2 of Reference' 5~are met, then a measurement is not' required per Technical Specification 4.1.1.3.b.

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ENCLOSURE 5 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS l' AND 2.

DOCKET NOS 50-327 AND 50-328 (TVA-SON-TS-95-07)

WESTINGHOUSE SAFETY EVALUATION CHECK LIST 93-117,R1 I

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. '. l SECL-93-117, R1 Customer Reference No(s).

Westinghouse Reference No(s).

WESTINGHOUSE SAFETY EVALUATION CHECK LIST

1) NUCLEAR PLANT (S): Seouovah Units 1 and 2

) CHECK LIST APPLICABLE TO: Conditional Exemotion from Measurement of EOL MTC

3) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.59(b) has been prepared to the extent required and is attached. If a safety evaluation is not required or is incomplete for any rearon, explain on Page 2. Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed.

CHECK LIST - PART A - 10CFR50.59(a)(1) 3.1) Yes _ No_X. A change to the plant as described in the FSAR?

3.2) Yes _ No_X_ A change to procedures as described in the FSAR?

3.3) Yes _ No_X_ A test or experiment not described in the FS AR?

3.4) Yes _X _ No _._ A change to the plant technical specifications Appendix A to the Operating License)?

4) CHECK LIST - PART B - 10CFR50.59(a)(2) (Justification for Part B answers must be included on page 2.)

4.1) Yes __._ No_X_ Will the probability of an accident previously evaluated in the FSAR be increased?

4.2) Yes _ No_X_ Will the consequences of an accident previously evaluated in the FSAR be increased?

4.3) Yes _ No_X_ May the possibility of an accident which is different than any already evaluated in the FSAR be created?

4.4) Yes __ No_X_ Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

4.5) Yes _ No.X_ Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

4.6) Yes _ No _X_ May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?

4.7) Yes _ No.X_ Will the margin of safety as described in the bases to any technical specification be reduced?

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If the answer to any of the above questions is unknown, indicate under 5.) REMARKS and l explain below.

If the answer to any of the above questions in Part A (3.4)'or Part B cannot be answered in the negative, based on written safety evaluation, the change review would require an application for license amendtnent as required by 10CFR50.59(c) and submitted to the NRC pursuant to 10CFR50.90.

5) REMARKS:

The answers given in Section 3, Part A, and Section 4, Part B, of the Safety Evaluation Checklist, are based on the attached Safety Evaluation.

FOR FSAR UPDATE Section: N/A Pages: Tables: Figures:

Reason for / Description of Change: N/A 6.0 APPROVAL LADDER Nuclear Safety Preparer:

g O. > - ~ s L.V. Tomasic Date:

9.u.9r Nuclear Safety Reviewer: T.LK ch(n V {#

h Date: Y /( I I

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. SECL 93117, R1 Sequoyah Nuclear Plant, Units 1 and 2 Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement Safety Evaluation

1.0 BACKGROUND

Revision I removes proprietary material from this safety evaluation.

For FSAR accident analyses, the transient response of the plant is dependent on reactivity feedback effects, in particular, the moderator density coefficient (MDC) and the Doppler power coefficient.

Because of the sensitivity of accident analysis results to the MDC value assumed, it is important that the actual core MDC remain within the bounds of the limiting values assumed in the FSAR accident analyses. While core neutronics analyses will have confirmed that the MDC is within these bounds, the Technical Specifications also place limits on the moderator temperature coefficient (MTC) that can be obtained during normal operation. MTC measurements are performed at the beginning of the cycle prior to initial operation above $% rated thermal power. Most plants also currently have a requirement to measure the MTC at rated thermal power conditions within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.

In order to ensure a bounding accident analysis, the MDC is assumed to be at its most negative value for the analysis conditions appropriate to each accident. The most positive MDC limiting value is based on the end of life (EOL) core conditions corresponding to maximum fuel burnup and minimum boron concentration assuming 100% rated thermal power. Two different Technical Specification bases relating the accident analysis MDC to the most negative MTC have been previously licensed for Westinghouse plants as described in Chapter 2 of Reference 1.

Most accident analyses use a constant MDC designed to bound the MDC at the worst set of initial conditions as well as at the most limiting set of transient conditions. This value of MDC forms the licensing basis for the FSAR accident analysis as well as the bases for the current EOL MTC Technical Specification requirements. Converting the MDC used in the accident analyses to a corresponding MTC is a simple calculation which accounts for the rate of change of moderator density with temperature at the conditions of interest.

Technical Specifications place both Limiting Condition for Operation (LCO) and Surveillance Requirement (SR) constraints on the MTC, based on the accident analysis assumptions of the MDC.

De most negative MTC LCO limit applies to Modes 1,2, and 3, and requires that the MTC be less negative than the specified !!mit value for the all rods withdrawn, end of cycle life, rated thermal power condition. To demonstrate compliance with the most negative MTC LCO, the Technical Specification SR calls for verification of the MTC after 300 ppm equilibrium boron concentration is obtained. Because the HFP MTC value will gradually become more negative with further core burnup and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOL LCO limit. To account for this effect, the 300 ppm SR value is sufficiently less negative than the EOL LCO limit value, providing assurance that the LCO limit wi31 be me't as long as the 300 ppm surveillance criterion is met.

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SECL-93-l!7, RI Currently, the Technical Specifications require measurements of MTCs at BOL to verify the most positive MTC limit and near EOL to verify the most negative MTC limit. At BOL, the measurement of the isothermal temperature coefficient is relatively simple to pert'orm since it is done at hot zero power isothermal conditions and is not complicated by changes in the enthalpy rise or the presence of xenon. The measurement made near EOL differs from the BOL measurement as it is performed at or near hot full power conditions. MTC measurements at HFP are more difficult to perform due to small variations in soluble boron concentration and changes in xenon concentration and distribution, fuel temperature, and enthalpy rise created by small changes in the core average power during the measurement. Changes in each of these parameters must be accurately accounted for when reducing the measurement data or additional measurement uncertainties will be introduced. Even though these additional uncertainties may be small, the total reactivity change associated with the swing in moderator temperature is also relatively small. De resulting MTC measurement uncertainty created by even a small change in power level can then become significant and, if improperly accounted for, can yield misleading measurement results.

The method to calculate the revised predicted MTC for determining whether the EOL MTC SR is satisfied is described in Reference 1. If the revised predicted MTC meets the SR then the measurement is not required.

The purpose of this safety evaluation is to provide justification that the conditional exemption from the measurement of the 300 ppm SR MTC does not represent an Unreviewed Safety Question.

2.0 LICENSING BASIS The change to the MTC LCO and SR Technical Specifications and COLR sections is evaluated with respect to the acceptance criteria for the accidents addressed in Chapter 15 of the FSAR.

I 3.0 EVALUATION 3.1 Nuclear Design Evaluation ne Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement methodology is described in detail in Reference 1. De Technical Specification Bases of the most negative MTC LCO and SR and the values of these limits are not altered. Instead, a revised prediction is compared to the SR MTC to determine if the SR limit is met. The method for )

calculating the revised prediction is described in Reference 1. i 1

3.2 Impact on Safety Analysis ne safety analysis assumption of a constant MDC, and the actual value assumed, will NOT change.

Therefore, a conditional exemption from the HFP near-EOL 300 ppm MTC measurement based on I the margin to the SR limit will NOT have an adverse impact on the safe operation of the plant.

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, ,1 s 3.3 Impact on Technical Specifications The proposed Technical Specification SR (i.e. Specification 4.1.1.3) and COLR changes required to support the conditional exemption are described in Reference 1. Rese Technical Specification changes will NOT change the LCO or SR MTC limits (i.e. Specification 3.1.1.3). De Bases describing the derivation of these MTC limits from the safety analysis MDC will also remain unchanged.

He specific values of the SR and LCO MTCs remain unchanged by the conditional exemption methodology documented in Reference 1. One additional Technical Specification will be added (i.e.

6.9.1.15) which will define the Most Negative Moderator Temperature Coefficient Limit Report.

His cycle-specific report will provide the information required for calculating the margin to the SR limit. Reference I will also be added to the list of references in Technical Specification 6.9.1.14.

3.4 Impact on Other Safety Related Areas ne following safety related areas are not impacted by the conditional exemption of the HFP 300 ppm MTC:

- mechanical and fluid systems instrumentation and controls

- containment analysis

- radiological consequences

- non-LOCA analysis

- LOCA and LOCA related analyses, including Large Break and Small Break LOCA hydraulic forces, post-LOCA suberiticality, and hot leg switchover

- steam generator tube rupture

- probabilistic risk assessment

- emergency operating procedures

- protection system setpoints 4.0 DETERMINATION OF NO UNREVIEWED S AFETY QUESTION ne evaluation presented above forms the basis upon which specific responses to the questions in Section 4 of the Checklist can be provided.

1. Will the probability of an accident previously evaluated in the FSAR be created ?

No. The conditional exemption of the most negative MTC measurement does not change the most negative MTC SR and LCO limits in the Technical Specifications. Since these MTC values are unchanged, and since the basis for the derivation of these values from the safety analysis MDC is unchanged, the constant MDC assumed for the FSAR safety analyses will also remain unchanged. Herefore, no change in the modeling of the accident analysis conditions or response is necessary in order to implement the change to the conditional exemption methodology.

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2. Will the consequences of an accident previously evaluated in the FSAR be increased :

No. Since the constant MDC assumed in the safety analyses is not changed by the conditional exemption of the most negative MTC SR measurement, the consequences of an accident i previously evaluated in the FSAR are not increased. The dose predictions presented in the i FSAR for a SGTR remain valid such that more severe consequences will not occur.

Additionally, since mass and energy releases for LOCA and steamline break are not increased as a result of the unchanged MDC, the dose predictions for these events presented in the FSAR l also remain bounding.

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3. May the possibility of an accident which is different than any already evaluated in the FSAR be created 7 1

l No. Since the EOL MTC is not changed by the conditional exemption methodology of l

Reference 1, the possibility of an accident which is different than any already evaluated in the i FSAR has not been created. No new failure modes have been defined for any system or component not bas any new limiting single failure been identified. Consetvative assumptions for MDC have already been modeled in the FSAR analyses. These assumptions will remain valid since the conditional exemption methodology documented in Reference 1 does not change the safety analysis MDC nor the Technical Specification values of the MTC. l l

4. Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be created ?

No. He probability of a malfunction of equipment important to safety will not increase due to the coLditional exemption methodology documented in Reference 1. As stated previously, component and system performance will not be adversely affected since the methodology in Reference 1 does not change the MDC assumed as part of the current analyses of record.

5. Will the consequences of a malfunction of equipment important to safety previously evaluatM; in the FSAR be increased ?

No.ne consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not increase since the MDC assumed in the safety analyses remains unchanged by the conditional exemption methodology of Reference 1. No malfunction of equipment other than those currently defined in the FSAR are expected as a result of the conditional exemption from the tnost negative MTC measurement. The evaluations performed for SGTR, steamline break and LOCA have confirmed that the dose predictions presented in the FSAR remain bounding. He acceptance criteria for radiological consequences continue to be met.

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6. May the possibility of malfunction of equipment important to safety different than any already  :

evaluated in the FSAR be created ?

No.ne conditional exemption methodology of Reference I will not create the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR.

All original design and performance criteria continue to be met such that there is no unanticipated malfunction of equipment expected.

7. Will the margin of safety as slescribed in the bases to any Technical Specification be reduced '

No, ne evaluation of the conditional exemption methodology documented in Reference I has taken into account the applitable Technical Specifications and has bounded the conditions under which the Specifications peruit operatior, he applicable Technical Specifications are the Surveillance specification 4.1.!.3 and for those plants with a COLR, Reference 1 is added to the list of references in Specificailon 6.9.1.14a. An additional Specification 6.9.1.15 is added to define the requirements for the Most Negative Moderator Temperature Coefficient Limit Report, which is described in Appendices A, B and D of Reference 1. The COLR has also been modified as described in Appendix B of Reference 1. The analyses which support these Technical Specifications have been evaluated. He results as presented in the FSAR remain bounding since the MDC assumed in the safety analyses and the LCO and SR MTCs in the Technical Specifications remain unchanged. Therefore, the margin of safety, as defined in the bases to these Technical Specifications, is not reduced.

5.0 CONCLUSION

It is concluded that operation with a conditional exemption from the most negative MTC measurement as described in Reference 1 and this evaluation does not involve an unreviewed safety question as defined in 10 CFR 50.59.

6.0 REFERENCES

1. WCAP-13749, " Safety Evaluation Supporting The Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement", May,1993.

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