ML13331A973
| ML13331A973 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/31/1987 |
| From: | Baskin K, Holcombe J San Diego Gas & Electric Co, Southern California Edison Co |
| To: | |
| Shared Package | |
| ML13331A970 | List: |
| References | |
| NUDOCS 8709080406 | |
| Download: ML13331A973 (15) | |
Text
BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDISON
)
COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY )
for a Class 104(b) License to Acquire,
)
DOCKET NO. 50-206 Possess, and Use a Utilization Facility as
)
Part of Unit No. 1 of the San Onofre Nuclear )
Amendment No. 143 Generating Station
)
SOUTHERN CALIFORNIA EDISON COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY, pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 143.
This amendment consists of Proposed Change No. 170 and 171 to Provisional Operating License No. DPR-13. Proposed Change Nos. 170 and 171 modify the Technical Specifications incorporated in Provisional Operating License No. DPR-13 as Appendix A.
Proposed Change No.170 is a request to revise Appendix A Technical Specification 3.9, "Core Average Burnup," to be a Moderator Temperature Coefficient based specification. This change will revise this limiting condition for operation to be based directly upon the safety parameter that the core burnup specification was designed to limit.
8709080406 870831 PDR ADOCK 05000206 PDR
-2 Proposed Change No. 171 revises Appendix A Technical Specifications 3.10, "Incore Instrumentation," and 3.11, "Continuous Power Distribution Monitoring," to incorporate more frequent correlation verification of the excore axial offset monitoring instrumentation and revise the formula for determining incore axial offset. These changes will facilitate a closer surveillance of incore flux distribution conditions.
In the event of conflict, the information in Amendment Application No. 143 supersedes the information previously submitted.
Based on the safety analysis provided in the Description of Proposed Change and Safety Analysis, it is concluded that (1) the proposed change does not involve an unreviewed safety question as defined in 10 CFR 50.59, nor does it present significant hazards considerations not described or implicit in the Final Safety Analysis, and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.
Pursuant to 10 CFR 170.12, the fee of $150 is herewith remitted.
-3 Subscribed on this JV day of
, 1987.
Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY By:
Kenneth P. Baskin Vice President Subscribed and sworn o before me this 7/4 day of 15' ff OFFICIAL SEAL AGNESGCRABTREE Notary Public-California LOS ANGELES COUNTY My Comm. Exp. Sep. 14, 1990 Not Public in and for the County of Los Angeles, State of California My Commission Expires:
y f/
/ff Charles R. Kocher James A. Beoletto Attorneys for Southern California Edison Company By:
Jami Beb etto\\
-4 Subscribed on this @Q day of 1987.
Respectfully submitted, SAN DIEGO GAS & ELECTRIC COMPANY C. Holcombe Vice President Subscribed and sworn to before me this
() 4ay of
//
J.4 4.
No y
Publ ic 0and or County of S iego, St fe of Cali-fdnia My Commission Expires:-& "-
OFFCUAL.SEAL JILL QUIGLEY NOTARY PUBUC-CAUFORNIA David R. Pigott PRINCIPAL OFFICE IN Samuel B. Casey SN M DIExGOp 1989 Orrick, Herrington & Sutcliffe MyCommiionExp.Mrch7,1989 Attorneys for San Diego Gas & Electric Company By:
David R. Pigott
I*
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of SOUTHERN
)
CALIFORNIA EDISON COMPANY
)
and SAN DIEGO GAS & ELECTRIC
)
Docket No. 50-206 COMPANY (San Onofre Nuclear
)
Generating Station Unit No. 1 CERTIFICATE OF SERVICE I hereby certify that a copy of Amendment Application No. 143 was served on the following by deposit in the United States Mail, postage prepaid, on the 31st day of August
, 1987.
Henry 3. McGurren, Esq.
Staff Counsel U.S. Nuclear Regulatory Commission Washington, D.C.
20545 David R. Pigott, Esq.
Samuel B. Casey, Esq.
Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 L. G. Hinkleman Bechtel Power Corporation P.O. Box 60860, Terminal Annex Los Angeles, California 90060 Michael L. Mellor, Esq.
Thelen, Marrin, Johnson & Bridges Two Embarcadero Center San Francisco, California 94111 Huey Johnson Secretary for Resources State of California 1416 Ninth Street Sacramento, California 95814 Janice E. Kerr, General Counsel California Public Utilities Commission 5066 State Building San Francisco, California 94102
-2 James McGuffin Western Regional Manager Westinghouse Electric Corporation Post Office Box 2728 Pittsburgh, Pennsylvania 15230 A. I. Gaede 23222 Cheswald Drive Laguna Niguel, California 92677 Frederick E. John, Executive Director California Public Utilities Commission 5050 State Building San Francisco, California 94102 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 am"Jes A. Be YttoN
DESCRIPTION OF PROPOSED CHANGE AND SAFETY ANALYSIS OF PROPOSED CHANGE NO. 170 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE DPR-13 This is a request to revise Section 3.9, "CORE AVERAGE BURNUP" of Appendix A Technical Specification for San Onofre Nuclear Generating Station, Unit 1.
DESCRIPTION Specification 3.9, "CORE AVERAGE BURNUP" serves to establish a core burnup limit for San Onofre Nuclear Generating Station, Unit 1 (SONGS 1) core.
The intention is to maintain the core within the end-of-life (EOL) design limits assumed in the safety analyses. The EOL design limits provide assumed bases for determining acceptable results for certain safety analyses.
However, since the moderator temperature coefficient (MTC) becomes more negative as the fuel accumulates burnup and it is the assumed MTC that is the limiting safety analyses assumption that restricts core average burnup, restricting core average burnup is tantamount to restricting the MTC.
Accordingly, this proposed change would provide assurance that the most negative moderator temperature coefficient (MTC) used in supporting accident analyses is not exceeded during core operations at high fuel burnups.
The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the accident and transient analyses. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. The specification will require that the MTC be compared to a predetermined calculated value that has considerable margin to the MTC assumed in the limiting steam break analysis. However, in no case will the safety analysis assumption be exceeded. In order to ensure this, in the event the predetermined calculated MTC value is exceeded, the MTC is to be remeasured and compared to the EOL limit of -3.8 x 10-4 Ak/k/OF every 14 effective full power days (EFPD) during the remainder of the fuel cycle.
This EOL MTC limit is conservative in that it is less negative than the
-4 x 10-4 Ak/k/*F assumed in the safety analyses. In the event that the EOL MTC limit is exceeded, the reactor will be shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of exceeding the limit.
Consistent with the above discussed changes, the title of the specification is accordingly retitled "MODERATOR TEMPERATURE COEFFICIENT."
This change is administrative in nature and does not impact any safety analyses.
EXISTING TECHNICAL SPECIFICATION See Attachment 1 PROPOSED TECHNICAL SPECIFICATION See Attachment 2
(1)
-2 SAFETY EVALUATION The proposed change as discussed above shall be deemed to constitute a significant hazard consideration if positive findings are made in any of the following areas:
- 1.
Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No During the reload design and evaluation for each cycle, all appropriate safety parameters are evaluated and reviewed for acceptability. The proposed changes to the Technical Specifications described herein are set in consideration of the safety evaluation for the current and subsequent reload cycles of San Onofre Unit 1. These evaluations have demonstrated that core design and safety limits will be satisfied, assuming the requested MTC technical specification change is approved and the facility is operated in accordance with this limiting condition for operation.
The previous specification restricted core average burnup in an attempt to ensure that an MTC assumed in the steam line break analysis was not exceeded. The proposed specification will directly restrict the MTC, irregardless of core average burnup, in order to assure the integrity of the safety analysis.
Since the MTC limit is the same -3.8 x 10 ak/k/oF that was assumed in the basis of the existing technical specification, there is no increase in the probability or consequences of previously evaluated accidents. Therefore, it is concluded that operation of the facility in accordance with this proposed change will not increase the probability or consequences of an accident previously evaluated.
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No As stated above, all appropriate safety parameters have been and will be evaluated during the reload design and evaluation for the current and subsequent cycles. Operation of the facility in accordance with this proposed change assures that the plant will remain within the assumptions of the safety analyses. Therefore, it is concluded that this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
-3 Response: No Operation of the facility in accordance with this proposed change will not cause reduction in a margin of safety, since measurements of the MTC during reactor operations will assure that the MTC will be less negative than a predetermined calculated MTC value, but in no case will the MTC be allowed to be more negative than the -4.0 x 10-4 Ak/k/oF limit used in the accident analyses for San Onofre Unit 1. In order to provide an added margin of safety, the limiting MTC that requires plant shutdown is
-3.8 x 10-4 Ak/k/*F. Thus, the plant will be assured of being within its assumed margin of safety. Therefore, it is concluded that this proposed change will not result in a significant reduction in a margin of safety.
The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards cohsiderations. This proposed change is most similar to example (vii) related to a change to make a license conform to changes in the regulations where the license change results in a very minor change to facility operations clearly in keeping with the regulations. In this case, the change will allow for the monitoring of a parameter associated with the core burnup, rather than the burnup itself.
Since this parameter (MTC) is the important value for safety analyses and it is currently used in the latest revision of the STS, this is an appropriate revision.
The proposed change to the title and the basis are most similar to example (I) of 48 FR 14870 since they are purely administrative changes to achieve consistency throughout the Technical Specification. In this case, the revision is necessary to be consistent with the changes to the body of the specification.
SAFETY AND SIGNIFICANT HAZARDS DETERMINATION Based on the safety evaluation, it is concluded that: (1) the proposed change does not involve a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Environmental Statement. - Existing Specification Section 3.9 -
Proposed Specification Section 3.9 3.9 CORE AVERAGE BURNUP 2
Applicability:
Applies to average burnup of the core.
11/13/70 Objective:
To establish an average burnup limit for the core.
Specification:
The average burnup of the core shall not exceed 21,000 MWD/MTM.
20 Basis:
To ensure that the cold water accident analyses and the 2/18/75 steam line break accident analysis remain applicable for 8
Cycle 4 and succeeding cycles the core average burnup should 4/13/73 not exceed 21,000 MWD/MTM. The moderator temperature coefficient becomes more negative with increasing burnup and 20 a more negative temperature coefficient would increase the 2/18/75 severity of the consequences from cold water accidents.. The analyses of the consequences of cold water accidents use a moderator temperature coefficient of reactivity of -4.Oxl0 4 Ak/koF.
In order to provide a margin of safety, the 8
reactor should not be operated with a temperature 4/13/73 coefficisnt of reactivity more negative than -3.8xl0~
4 Ak/k/ F.
Cycle 4 analyses show that the temperature coefficignt of reactivity will be less negative than
-3.8xl0 Ak/k F at 21,000 MWD/MTM.
The limiting reactivity coefficient used in the steam line break accident analysis is given as a function of keff and average moderator temperature in Figure 14 of Amendment 18 to the FSAR. In order to ensure that the safety analysis remains valid, the reactor should not be operated with a reactivity coefficient more negative.than the limit implied by Figure 14 of Amendment 18.
Cycle analyses show that the overall reactivity coefficient will not be more negative 20 than the specified limit up to a core-average burnup of 2/18/73 21,000 MWD/MTM.
References:
(1) Final Engineering Report and Safety Analysis, Paragraph 3.4.
(2) Supplement No. I to Final Engineering Report and Safety Analysis Section 5, Question 8 and 9.
(3) Final Engineering Report and Safety Analysis, Paragraph 3.9.
3-90 Revised:
10-30-78 3.9 MODERATOR TEMPERATURE COEFFICIENT (MTC)
APPLICABILITY:
Applies to negative moderator temperature coefficient (MTC) during core operations in MODES 1, 2 and 3.
OBJECTIVE:
To establish negative MTC limits for the core.
SPECIFICATION:
- a.
The MTC shall be less negative than - 3.8 x 10-4 Ak/k/oF for all rods withdrawn, end of cycle life (EOL) and the RATED THERMAL POWER condition.
- b. In order to assure that the above negative MTC limit is not exceeded, the MTC shall be measured at any THERMAL POWER and extrapolated to the RATED THERMAL POWER value to be compared to the predetermined, calculated negative MTC within 7 effective full power days (EFPD) of reaching an equilibrium boron concentration of 300 ppm.
The predetermined calculated MTC value (at RATED THERMAL POWER conditions with all rods withdrawn) is determined as follows:
-3.1 x 10-4 Ak/k/oF MTC at a nominal core average coolant temperature of 575.0oF, and MTC increasing linearly with decreasing average coolant temperature to -2.5 x 10-4 Ak/k/oF at a nominal core average coolant temperature of 551.5 0F.
ACTION:
In the event the comparison of specification b indicates the MTC is more negative than the applicable value given above, the MTC shall be remeasured, and compared to the EOL MTC limit of -3.8 x 10-4 Ak/k/oF at least once per 15 EFPD during the remainder of the cycle. If the measured MTC is more negative than the -3.8 x 10-4 Ak/k/oF limit any time during the remainder of the cycle, the reactor shall be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after exceeding the negative MTC limit.
BASIS:
The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the San Onofre Unit 1 accident and transient analyses.
The limiting MTC used in the steam line break accident analysis is given as a function of keff and average moderator temperature in Figure 14 of Amendment 18 to the FSAR. In order to ensure that that safety analysis remains valild, the reactor should not be operated with a MTC more negative than the limit implied by Figure 14 of Amendment 18.
The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated above will require extrapolation to those conditions in order to permit an accurate comparison.
S
-2 The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the San Onofre Unit 1 analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition, and a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value
-4.0 x 10-4 sk/k/0 F. The MTC predetermined calculated value for surveillance purposes represent conservative values (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and are obtained by making these corrections to the limiting MTC value of -4.0 x 10-4 Ak/k/oF. In order to provide a margin of safety, the reactor should not be operated with an MTC more negative than -3.8 x 10-4 Ak/k/*F.
The measurement of the MTC at the end of the fuel cycle is adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in boron concentration associated with fuel burnup.
References:
(1) Final Engineering Report and Safety Analysis, Paragraph 3.4 (2) Supplement No. 1 to Final Engineering Report and Safety Analysis Section 5, Question 8 and 9.
(3) Final Engineering Report and Safety Analysis, Paragraph 3.9.
DESCRIPTION OF PROPOSED CHANGE AND SAFETY ANALYSIS OF PROPOSED CHANGE NO. 171 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE DPR-13 This is a request to revise Section 3.10, "INCORE INSTRUMENTATION" and 3.11 "CONTINUOUS POWER DISTRIBUTION MONITORING" of Appendix A Technical Specifications for San Onofre Nuclear Generating Station, Unit 1.
DESCRIPTION Technical Specifications 3.10 and 3.11 relate to the incore instrumentation and limits of axial offset for power operation. The instrumentation and limits are necessary to assure that the core physics are maintained in a manner consistent with the assumptions in the applicable safety analyses.
Proposed Change No. 171 would revise these two sections of the technical specifications to provide added assurance that the plant is operated within the assumptions of the safety analyses.
Proposed Change No. 171 would revise Specification 3.10 to require a fixed interval for the correlation verification of incore versus excore data.
The proposed frequency of every 180 effective full power days (EFPD) provides greater assurance that the core is being operated in a manner consistent with that assumed in the fuel cycle reload analysis. This revision specifies the frequency of correlation verification, and in between the verifications, the addition of a new term (FCC) in 3.11 that will account for any differences encountered in the monthly correlation check by restricting core operating limits. The addition of the new term in 3.11 allows for the replacement of the reference to the + 3 percent restriction of the incore axial offset, since the differences may vary according to the latest correlation check.
- However, the increased mandatory correlation verification frequency will assure that the FCC term does not mask actual changes in core physics characteristics due to the accumulation of burnup. Additional revisions to 3.10 are proposed to delete part E, which applies to a previous operating cycle and is no longer necessary. The specification of a mode applicability of "MODE 1 above 90%
RATED THERMAL POWER" and a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action time to reduce power merely codifies what has been the general operating practice. This is acceptable since the current technical specification states that Mode 1 operation below 90% power is acceptable. Revisions are also proposed to the structure of the specification to make the structure of 3.10 consistent, to the extent possible, with the STS.
Proposed Change No. 171 would revise Specification 3.11 to include a new incore axial offset equation that has a variable factor to account for statistical variations that could artificially alter the incore-excore correlation. However, in no case will the proposed term FCC be allowed to be less than the value of 3.0, to preserve the assumptions made in the design analyses. The specification of a mode applicability of "MODE 1 above 90%
RATED THERMAL POWER" and a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action time to reduce power merely codifies what has been the general operating practice.
- g g
-2 This is acceptable since the current Technical Specification states that Mode 1 operation below 90% is acceptable. Additional revisions to 3.11 are proposed to restructure the specification in a manner consistent with the Standard Technical Specification. The proposed revisions to the basis of 3.11 are necessary to describe the basis for the proposed factor FCC.
EXISTING TECHNICAL SPECIFICATION See Attachment 1 PROPOSED TECHNICAL SPECIFICATION See Attachment 2 SAFETY EVALUATION The proposed change as discussed above shall be deemed to constitute a significant hazard consideration if positive findings are made in any of the following areas:
- 1.
Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated.
Response: No The proposed changes to 3.10 would require that the correlation verification be performed at a more frequent interval.
The revision to the frequency will provide greater assurance that the calculated incore axial offset reflects actual core conditions and not errors in detector output. The addition of the FCC terms to the incore axial offset equation of 3.11 will account for detector drift, but in no case will the Incore Axial Offset equation allow a smaller value than was mandated by the existing specifications.
Therefore, these changes merely allow for a more accurate measurement of core operating parameters and do not alter any accident probabilities and/or consequences.
The remaining proposed changes to 3.10 and 3.11 are administrative in nature, and do not alter any safety analyses. Therefore, it is concluded that this proposed change will not cause a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No As stated above, the parameters being restricted and monitored by these specifications ensure that the plant is operated in a manner consistent with the safety analyses.
The intervals between checks and the offset equation will be modified, but the revisions are also consistent with the
-3 safety analysis assumptions regarding core operating limits.
Therefore, it is concluded that this proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response: No As explained in the responses to questions 1 and 2, this proposed change will not alter the safety analyses assumptions. Therefore, it is concluded that this proposed change will not result in a significant reduction in a margin of safety.
The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards considerations.
Example (11) related to a change that constitutes an additional limitation, restriction or control not presently included in the technical specifications. Since the assumptions and results of applicable safety analyses remain unchanged, it is concluded that the proposed change does not constitute a significant hazards consideration.
The proposed change to the bases are most similar to example (1) of 48 FR 14870 since it is purely an administrative change to achieve consistency throughout the Technical Specifications. In this case, the changes are necessary to be consistent with the revisions to the body of the specification.
SAFETY AND SIGNIFICANT HAZARDS DETERMINATION Based on the safety evaluation, it is concluded that:
(1) the proposed change does not involve a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Environmental Statement. -
Existing Specification Section 3.10 and 3.11 -
Proposed Specification Section 3.10 and 3.11