ML13316B156
| ML13316B156 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/28/1988 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML13316B155 | List: |
| References | |
| DPR-13-A-112 NUDOCS 8811230401 | |
| Download: ML13316B156 (10) | |
Text
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+ o UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY DOCKET NO. 50-206 SAN ON0FRE NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 112 License No. DPR-13
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Southern California Edison Company and San Diego Gas and Electric Company (the licensee) dated August 31, 1987 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the.regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
P PDC
- 2. Accordingly, the license is amended by changes to the Technical Speci fications as indicated in the attachment to this license amendment, and paragraph 3.B. of Provisional Operating License No. DPR-13 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 112, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical
.Specifications, except where otherwise stated in specific license conditions.
- 3. This license amendment is effective as of the date of its issuance and must be fully implemented no later than 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION George Knighto Director Projet Directo te V Division of Reactor Projects -
- III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: October 28, 1988
ATTACHMENT TO LICENSE AMENDMENT NO. 112 PROVISIONAL OPERATING LICENSE NO. DPR-13 DOCKET NO. 50-206 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT ii ii 38e 39 39 39a 39a 39b 39b 39c 39c 39d 39d
Page 3.5.3 Control and Shutdown Rod Misalignment......
33b 3.5.4 Rod Position Indicating System.........
33g 3.5.5 Containment Isolation Instrumentation......
33j 3.5.6 Accident Monitoring Instrumentation.......
33n 3.5.7 Auxiliary Feedwater Instrumentation.
33p 3.5.8 Radioactive Liquid Effluent Instrumentation.
33s 3.5.9 Radioactive Gaseous Process and Effluent Monitoring Instrumentation..........
33v 3.5.10 Radiation Monitoring Instrumentation......
33y 3.6 CONTAINMENT.
34 3.7 AUXILIARY ELECTRICAL SUPPLY 36 3.8 FUEL LOADING AND REFUELING....
........ 38 3.9 MODERATOR TEMPERATURE COEFFICIENT.
38e 3.10 INCORE INSTRUMENTATION...........
39a 3.11 CONTINUOUS POWER DISTRIBUTION MONITORING........
39c 3.12 CONTROL ROOM EMERGENCY AIR TREATMENT SYSTEM 39e 3.13 SHOCK SUPPRESSORS (SNUBBERS) OPERABILITY........
39g 3.14 FIRE PROTECTION SYSTEMS OPERABILITY 39i 3.15 RADIOACTIVE LIQUID EFFLUENTS......
39n 3.16 RADIOACTIVE GASEOUS EFFLUENTS 39q 3.17 DOSE...
39w 3.18 RADIOLOGICAL ENVIRONMENTAL MONITORING..........
39x 3.19 SOLID RADIOACTIVE WASTE.....
39ff 3.20 OVERPRESSURE PROTECTION SYSTEMS...........
39gg SECTION 4 SURVEILLANCE REQUIREMENTS........
40 4.1 OPERATIONAL SAFETY ITEMS...
40b 4.2 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEM..
45 4.3 CONTAINMENT SYSTEMS....
47 4.4 EMERGENCY POWER SYSTEM PERIODIC TESTING..........
52 11 Amendment No. 00,
- id2, 112
- 38e 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/0F 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/ 0F MTC at a nominal core average coolant temperature of 575.OoF, 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 the-safety analysis remains valid, 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.
Amendment No. 112
39 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 Ak/k/oF. 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/0 F. In order to provide a margin of safety, the reactor shoyld not be operated with an MTC more negative than -3.8 x 10- 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.
Amendment No.
1, 2
-39a 3.10 INCORE INSTRUMENTATION APPI.ICABILITY:
MODE 1 above 90% RATED THERMAL POWER OBJECTIVE:
To specify the type and frequency of incore measurements used to verify linear power density values.
SPECIFICATION:
- a. A power distribution measurement shall be performed every 30 effective full power days (EFPD) and after attainment of equilibrium xenon upon return to power following a refueling shutdown.
- b. The incore instrumentation system shall be used to accomplish the Correlation Verification of incore versus excore data for the axial offset monitoring system prior to exceeding 90% of RATED THERMAL POWER following each refueling and at least once per 180 effective full power days (EFPD) thereafter.
Subsequent to the Correlation Verification and for the duration of each cycle, incore instrumentation shall be used to perform a Correlation Check of the axial offset monitoring system every 30 EFPD.
C. A core thermocouple map shall be taken every 30 EFPD and after attainment of equilibrium xenon upon return to power following a refueling shutdown.
ACTION:
A. If the correlation check, power distribution measurement or core thermocouple map described above cannot be made within the prescribed time, a maximum of 15 EFPO will be allowed for equipment correction.
B. In the event that Specification a, b and c cannot be met during the 15 EFPD allowed for corrective action, within one hour action shall be taken such that THERMAL POWER is restricted to less than or equal to 90% of RATED THERMAL POWER until these specifications can be met.
BASIS:
The flux mapping system is used to measure the core power distribution and to correlate incore versus excore data for the axial offset system. Measurements made with the flux mapping system every 30 effective full power days and upon return to power following a refueling shutdown will monitor the core power distribution to confirm that the maximum linear power density remains below allowable values. The Amendment No. 8, 112
39b axial offset system will monitor the axial core power distribution in a continuous manner. If the Correlation Verification or Correlation Check is not performed, the 90%
of full thermal power restriction assures safe operation of the reactor. In addition, core thermocouples provide an independent means of measuring the balance of power among the core quadrants.
The flux mapping system and the thermocouple system are not integral parts of the Reactor Protection System. These systems are, rather, surveillance systems which may be required in the event of an abnormal condition such as a power tilt or a control rod misalignment. Since such a condition cannot be predicted, it is prudent to have the surveillance systems in an operable state. The 90% of full power restriction, used when these measurements cannot be taken as scheduled, is applied to minimize the probability of exceeding allowed peaking factors.
Operation for a 180 effective full power day period prior to reperforming the correlation verification is acceptable on the basis that the allowed incore axial offset limits are reduced by the amount in percent of incore axial offset that the monthly correlation check differs from the correlation.
Amendment No. /.
112
39c 3.11 CONTINUOUS POWER DISTRIBUTION MONITORING APPLICABILITY:
MODE 1 above 90% RATED THERMAL POWER OBJECTIVE:
To provide corrective action in the event that the axial offset monitoring system limits are approached.
SPECIFICATION:
The incore axial offset limits shall not exceed the functional relationship defined by:
2.89/P -
2.1225 For positive offsets:
IAO - -----
FCC 0.03021 2.89/P - 2.1181 For negative offsets:
IAO =
+ FCC
-.03068 where IAO = incore axial offset P = fraction of RATED THERMAL POWER FCC = The larger of 3.0 or the value in percent of incore axial offset by which the current correlation check differs from the incore-excore correlation.
ACTION:
A. If the incore limit defined by the Specification, as measured by the excore axial offset system, is exceeded by both axial offset monitoring channels, within one hour action shall be taken such that THERMAL POWER is restricted to less than or equal to 90% of RATED THERMAL POWER until these specifications can be met.
B. If it is determined that one of the excore axial offset monitoring channels is inoperable, the remaining OPERABLE excore axial offset channel shall be used to provide power distribution information. In addition, one NIS channel current shall be logged every two hours and axial offset information determined from this data until the inoperable channel has been returned to OPERABILITY.
C. If both excore axial offset channels are declared ino erable, at least three NIS channel currents shall be ogged every two hours and offset information determined from these data. If no method for determining axial offset is available, within one hour action shall be taken such that THERMAL POWER is restricted to less than or equal to 90% of RATED THERMAL POWER until these specifications can be met.
Amendment No. J1 35 ES 112
39d BASIS:
The percent full power axial offset limits are conservatively established considering the core design peaking factor, analytical determination of the relationship between core peaking factors and incore axial offset considering a wide range of maneuvers and core conditions, and actual measurements relating incore axial offset to the axial offset monitoring systems. The axial offset limit established from the incore versus excore data have been reduced by an amount equivalent to FCC to allow for burnup and time dependent differences between the periodic correlation verification and the monthly correlation check. Correcting the allowed incore axial offset limits by an amount equal to FCC maintains plant operation within the original safety analysis assumptions.
Should a specific cycle analysis establish that the analytical determination of the relationship between core peaking factors and incore axial offset has changed in a manner warranting modification to the existing envelope of peaking factor (1,2), then a change to functional relationship of the specification shall be submitted to the Commission. The incore-excore data correlation is checked or verified periodically as delineated in Specification 3.10, INCORE INSTRUMENTATION.
Reducing power in cases when limits are approached or exceeded, will assure that design limits which were set in consideration of accident conditions are not exceeded.
Prior to installation of the axial offset monitoring system, the NIS system was used to monitor axial offset and showed there is considerable margin between axial offset normally seen and limits established in consideration of design peaking factors.
References:
(1) Supporting Information on Periodic Axial Offset Monitoring, San Onofre Nuclear Generating Station, Unit 1, September, 1973 (2) Supporting Information on the Continuous Axial Offset Monitoring System, San Onofre Nuclear Generating Station, Unit 1, July, 1974 (3) Description and Safety Analysis, Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1 Cycle 5, January, 1975, Westinghouse Non-Proprietary Class 3.
Amendment No. JJ.
112