ML13316B133
| ML13316B133 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/21/1988 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML13316B132 | List: |
| References | |
| DPR-13-A-111 NUDOCS 8810310103 | |
| Download: ML13316B133 (8) | |
Text
p effl REG0 UNITED STATES Sa 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.111 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 May 26, 1988 complies with the standards and require ments 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.
- I3O31 0103 S 1 02./
PDR A:Ic K 'I02A pADC 05000206.
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-2
- 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. 111, 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 W. Knighton, Director Project Directorate V Division of Reactor Projects -.III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: October 21, 1988
ATTACHMENT TO LICENSE AMENDMENT NO. 111 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 31 31 32 32 32a 32a 33 33 33a 33a
3.5.2 CONTROL ROD INSERTION LIMITS APPLICABILITY: MODES 1 and 2.
OBJECTIVE:
This specification defines the insertion limits for the control rods in order to ensure (1) an acceptable core power distribution during power operation, (2) a limit on potential reactivity insertions for a hypothetical control rod ejection, and (3) core subcriticality after a reactor trip.
SPECIFICATION: A. Except during low power physics tests or surveillance testing pursuant to Specification 4.1.1.G, the Shutdown Groups and Control Group 1 shall be fully withdrawn, and the position of Control Group 2 shall be at or above the 21-step uncertainty limit shown in Figure 3.5.2.1.
B. The energy weighted average of the positions of Control Group 2 shall be at least 90% (i.e. > Step 288) withdrawn after the first 20% burnup of a core cycle. The average shall be computed at least twice every month and shall consist of all Control Group 2 positions during the core cycle.
ACTION:
A. With the control groups inserted beyond the above insertion limits either:
- 1. Restore the control groups to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or
- 2. Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figure, or
- 3. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
B. With a single dropped rod from a shutdown group or control group, the provisions of Action A are not applicable, and retrieval shall be performed without increasing THERMAL POWER beyond the THERMAL POWER level prior to dropping the rod. An evaluation of the effect of the dropped rod shall be made to establish permissible THERMAL POWER levels for continued operation. If retrieval is not successful within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the time the rod was dropped, appropriate action, as determined from the evaluation, shall be taken. In no case shall operation longer than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> be permitted if the dropped rod is worth more than 0.4% A k/k.
BASIS:
During STARTUP and POWER OPERATION, the shutdown groups and control group 1 are fully withdrawn and control of the reactor is maintained by control group 2. The control group insertion limits are set in consideration of maximum specific power, shutdown capability, and the rod ejection accident. The considerations associated with each of these quantities are as follows:
31 Amendment No.
8,111
- 1. The initial design maximum value of specific power is 15 kW/ft. The values of FJH and FQ total associated with this specific power are 1.75 and 3.23, respectively.
A more restrictive limit on the design value of specific power, F2H and FO is applied to operation in accordance with the current safety analysis including fuel densification and ECCS performance. The values of the specific power, F H and FO are 13.7 kW/ft, 1.57 and 2.89, respectively. At partial power, the F2H maximum values (limits) increase according to the following equation, FAH (P) -
1.57 [1 + 0.2 (1-P)], where P is the fraction of RATED THERMAL POWER. The control group insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution.
- 2. The minimum shutdown capability required is 1.25% Ap at BOL, 1.9% Ap at EOL and defined linearly between these values for intermediate cycle lifetimes. The rod insertion limits ensure that the available SHUTDOWN MARGIN is greater than the above values.
- 3. The worst case ejected rod accident (8) covering HFP-BOL, HZP-BOL, HFP-EOL shall satisfy the following accident safety criteria:
a) Average fuel pellet enthalpy at the hot spot below 225 cal/gm for nonirradiated fuel and 220 cal/gm for irradiated fuel.
b) Fuel melting is limited to less than the innermost 10%
of the fuel pellet at the hot spot.
Low power physics tests are conducted approximately one to four times during the core cycle at or below 10% RATED THERMAL POWER. During such tests, rod configurations different from those specified in Figure 3.5.2.1 may be employed.
It is understood that other rod configurations may be used during physics tests. Such configurations are permissible based on the low probability of occurrence of steam line break or rod ejection during such rod configurations.
Operation of the reactor during cycle stretch out is conservative relative to the safety considerations of the control rod insertion limits, since the positioning of the rods durihg stretch out results in an increasing net available SHUTDOWN MARGIN.
32 Amendment No.
3
Compliance with Specification B prevents unfavorable axial power distributions due to operation for long intervals at deep control rod insertions.
The presence of a dropped rod leads to abnormal power distribution in the core. The location of the rod and its worth in reactivity determines its effect on the temperatures of nearby fuel. Under certain conditions, continued operation could result in fuel damage, and it is the intent of ACTION B to avoid such damage.
References:
(1) Final Engineering Report and Safety Analysis, revised July 28, 1970.
(2) Amendment No. 18 to Docket No. 50-206.
(3) Amendment No. 22 to Docket No. 50-206.
(4) Amendment No. 23 to Docket No. 50-206.
(5) Description and Safety Analysis, Proposed Change No. 7, dated October 22, 1971.
(6) Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1, Cycle 4, WCAP 8131, May 1973.
(7) Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1, Cycle 5, January 1975, Westinghouse Non Proprietary Class 3.
(8) An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods, NCAP-7588, Revision 1-A, January 1975.
32a Amendment No.,140, 50,111
THIS PAGE INTENTIONALLY LEFT BLANK.
33 Amendment No. 1
CONTROL GROUP INSERTION LIMITS FULLY WITHDRAWN 320 300-T
___261~
250-I 24011 OO FULLY 0 0 250 150 0 o 100-35 FULLY INSERTED 0
0 10 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.5.2.1 33a Amendment No. $6.11