ML14113A460
| ML14113A460 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/10/1980 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13316A521 | List: |
| References | |
| NUDOCS 8012120426 | |
| Download: ML14113A460 (4) | |
Text
DESCRIPTION OF PROPOSED CHANGE AND SAFETY ANALYSIS PROPOSED CHANGE NO. 100 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE DPR-13 This is a request to revise Appendix A Technical Specifications 2.1, "REACTOR CORE - Limiting Combination of Power, Pressure, and Temperature," 3.5.2, "Control Group Insertion Limits," and 3.11, "Continuous Power Distribution Monitoring."
Reason for Proposed Change By letter dated February 8, 1980, Proposed Change No. 88 to the Technical Specifications was submitted for NRC review and approval. Proposed Change No. 88 provided revisions to ensure that the Technical Specification objec tives were met with the loading of the reactor core for Cycle 8 operation.
The NRC issued Amendment No. 49 to Provisional Operating License DPR-13 on May 19, 1980, thereby approving the revisions of Proposed Change No. 88.
The steam generator repair program now underway at San Onofre Unit I has made it necessary to determine the impact which these modifications will have on the safety analysis for transients and accidents included in the Final Safety Analysis (FSA). Accordingly, a reanalysis was performed and documented in the enclosed report entitled, "Reload Safety Evaluation, San Onofre Nuclear Gener ating Station Unit 1, Cycle 8, Revision 1," dated October, 1980. Proposed Change No. 100 provides the Technical Specification changes which are neces sary to assure operation within safety limits as a result of the revised Reload Safety Evaluation.
Existing Specifications The existing specifications are as constituted in Sections 2.1, 3.5.2 and 3.11 of Appendix A to Provisional Operating License DPR-13.
Proposed Specifications Technical Specification 2.1, "Item (2) would be revised to read:
"The combination of reactor power and coolant temperature shall not exceed the locus of points established for the RCS pressure in Figure 2.1.1. If the actual power and temperature is above the locus of points for the appropriate RCS pressure, the safety limit is exceeded."
Figure 2.1.1 would be replaced with the figure of Enclosure 1.
Table 2.1, Item 4 would be revised to read:
"4. Variable Low Pressure 26.15 (0.984 LIT + Tavg) -
14341"
0
-2 The Basis for Technical Specification 3.5.2 (Item 1, 2nd para.) would be revised to read:
"A more restrictive limit on the design maximum value of specific power, F N and F is applied to operation in accordance with the current AH Q safety analysis including fuel densification and ECCS performance. The values of the specific power, F H and FQ are 13.7 kW/ft., 1.55 and 2.89 respectively. The control group insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution."
Technical Specification 3.11, Item A woulo be revised to read:
"A. The incore axial offset limits shall not exceed the functional relationship defined by:
For positive offsets:
1AO = 2.89/P -
2.1225 3.0 0.03021 2.89/P -
2.1181 For negative offsets:
IAO =
-0.03068
+ 3.0 where IAO = incore axial offset P
= fraction of rated thermal power" The balance of Technical Specifications 2.1, 3.5.2 and 3.11 would remain as constituted in Appendix A to Provisional Operating License No. DPR-13.
Safety Analysis The Reload Safety Evaluation was reanalyzed assuming a maximum of 20% steam generator tube plugging which results in a reduction of 6.6% in Reactor Coolant System Thermal Design Flow. The results of the evaluation are documenteo in the enclosed report entitled, "Reload Safety Evaluation, San Onofre Nuclear Generating Station Unit 1, Cycle 8, Revision 1," dateo October, 1980.
Based upon the analysis provided in the report discussed above, 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.
ENCLOSURE 1
Figure 2.1...
Safety Limits Temperature, Power, Pressure F
RCS Flow -
195,000 GPM
.1
-:7- 0
.1 I
I I
I.
30-7v 20..
7 7
-61 26 80
~ 590 58
-7 RCS Pressure 570 2250 PSI 360
- 2100 PSI
- 55. -
.8 1 40
-.2 ractioh of Nomina Power (1347 Mt) 1L6 16