ML13331B045
| ML13331B045 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/26/1988 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13331B043 | List: |
| References | |
| NUDOCS 8806020063 | |
| Download: ML13331B045 (9) | |
Text
0 3.5.2 CONTROL GROUP INSERTION LIMITS APPLICABILITY:
This standard applies to the insertion limits for the 61 control banks during Startup and Power Operation.
6/11/81 OBJECTIVE:
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.
The positions of all control rods shall be at or above 61 the limits shown in Figure 3.5.2.1 except during low 6/11/81 power physics tests.
B.
The energy weighted average of the positions of 21 control bank 2 shall be at least 90% withdrawn after 5/13/75 the first 20% burnup of a core cycle The average shall be computed at least twice every month and shall consist of all control bank 2 positions during the core cycle.
C.
If it is determined that a.rod has been dropped, retrieval shall be performed without increasing power level.
An evaluation of the effect of the dropped rod shall be made to establish permissible 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 />, 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.
D.
(Deleted)
BASIS:
During Startup and Power Operation, the shutdown groups are fully withdrawn and control of the reactor is maintained by 61 the control groups.
The insertion limits are set in 6/11/81 consideration of maximum specific power, shutdown capability, and the rod ejection accident. The considerations associated with each of these quantities are as follows:
- 1.
The initial design maximum value of specific power is 15 kW/ft.
The values of F H and F total associated with this specific power are 1.75 and 3.23, respectively.
A more restrictive lmit on the design maximum value of specific power, F H and F is applied to operation in accordance with tfe current safety analysis 60 including fuel densification and ECRS performance.
6/8/81 The values of the specific power, F H and F are 13.7 kW/ft, 1.57 and 2.89, respectively. At partial power 88 the F H maximum values (limits) increase according 11/23/84 to the following equation, 8806020063-880526 34 ei PDR ADOCK 0500026 Revised:
3/4/85 DCD
N 88 F H (P) = 1.57 [1 + 0.2 (1-P)],
where P is the 8
fraction of rated power.
The control group insertion limits in conjunction with Specification B prevent 60 exceeding these values even assuming the most adverse 6/8/81 Xe distribution.
- 2.
The minimum shutdown capability required is 1.25% Ap p 21 at BOL, 1.9% Ap at EOL and defined linearly between 5/13/75 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 54 225 cal/gm for nonirradiated fuel and 220 cal/gm 5/29/80 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% 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 consideratons of the control rod insertion limits, since the positioning of the rods during stretch out results in an increasing net available shutdown.
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 61 its worth in reactivity determines its effect on the 6/11/81 temperatures of nearby fuel. Under certain conditions continued operation could result in fuel damage, and it is the intent of Specification C to avoid such damage.
3-50 Revised:
3/4/85
References:
(1) Final Engineerhg Report and Safety Analysis, revised July 28, 1970.
4 1/18/72 (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.
21 5/13/75 (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 54 Kinetics Kethods, WCAP-7588, Revision 1-A, January, 1975.
5/29/80 3-51 Revised:
8/1/80
v-,I G UR 1") 3. t3. 2 LIMITING CONDITION FOR OPERATION-CONTROL GROUP INSERTION LIMITS 8
100 CONTROL GROUP I 90 10 40 2o 1o 3.0 I-.I 0
10 20 30 40 50 so 1o so 90 100 POWER (PERCENT OF 1347 MWT)
PROPOSED TECHNICAL SPECIFICATION 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:
- 1. The initial design maximum value of specific power is 15 kW/ft. The values of FXH 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, FXH 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, FXH and F are 13.7 kW/ft, 1.57 and 2.89, respectively. At partial power, the F H maximum values (limits) increase according to the following equation, F2H (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 during stretch out results in an increasing net available SHUTDOWN MARGIN.
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, WCAP-7588, Revision 1-A, January 1975.
9076F
CONTROL GROUP INSERTION LIMITS FULLY WITHDRAWN 320 300 J
-261 250
-261
- 240 z
oD 200 C')
0 0
w u)150-0 0 100 -pV 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