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Southern Nuclear Design Calculation Plant: Vogtle Unit: 1 &2 Calculation Number: X6CNA 14 Sheet: 46 Miscellaneous Design Inputs 21. Iodine boiling point= 1 B4 C = -363 F
Reference:
Page B-1, "CRC Handbook of Chemistry
& Physics" 22. Density of Refueling Cavity and Spent Fuel Pool Water@ 130 F = 61.SS lbm/cu ft
Reference:
See Attachment C2. 23. Density of CVCS letdown flow= 0.99 glee (Attachment C2) V17 Page 1 of 5
Reference:
The density is used to convert the letdown activity from µCi/g to µCi/cc, which are the units used by the CVCS letdown rad monitor RE-4BOOO (Design Input #1 & Attachment CS). Based on at-power CVCS letdown parameters from the Unit 1 and 2 IPCs (Attachment CS), the average temperature and pressure at the radiation measurement location are 9B.S F and 3BS psig. 24. Average Decay Gamma Energies for RE-4BOOO principle isotopes (Attachment CB) Isotope Average
Reference:
Brookhaven National Laboratory National Gamma Nuclear Data Center decay data Energy (httQ://www
.orau.org/QtQ/PTP%20Librarv/librarv/DOE/bnl/nu (MeV) clidedata/table.htm
) 1-131 0.3B2 Copies of web pages in Attachment CB 1-132 2.20 1-133 0.607 1-134 2.SO 1-135 1.SS Co-5B 0.97S Co-60 2.S1 Cs-134 1.SS Cs-136 2.12 Cs-137 O.S6S Cs-13B 2.31 Southern Nuclear Design Calculation Plant: Vogtle Unit: 1 &2 Calculation Number: X6CNA 14 Recognition Category S: System Malfunctions Notice of Unusual Event SU4: Fuel Clad Degradation. Operating Mode Applicability: Power Operation (Mode 1) Startup (Mode 2) Hot Standby (Mode 3) Hot Shutdown (Mode 4) Emergency Action Levels: 1 OR 2 V17 Page 2 of 5 Sheet: 61 SU4 EAL 1: CVCS Letdown radiation monitor RE-48000 reading greater than 5 µCi/cc indicating fuel clad degradation greater than Technical specificatio n There are two Technical Specification limits on RCS coolant activity:
- SR 3.4.16.1: Gross specific activity
< 100/t= µCi/gm
- SR 3.4.16.2: Dose Equivalent 1-131 (DE 1-131) < 1.0 µCi/g Per section B.3.4.16, page 83.4.16-2 of VEGP Tech Spec Bases, noble gas activity in the reactor coolant assumes 1 % failed fuel, which closely equals the LCO limit of 1001r; µCi/gm for gross specific activity.
The EAL threshold will be calculated for each Tech Spec limit condition
. Per pages 12 and 13 of X6AZ01 A, the principle isotopes detected by RE-48000 are 1-131, 1-133, Co-58, Co-60, Cs-134, and Cs-137. However, per Section B-12-3-2 and Figure B-12-2 of 1X6AZ01-10004 & 2X6AZ01-10004, RE-48000 will detect gammas of energies down to -0.1 MeV. I .I I i *' i I I I ' ---,_ I _,....-....-
- I i/ I i ) I ' I I I I I I I I I l j ' I I ' ' ! I ' :I.I C.:l 0.3 O.* 0.:5 o.* 0.'T :J.a 0.!1 1.0 ... 1.2 (NCRGY t...EV£L.
Uh\I) Figure B-12-2 I I Thus the other I, Co, and Cs isotopes listed in FSAR Table 11.1-2 should be included if their average decay gamma energies exceed 0.1 MeV.
V17 Southern Nuclear Design Calculation Page 3 of 5 Plant: Vogtle Unit: 1 &2 Calculation Number: X6CNA 14 Sheet: 62 Per L TR-CRA-06-179 attached to WEC-SNC letter GP-18006, the MURPU coolant activities may be adjusted upward 2% to account for the increase in core thermal power from 3565 MWt to 3636 MWt. Thus, the Co and Cs MURPU 1 % defect activity are equal to their pre-MURPU 1 % Defect activities multiplied by 1.02. The Co and Cs activities corresponding to the 1.0 µCi/g DE 1-131 Tech Spec limit are the products of their MURPU 1 % defect activities and the ratio of the 1-131 DE 1-131 concentration to its equilibrium concentration (0. 7 4/2.91 ). The activities, expressed in µCi/g are summed and then multiplied by the CVCS letdown flow density (0.99 glee) to convert them to µCi/cc. The EAL threshold is the minimum of the 1 % Defect and the 1.0 µCi/g DE 1-131 activities
. 1.0 µCi/g MURPU Pre-MURPU DE 1-131 1% Defect 1% Defect Isotope Coolant Coolant Coolant Activity Activity Activity
(µCi/g) (µCi/g) (µCi/g) 1-131 0.74 2.91 1-132 0.75 2.96 1-133 1.41 5.56 1-134 0.18 0.69 1-135 0.69 2.72 Co-58 3.89E-03 1.53E-02 1.50E-02 Co-60 4.93E-04 1.94E-03 1.90E-03 Cs-134 5.97E-01 2.35 2.3 Cs-136 7.52E-01 2.96 2.9 Cs-137 3.89E-01 1.53 1.5 Total= 5.5 21.7 µCi/g Total= 5.5 21.5 µCi/cc CVCS Letdown Density = 0.99 glee Given the RG 1.97 R2 required system accuracy (Acceptance Criterion 3), the threshold is rounded down from 5.5 to 5 µCi/cc. NOTE: SU4 EAL2 not determined in this calculation.
V17 Southern Nuclear Design Calculation Page 4 of 5 Plant: Vogtle Unit: 1 &2 Calculation Number: X6CNA 14 Sheet: CS-1 Attachment C5 -VEGP 1 &2 CVCS Letdown Radiation Monitor (RE-48000)
Readings V17 Southern Nuclear Design Calculation Page 5 of 5 Plant: Vogtle Unit: 1 &2 Calculation Number: X6CNA 14 Sheet: C5-2 Attachment C5 -VEGP 1 &2 CVCS Letdown Radiation Monitor (RE-48000)
Readings I I I I I I I I I I 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity V18 Page 1 of 3 RCS Specific Activity 3.4.16 LCO 3.4.16 The specific activity of the reactor coolant shall be within limits. APPLICABILITY
- MODES 1 and 2, MODE 3 with RCS average temperature (Tav9) 500°F. ACTIONS ------------
N 0 TE--------------------------------------------------------
L CO 3.0.4c is applicable
. CONDITION A. DOSE EQUIVALENT A.1 1-131>1.0
µCi/gm. AND A.2 B. Gross specific activity of B.1 the reactor coolant not within limit. AND B.2 Vogtle Units 1 and 2 REQUIRED ACTION COMPLETION TIME Verify DOSE Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> EQUIVALENT 1-131 within the acceptable region of Figure 3.4.16-1. Restore DOSE EQUIVALENT 1-131 to within limit. Perform SR 3.4.16.2. Be in MODE 3 with Tav9 < 500°F. 3.4.16-1 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 4 hours 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (continued)
Amendment No. 137 (Unit 1) Amendment No. 116 (Unit 2)
ACTIONS (continued)
CONDITION C. Required Action and C.1 associated Completion Time of Condition A not met. OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1. SURVEILLANCE REQUIREMENTS REQUIRED ACTION Be in MODE 3 with Tavg < 500°F. SURVEILLANCE SR 3.4.16.1 SR 3.4.16.2 Verify reactor coolant gross specific 100/E µCi/gm. ----------------------------N 0 TE-----------------------------
0 n ly required to be performed in MODE 1. Verify reactor coolant DOSE EQUIVALENT 1-131 specific 1.0 µCi/gm. V18 Page 2 of 3 RCS Specific Activity 3.4.16 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period (continued)
Vogtle Units 1 and 2 3.4.16-2 Amendment No. 158 (Unit 1) Amendment No. 140 (Unit 2) l I V18 Page 3 of 3 RCS Specific Activity 3.4.16 250 ,g> (.) ..:'; !::: :::;; 200 :::; UNACCEPTABLE OPERATION ,__ (.) u:: C3 w 150 a. (/) ,__ g (.) Ii:.: < :::;; 100 Ci a. M ,__ z w ACCEP ABLE ...J OPER TION :::> 50 0 w w (/) 0 0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4.16-1 REACTOR COOLANT DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACITVITY
>1 mCi/gram DOSE EQUIVALENT 1-131 Vogtle Units 1 and 2 3.4.16-4 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2)
V19 Page 1 of 1 RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE; b. 1 gpm unidentified LEAKAGE; c. 10 gpm identified LEAKAGE; and d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG). APPLICABILITY
- MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A RCS operational A.1 Reduce LEAKAGE to LEAKAGE not within within limits. limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
B. Required Action and B.1 Be in MODE 3. associated Completion Time of Condition A not AND met. B.2 Be in MODE 5. OR Pressure boundary LEAKAGE exists. OR Primary to secondary LEAKAGE not within limit. Vogtle Units 1 and 2 3.4.13-1 COMPLETION TIME 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 6 hours 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Amendment No. 144 (Unit 1) Amendment No. 124 (Unit 2)
Table 3.3.2-1 (page 2 of 7) V20 Page 1 OF 1 ESFAS Instrumentation 3.3.2 Engineered Safety Feature Actuation System Instrumentation FUNCTION
- 2. Containment Spray a. Manual Initiation
- b. Automatic Actuation Logic and Actuation Relays c. Containment Pressure High -3 APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 1,2,3,4 1,2,3,4 1,2,3 REQUIRED SURVEILLANCE CHANNELS CONDITIONS REQUIREMENTS 2 B 2 c 4 E SR 3.3.2.6 SR 3.3.2.2 SR 3.3.2.3 SR 3.3.2.5 SR 3.3.2.1 SR 3.3.2.4Q>Gl SR 3.3.2.7Ql01 ALLOWABLE VALUE NA NA 22.4 psig NOMINAL TRIP SETPOINT NA NA 121.5 psig I (continued)
(i) If the as-found channel setpoint is outside its predefined as-found tolerance
, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. Gl The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance
- otherwise
, the channel shall be declared inoperable.
Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance
. The methodologies used to determine the as-found and the as-left tolerances are specified in NMP-ES-033-006, Vogtle Setpoint Uncertainty Methodology and Scaling Instructions
. Vogtle Units 1 and 2 3.3.2-10 Amendment No. 165 (Unit 1) Amendment No. 147 (Unit 2)