ML11312A052

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Calculation IN1-245, Setpoint Calculation for PS-2390A & B, Condensate Tank Low Level Transfer, Revision 2
ML11312A052
Person / Time
Site: Pilgrim
Issue date: 10/01/2009
From: Richard D
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
References
IN1-245, Rev 2
Download: ML11312A052 (35)
v • d • e


Text

ATTACHMENT 9.2 ENGINEERING CALCULAT1ON COVER PAGE Sheet 1 of 13

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CALCULATION ()EC # 12609 (2) Page I of COVER PAGE (3) Design Basis Calc. [0 YES I-]NO 1(4) I*CALCULATION [ EC Markup

  • )Calculation No: IN11-245 ( Revision: 2 (7)

Title:

Setpoint Calculation for PS-2390A & B, Condensate Tank -")Editorial:

Low Level Transfer E]1 YES O*NO (9) System(s): 23 (10) Review Org (Department): I & C Design (11) Safety Class: (12) Component/Equipment/Structure Type/Number: ___________

Type/NPSber:

Z Safety / Quality Related PS 2390A PS 2390B r- Augmented Quality Program r- Non-Safety Related (13) Document Type: CALC (14) Keywords (Descriptionlropical Codes):

REVIEWS (15) Name/Sian ur/Dae (16) NameSignture.0"nat 0D. Richard Aai&Y---*4,z(-

Responsible Engineer Design Verifier lupervisor/Approval Reviewer Comments Attached El Comments Attached

CALCULATION REFERENCE SHEET ATTACHMENT ATTACHMENT 9.3 CALCULATION REFERENCE SHEET Sheet 2 of 13 CALCULATION CALCULATION NO: IN1-245 REFERENCE SHEET REVISION: 2 I. EC Markups Incorporated N/A to NP calculations)

II. Relationships: Sht Rev Input Output Impact Tracking I

Doc Doc Y/N No.

1. Calculation M-501 1 X 03 N
2. FSAR X 0 N
3. PNPS Tech Spec X X Y
10. SM418 2 3 X X Y 5.

III. CROSS

REFERENCES:

1. NRC Regulatory Guide 1.105, Rev. 3
2. ENN Engineering Guide ENN-IC-G-003, Rev. 0
3. ASME Steam Tables, 5th Edition
4. ENN-MS-S-009-PNP, Rev. 0
5. PNPS Equipment Qualification Master List, Rev. 45
6. SUDDS/RF92-039, Rev. 0
7. IAS PS-2390A PARAMETERS
8. IAS PS-2390B PARAMETERS
9. G. E. Data Sheet 225A5750
10. Dwg. M209, Rev. 66
11. Dwg. M243, Rev. 51
12. Dwg. M244 sh. 1, Rev. E30
13. Dwg. M184, Rev. E14
14. Dwg. E153, Rev. 2
15. Dwg. M1J14-14 sh.1, Rev. 26
16. Dwg. M1J14-14 sh. 2, Rev. E4
17. Dwg. M1J16-10, Rev. 25
18. Dwg. M1J17-12, Rev. 24
19. Dwg. M1J19-19, Rev. 16
20. Dwg. M1J20-5, Rev. 15
21. Dwg. C-338, Rev. 3
22. Dwg. SM418 sh.3, Rev. E2
23. ABB Impell Project Instruction No. 25-226-PI-001, Rev. 2
24. PNPS Procedure 8.M.2-2.5.6 IV. SOFTWARE USED:

Title:

N/A Version/Release: Disk/CD No._

V. DISK/CDS INCLUDED:

Title:

N/A Version/Release Disk/CD No.

VI. OTHER CHANGES: I

ATAHMENT1 9.4 RECORD OF REVISION SheeT 9 eorfR-N-13-082 RECORD OF REVIsIO Etrg3oretieAcin Sheetslto Revision 13epord of Reiifslon This change is incorporating Engineering Change EC1I2609. The changes are a result of Entergy Corrective Action Report CR-PNP-2006-O1 802-2 CA15.

4.

-U

4. TABLE OF CONTENTS page 1.0 Calculation Cover Page 1 2.0 Calculation Reference Sheet 2 3.0 Record of Revision 3 4.0 Table of Contents 4 5.0 Purpose 4 6.0 Conclusion 4 7.0 Input and Design Criteria 5 8.0 Assumptions 11 9.0 Method of Analysis 11 10.0 Calculation 11 11.0 Attachments 13
5. PURPOSE The calculation provides the uncertainty analysis for the CST level switches PS-2390A and PS2390B. Itdetermines the level setpoint of the switches which ensures automatic transfer of the HPCI suction from the CST to Suppression Pool before CST inventory is depleted.

This revision to calculation IN1 -245 is incorporating a change to the CST low water level Suppression Pool transfer setpoint analytical limit. Reference calculation M-501. The change is being made to resolve concerns relating to vortexing in the CST resulting in air ingestion in the suction of the HPCI pump. These concerns were identified in Entergy Corrective Action Report CR-PNP-2006-01802-CA15.

The calculation will support Nuclear Change EC1 2609 and will require FSAR and Tech Spec revision.

6. CONCLUSION The calculation determined the following:

Trip Setpoint:

58 inches from tank zero, or 8.6 psig (this includes hydrostatic head pressure correction for the elevation difference between the measured level of the tanks and the location of the pressure switches; i.e. +6.5psig) see attachment 2.

Technical Specification Allowable Value:

>-46 inches from tank zero, or >8.17 psig (this includes hydrostatic head pressure correction for the elevation difference between the measured level of the tanks and the location of the pressure switches; i.e. +6.5psig) see attachment 2.

Note, the setpoint Allowable Value listed in Technical Specification Table 3.2.B requires revision. The new value is > 46 inches above tank zero.

Setpoint Analytical Limit:

43 inches from tank zero, or 8.05 (psig) (this includes hydrostatic head pressure correction for the elevation difference between the measured level of the tanks and the location of the pressure switches; i.e. +6.5psig) see attachment 2.

Setpoint Reset Value:

< 141 inches from tank zero, or "11.6 (psig) (this includes hydrostatic head pressure correction for the elevation difference between the measured level of the tanks and the location of the pressure switches; i.e. +6.5psig) see attachment 2 and note 4.

No Adiust Limit:

8.4 to 8.8 psig Surveillance Interval:

Once per 3 months M&TE Accuracy:

+/-0.03 psig (Specified in section 5 of PNPS Procedure 8.M.2-2.5.6)

7. INPUT AND DESIGN CRITERIA Section 7.4.3.2.5 of the FSAR describes the control scheme of the CST low level Suppression Pool transfer. It states three pump suction valves are provided in the High Pressure Core Injection (HPCI) System. One valve lines up to the pump suction from the CST and the other two from Suppression Pool. The CST is the preferred source. Upon HPCI initiation if the suction valve of the CST is closed, the initiation signal opens it. If the water level in the CST falls below the Suppression Pool transfer level setpoint, the Suppression Pool suction valves open automatically after a 3 second time delay. The time delay for the suppression pool suction valve opening is introduced to prevent false (transient) signals from initiating suction transfer. When the suppression suction valves are fully open, the CST suction valve automatically closes. Pressure switches PS-2390A and B are used to detect the CST low water level condition. Either switch can initiate opening of the Suppression Pool suction valves.

The CST suction valve does not actually receive a signal to close until after the 3 second time delay plus the cycle time of the suppression pool suction valves to fully open. The delay is insignificant with respect to the change in water level in the CST and is accounted for in the determination of the analytical limit.

i.e. CST Tank Level Change = (HPCI + RCIC Flow)(Time Delay)/(CST Surface Area)

Figure 7.1 - Condensate Storage Tank HPCI Pump Suction Low Level

References:

Dwg. M-209, Dwg. M-243 Dwg. M-244, Dwg. E-153 Dwg. M1J14-14 shl, Dwg. M1J16-10 Dwg. M1J17-12, Dwg. M1P430-12 Calculation Input Data for the uncertainty/setpoint determination is provided in Table 7.1.

TABLE 7.1 - Loop Module Data Sheet Title Descriptive value References Remarks Component ID PS-2390A/B M209 Service Description Provide signal to open MO-2301-35 & MO-2301- FSAR 7.4.3.2.5 36 on low CST level Location Reactor Bldg. Aux. Bay, SUDDS/RF #92-East Wall 039 Attach. K El. 8'- 12" Dwg. M184 Manufacturer IAS PS-2390A&B Static-O-Ring Parameters Model Number IAS PS2390A&B 6N-AA2-X5PP Parameters Quality Category Q ENN-MS-S-009-PNP Adjustable Range 2 - 25 psig GE Instrument See Note 11 (55.4 - 692 inches H20) Data Sheet 225A5750 Process Calibrated Range N/A Setpoint calibration only Input Signal Calibrated N/A Setpoint Range calibration only Output Signal Calibrated N/A Setpoint Range calibration only Reference Accuracy (RA) N/A See Note 8 Drift (DR) +/- 0.38 psig Attachment 1 See Note 8 Static Pressure Effect (SP) N/A This error is applicable to DP instruments only External Pressurization N/A See Note 13 Effect (EP)

Overpressure Effect (OP) N/A See Note14 Temperature Effect See Note 3 Normal (TE) +/- 0.175 psig Accident (ATE)

Humidity Effect (HE) N/A See Note 15 Radiation Effect (RE) N/A PNPS See Note 5 Environmental Qualification Master List Seismic Effect (SE) +/- 0.25 psig Attachment See Note 6 Insulation Resistance N/A See Note 7 Effect(IR)

Power Supply Effect (PS) N/A See Note 7 Indicator Reading N/A No Indicator

Uncertainty (R)

Process Measurement N/A See Note 16 Effect (PM)

Primary Element Accuracy N/A No Primary (PE) Element Measurement and Test N/A Procedure See Note 8 Equipment Uncertainty 8.M.2-2.5.6 (MTE)

Technical Specification if _18 inches above tank zero Tech Spec Table See Note 2 Applicable 3.2.B Analytical Limit (AL) or 43 inches above tank zero Calculation M- See Note 10 Nominal Process Limit 501 (NPL)

Allowable Value (AV) -46 inches above tank zero SetPoint:

Trip (NTSP) 58 inches above tank zero Reset (RSV) <141 inches above tank See Note 4 zero Calibration Frequency Once per 3 months Tech Spec Table

+/- 25% 4.2.B and section 1; Definitions As-Found Tolerance (AFT) +/- 0.43 psig Section 10 This calculation As-Left Tolerance (ALT) +/- 0.2 psig Procedure 8.M.2-2.5.6 Calibration Procedure No. 8.M.2-2.5.6 Module Algorithm N/A EQ and/or Functional N/A Equipment is Operating Environment non-EQ Safety Function / Other See Note 1 Functional requirements Function Duration N/A Note 5 The function duration is not important because the switches are non-EQ Normal Operation Upper N/A Decreasing Limit (NUL) Setpoint Normal Operation Lower 126 inches from tank zero See note 9 Limit (NLL)

Operating Margin (OM)

OpertinMarinOM)_F________ 68 inches Section 10 IThis calculation_

I

Notes

1. The safety function of the switches is to provide signal to initiate transfer of HPCI pump suction from the. CST to suction from the suppression pool upon CST low water level condition. These switches are listed as 'Q' in the 0-List.
2. Tech Specs Table 3.2.B specifies Condensate Storage Tank Low Level Trip Function must be

>18" above tank zero, minimum of 2 operable Instrument Channels per Trip System. Tech Spec Table 4.2.B specifies Calibration Frequency once/3 months.

3. Per Attachment 3; the Temperature Effect (TE) is 2% of Full Range (FR.) per 100 0 F. Per FSAR Table 10.9-1 and 10.9-2, maximum and minimum temperatures for the area the switches are located are 105 0 F and 60°F respectively. The calibration temperature is assumed to be 700 F. For conservatism, the larger AT is used to determine TE:

AT=105-70AT=35 0 F AT 35 TE--+(2%of FR* -TTE--' (0.02

  • 25, 100 10010 TE---- 0. 175 psig
4. Empirical data from obtained from the PNPS calibration records (Reference PNPS Procedure 8.M.2-2.5.6) show that PS2390A and PS2390B reset differential is within the G.E. Instrument Data Sheet Specifications for Reset Span Differential, 0.5 to 3 psid (see Reference 9). Adding the upper limit of the pressure switch reset differential span to the plant setpoint confirms the switch will not reset until after the CST level returns to a level greater than the NLL (Normal Operation Lower Limit). This will prevent opening the CST suction valve to HPCI/RCIC until this level is achieved.
5. Per the PNPS Equipment Qualification List, the pressure switches are not EQ qualified. The switches are not required to function for a harsh environment. Therefore, this calculation will be performed for normal operating conditions only.
6. The switches are seismically qualified. Per the seismic qualification test results, (see Attachment 3) the SOR switches function normally before and after seismic simulation. The vendor specifies an accuracy of +/- 1% Full Range. This value will be used in this calculation for the seismic effect (SE).
7. The Insulation Resistance Effect and Power Supply Effect are applicable to low energy analog signals and do not apply to ON/OFF mechanical devices. The switches PS-2390A&B are ON/OFF mechanical devices that open and close 125V DC contacts. These errors are not applicable to these devices.

Notes (cont.)

8. The sensor drift was statistically analyzed in Attachment 1 in accordance with ABB Impell Project Instruction 25-226-PI-001. The analysis used the data obtained from empirical data from 11/6/87 through 8/22/2007 obtained from the PNPS as-found/as-left instrument calibration records (reference PNPS Procedure 8.M.2-2.5.6). The calibration frequency of the switches for the above data is once per 3 months
  • 25%. The value obtained from the statistical analysis has a probability and confidence of 95/95%.

From attachment 1, the drift value was determined to be +/-0.38 psi The sensor drift determined is considered to include the effects of measurement and test equipment (M&TE) and Reference Accuracy. Therefore, values will not be included in this calculation for M&TE Uncertainty or RA (Reference Accuracy).

9. The Condensate Storage Tanks provide the preferred supply to the HPCI and RCIC systems.

The Torus water storage provides the back up emergency HPCI and RCIC system supply. All suctions for the CST are located 10.5 feet (126 inches from the bottom of the tank; approximately 75,000 gallons) above the HPCI and RCIC suctions. Two stand pipes are in each tank, both are 10.5 feet high, one is for return water and the other is for the transfer pump supply. This ensures that there is a reserve available for HPCI and RC1C. The bottom penetration is normally lined up to the HPCI and RCIC systems.

10.Analytical Limit calculated in M-501, Rev 1, Minimum CST Level for Transfer of HPCI Pump Suction to Torus.

Analytical Limit, AL = 43 in; converted to psig (The density of water at 40 deg F is 62.426 lbs/ft) 43in*62.4261bs Yft3 ALpig - 1728in 3 /ft 3 1.553psig (rounded to 1.55 psig)

11. Pressure Switch Range is obtained from G.E. Instrument Data Sheet (Cross Reference 7),

and as read from data plate on switches. Exact replacements can no longer be purchased from SOR. Current model of similar equipment has a higher adjustable range, 3-30 psig instead of 2-25 psig, see Attachment 5. Replacement of this switch with new equipment, if needed at some future date will require revision to this calculation.

12. CST overflow line at 39.5 ft. (474 inches from the bottom of the tank).
13. The External Pressurization Effect is not applicable because both sides of the pressure switch sensing mechanism are referenced to atmosphere. (i.e. Both the tank and switch are vented to atmosphere.) Therefore any change in the external or ambient pressure at the switch will be cancelled by the change in pressure at the Condensate Storage Tank.
14. The Overpressure Effect is not applicable because the Condensate Storage Tank is 40 feet high and vented. The maximum pressure of the process is within the calibration range of PS-2390A & B. Therefore, it is not possible to overpressure PS-2390A & B.

Notes (cont.)

I

15. The housing of the pressure switch is NEMA 4 weather tight. Pressure switches are not normally affected by humidity and the manufacture does not specify an error due to humidity variation. Therefore, the humidity effect is considered to be insignificant, HE = 0 psig.
16. The Process Measurement Effect is due to process temperature changes and piping friction losses. Neither of these effects results in a non-conservative error. The first effect is not applicable because the setpoint determination is performed assuming the process temperature (including instrument line fluid) is at 40 degrees F. The temperature of the CST water is controlled at greater than 45 OF. See drawing SM418 sh. 2 for temperature control of the CST. Conversion of the CST level in inches of water to pressure in lbs/in2 is calculated at 400 F. The density of water at 40 degrees F is 62.426 lbs/ft3 . The conversion of the setpoint elevation to sensed pressure at PS2390A & B location is performed assuming the lowest temperature of the Condensate Storage Tank range of operability. This is a conservative method of analysis which results in a pressure switch measurement that corresponds to a CST level that is lower than actual. Also, fluid flowing through a piping system experiences a drop in pressure due to piping friction. The loss of process pressure due to piping friction results in a lower pressure at the location of the pressure switch which is located in the HPCI/RCIC suction piping. Reference drawing M209. This error is not applicable because it causes the pressure switch to trip at a higher CST level than required which is conservative.
8. ASSUMPTIONS CST temperature is assumed to be at the lowest value of the operating range. Refer to Drawing SM418 sh. 3. This is a conservative assumption which will result in an acceptable setpoint for all operating temperatures of the CST.
9. METHOD OF ANALYSIS This calculation is performed based on the methodology described in ENN-IC-G-003 "Instrument Loop Accuracy and Setpoint Calculation Methodology" and uses ABB Impell Project Instruction No. 25-226-PI-001 for analysis of the as-found and as-left instrument calibration data. The calculation has been prepared in accordance with EN-DC-126 "Engineering Calculation Process".
10. CALCULATION The setpoint for the CST low level Suppression Pool transfer is determined in accordance with rigor and equations prescribed in ENN-IC-G-003. The rigor as identified in ENN-IC-G-003 is type 1. The applicable equations used in the analysis follow.

Instrument Module Uncertainty, en The general form of the module uncertainty equations are:

en+ = +(RA 2 +DR 2 +TE 2+HE2+RE 2+pS 2+Sp 2+op2+S E2+Ep 2+ALT 2+MTE 2 +R2)112 +B e, = -(RA 2+DR 2+TE2+HE 2+RE 2+pS2 +Sp 2+op 2+S E2+EP 2+ALT 2+MTE 2+R2)/2 -B Channel Uncertainty, CU The general form of the channel uncertainty equation are:

CU+ =+(pM 2+PE 2 +e 12+.. .+e2 )112 +8 CU- =-(pM2+PE 2+e1 2+. ..+e2)1/2 -B Trip Setpoint. NTSP The general form of the trip setpoint equation is:

NTSP = AL+/-(CU+Margin)

As-Found Tolerance, AFT The general form of the As-Left Tolerance equation is:

2 21 2 AFTn = (RAn2 +DRn +ALTn )l Allowable Value, AV The general form of the Allowable Value equation is:

AV = NTSP+(AVTSM+Margin)

AVSTM = AFT Operatina Margin, OM The general form of the Operating Margin equation is:

OM = NLL - NTSP From the data in Table 7.1 the setpoint, Allowable Value and operating margin are determined using the above equations. Note, because the setpoint is decreasing, there are no biases and there is only one component in the loop being analyzed the following equations are applicable.

Switch Error eswitch+ =+(DR 2+TE 2 +SE 2+ALT 2 )112 +B

=+(0.382+0. 1752+0.252+0.22)1/2

=+0.53 psig Loop Error CU+ =+(0.e sigh2)1/2

=+0.53 psig

Trip Setpoint NTSP = AL+/-(CU+Margin)

= 1.55+(0.53+ 0)

= 2.08 psig, or 2.1 psig (The actual plant setting is obtained by adding the head correction from Attachment 2. The plant setpoint is 8.6 psig)

The Trip Setpoint is converted to inches from the bottom of the tank as follows:

(2.1 psig) (27.73 inWC / psi @ 68 degree F) = 58.2 inches, say 58 inches The As-Found Tolerance and Allowable Value are:

As-Found Tolerance 2 2 1 2 AFTswftch = (DRswch +ALTswkch )

= (0.382+0.22)1/2

= +/-0.43 psig, or (0.43psig) (27.73 inWC / psi @ 68 degree F) = +/-11.9 inches Allowable Value AV = NTSP-(AVTSM+Margin)

= 2.1 -(0.43 + 0)

= 1.67 psig (Adding the head correction from Attachment 2, the AV is 8.17 psig)

The Allowable Value is converted to inches from the bottom of the tank as follows:

(1.67 psig) (27.73 inWC / psi @ 68 degree F) = 46.3 inches, say 46 inches Operatina Margin The Operating Margin in inches is:

OM = NLL- NTSP

= (126 - 58) inches

= 68 inches

11. ATTACHMENTS
1. PS-2390NB Drift Data Analysis (6 pages)
2. PS-2390A & B Instrument Leg Head Correction (1 page)
3. Memo dated 10/19/1987 from Virginia Woldow to John Torbeck (1 page)
4. Boston Edison Notes of Telecon - dated 1/11/1999 (1 page)

ATTACHMENT 1 PS-2390A/B DRIFT DATA ANALYSIS CALC. NO. IN1-245 DATE DATA STATUS CAL DATA INTERVAL REMARKS DRIFT DATA 1ST OUTLIER 2ND OUTLIER PS-2390B 6/2/1994 AS FOUND 7.70 AS LEFT 7.70 9/6/1994 AS FOUND 7.70 96 0.00 0.00 0.00 AS LEFT 7.70 12/12/1994 AS FOUND 7.70 97 0.00 0.00 0.00 AS LEFT 7.70 3/14/1995 AS FOUND 7.70 92 0.00 0.00 0.00 AS LEFT 7.70 6/28/1995 AS FOUND 7.82 106 0.12 0.12 0.12 AS LEFT 7.82 9/29/1995 AS FOUND 7.65 93 -0.17 -0.17 -0.17 AS LEFT 7.65 1/3/1996 AS FOUND 7.89 96 0.24 0.24 0.24 AS LEFT 7.89, 4/3/1996 AS FOUND 7.90 91 0.01 0.01 0.01 AS LEFT 7.90 7/1/1996 AS FOUND 7.84 89 -0.06 -0.06 -0.06 AS LEFT 7.84 10/22/1996 AS FOUND 7.75 113 -0.09 -0.09 -0.09 AS LEFT 7.751 12/27/1996 AS FOUND 7.93 66 0.18 0.18 0.18 AS LEFT 7.93 4/7/1997 AS FOUND 7.85 101 -0.08 -0.08 -0.08 AS LEFT 7.85 7/14/1997 AS FOUND 7.75 98 -0.10 -0.10 -0.10 AS LEFT 7.75 10/16/1997 AS FOUND 7.83 94 0.08 0.08 0.08 AS LEFT 7.83 1/20/1998 AS FOUND 7.80 96 -0.03 -0.03 -0.03 AS LEFT 7.80 4/29/1998 AS FOUND 7.94 99 0.14 0.14 0.14 AS LEFT 7.94 _

7/29/1998 AS FOUND 7.89 91 -0.05 -0.05 -0.05 AS LEFT 7.89 _

10/19/1998 AS FOUND 8.00 82 0.11 0.11 0.11 AS LEFT 8.00 1/1311999 AS FOUND 8.20 86 0.20 0.20 0.20 AS LEFT 7.70 3/2/1999 AS FOUND 7.70 48 0.00 0.00 0.00 IN1_245.XLS Page I

ATTACHMENT 1 PS-2390A/B DRIFT DATA ANALYSIS CALC. NO. IN -245 AS LEFT 7.70 4/9/1999 AS FOUND 7.60 38 -0.10 -0.10 -0.10 AS LEFT 7.60 6/27/1999 AS FOUND 7.71 79 0.11 0.11 0.11 AS LEFT 7.71 9/1/1999 AS FOUND 7.70 66 -0.01 -0.01 -0.01

_AS LEFT 7.70 11/18/1999 AS FOUND 7.60 78 0.10 0.10 0.10 AS LEFT 7.80 2/24/2000 AS FOUND 7.71 98 -0.09 -0.09 -0.09 AS LEFT 7.71 6/2/2000 AS FOUND 7.71 99 0.00 0.00 0.00 AS LEFT 7.71 8/30/2000 AS FOUND 7.79 89 0.08 0.08 0.08 AS LEFT 7.79 12/11/2000 AS FOUND 7.50 93 -0.29 -0.29 -0.29 AS LEFT 7.82 2/28/2001 AS FOUND 7.76 89 -0.06 -0.06 -0.06 AS LEFT 7.86.

5/29/2001 AS FOUND 7.70 90 -0.16 -0.16 -0.1-6 AS LEFT 7.70 8/28/2001 AS FOUND 7.65 91 -0.05 -0.05 -0.05 AS LEFT 7.65 11/30/2001 AS FOUND 7.70 94 0.05 0.05 0.05 AS LEFT 7.701 2/25/2002 AS FOUND 8.00 87 0.30 0.30 0.30 AS LEFT 8.00 5/28/2002 AS FOUND 7.88 92 -0.12 -0.12 -0.12 AS LEFT 7.88 8/26/2002 AS FOUND 7.97 90 0.09 0.09 0.09 AS LEFT 7.97 11/26/2002 AS FOUND 7.71 92 -0.26 -0.26 -0.26 AS LEFT 7.711 2/24/2003 AS FOUND 7.70 90 -0.01 -0.01 -0.01 AS LEFT 7.701 5/27/2003 AS FOUND 7.60 92 -0.10 -0.10 -0.10 AS LEFT 7.60 _

8/26/2003 AS FOUND 7.60 91 0.00 0.00 0.00 AS LEFT 7.60 11/24/2003 AS FOUND 7.76 90 0.16 0.16 0.16 AS LEFT 7.76 3/1/2004 AS FOUND 7.40 98 -0.36 .-

0.36 -0.36 IN1_245.XLS Page 2

ATTACHMENT 1 PS-2390A/B DRIFT DATA ANALYSIS CALC. NO. IN1-245 AS LEFT 7.70 65/2712004 AS FOUND 7.90 87 0.20 0.20 0.20 AS LEFT 7.90

-8/23/2004 AS FOUND 8.30 88 0.40 0.40 0.40 AS LEFT 7.76 11/23/2004 AS FOUND 7.60 92 -0.16 -0.16 -0.16 AS LEFT 7.60 12/29/2004 AS FOUND 8.00 36 0.40 0.40 0.4O AS LEFT 8.00

-223/2005 AS FOUND 7.96 56 -0.04 -0.04 -0.04 AS LEFT 7.96 5/24/2005 AS FOUND 7.90 90 -0.06 -0.06 -0.06 AS LEFT 7.90 8/23/2005 AS FOUND 7.85 91 -0.05 -0.05 -0.05 AS LEFT 7.85 11/21/2005 AS FOUND 8.20 90 0.35 0.35 0.35 AS LEFT 7.84 P2/2/2006 AS FOUND 7.78 93 -0.06 -0.06 -0.06 AS LEFT 7.78 5/23/2006 AS FOUND 7.60 90 -0.18 -0.18 -0.18 AS LEFT 7.60 - _

8/22/2006 AS FOUND 7.70 91 0.10 0.10 0.10 AS LEFT 7.70 1i/22/2006 AS FOUND 7.30 92 -0.40 -0.40 -0.40 AS LEFT 7.80 2(23/2007 AS FOUND 7.90 93 0.10 0.10 0.10 AS LEFT 7.90 5/23/2007 AS FOUND 7.98 89 0.08 0.08 0.08 AS LEFT 7.98 8/22/2007 AS FOUND 7.90 91 -0.08 -0.08 -0.08 7.90 PS-2390A 6/2/1994 AS FOUND 7.70 AS LEFT 7.70 9/6/1994 AS FOUND 7.60 96 -0.10 -0.10 -0.10 AS LEFT 7.60 12/24/1994 AS FOUND 7.80 109 0.20 0.20 0.20 AS LEFT 7.80,

-314/1995 AS FOUND 7.80 80 0.00 0.00 0.00 AS LEFT 7.80 6/25/1995 AS FOUND 7.40 103 -0.40, -0.40 -0.40 IN1_245.XLS Page 3

ATTACHMENT 1 PS-2390A/B DRIFT DATA ANALYSIS CALC. NO. IN1-245 AS LEFT 7.75 9/29/1995 AS FOUND 7.62 96 -0.13 -0.13 -0.13 AS LEFT 7.62 1/3/1996 AS FOUND 7.97 96 0.35 0.35 0.35 AS LEFT 7.97 4/3/1996 AS FOUND 8.00 91 0.03 0.03i 0.03 AS LEFT 8.001 7/1/1996 AS FOUND 7.94 89 -0.06 -0.06 -0.06 AS LEFT 7.94 10/22/1996 AS FOUND 7.85 113 -0.09 -0.09 -0.09 AS LEFT 7.85 12/27/1996 AS FOUND 7.97 66 0.12 0.12 0.12 AS LEFT 7.97 4/7/1997 AS FOUND 7.90 101 -0.07 -0.07 -0.07 AS LEFT 7.901 7/14/1997 AS FOUND 7.85 98 -0.05 -0.05 -0.05 AS LEFT 7.85 10/16/1997 AS FOUND 7.85 94 0.00 0.00 0.00 AS LEFT 7.85 1/20/1998 AS FOUND 7.90 96 0.05 0.05 0.05 AS LEFT 7.90 4/29/1998 AS FOUND 7.88 99 -0.02 -0.02 -0.02 AS LEFT 7.88 7/29/1998 AS FOUND 7.85 91 -0.03 -0.03 -0.03 AS LEFT 7.85 10/19/1998 AS FOUND 7.90 82 0.05 0.05 0.05 AS LEFT 7.901 1/13/1999 AS FOUND 8.00 86 0.10 0.10 0.10 AS LEFT 8.00 3/2/1999 AS FOUND 8.00 48 0.00 0.00 0.00 AS LEFT 8.00 4/9/1999 AS FOUND 8.00 38 0.00 0.00 0.00

_AS LEFT 8.001 6/27/1999 AS FOUND 7.75 79 -0.25 -0.25 -0.25 AS LEFT 7.751 9/1/1999 AS FOUND 7.92 66 0.17 0.17 0.17

_AS LEFT 7.92 11/18/1999 AS FOUND 8.00 78 0.08 0.08 0.08 AS LEFT 8.00 2/24/2000 AS FOUND 8.01 98 0.01 0.01 0.01 7.65 IAS LEFT 6/2/2000 AS FOUND 7.66 99 .0.01 0.01 0.01 INi_245.XLS Page 4

ATTACHMENT 1 PS-2390A/B DRIFT DATA ANALYSIS CALC. NO. IN1-245 AS LEFT 7.66 8/30/2000 AS FOUND 7.81 89 0.15 0.15 0.15 AS LEFT 7.81 12/1/2000 AS FOUND 7.40 93 -0.41 -0.41 -0.41 AS LEFT 7.87 2/28/2001 AS FOUND 7.78 89 -0.09 -0.09 -0.09 AS LEFT 7.78 5/29/2001 AS FOUND 7.61 90 -0.17 -0.17 -0.17 AS LEFT 7.61 8/28/2001 AS FOUND 7.79 91 0.18 0.18 0.18 AS LEFT 7.79 11/30/2001 AS FOUND 7.70 94 -0.09 -0.09 -0.09 AS LEFT 7.70 2/25/2002 AS FOUND 7.90 87 0.20 0.20 0.20 AS LEFT 7.90 5/28/2002 AS FOUND 7.90 92 0.00 0.00 0.00 AS LEFT 7.90 8/26/2002 AS FOUND 7.85 90 -0.05 -0.05 -0.05 AS LEFT 7.85 11/26/2002 AS FOUND 7.77 92 -0.08 -0.08 -0.08 AS LEFT 7.77 2/24/2003 AS FOUND 8.00 90 0.23 0.23 0.23

_AS LEFT 8.00 5/27/2003 AS FOUND 7.60 92 -0.40 -0.40 -0.40 AS LEFT 7.60 8/26/2003 AS FOUND 7.80 91 0.20 0.20 0.20 AS LEFT 7.80 11/24/2003 AS FOUND 7.92 90 0.12 0.12 0.12 AS LEFT 7.921 3/1/2004 AS FOUND 7.90 98 -0.02 -0.02 -0.02 AS LEFT 7.90 5/27/2004 AS FOUND 7.64 87 -0.26 -0,26 -0.26 AS LEFT 7.64 8/23/2004 AS FOUND 7.60 88 -0.04 -0.04 -0.04 AS LEFT 7.60 11/23/2004 AS FOUND 7.80 92 0.20 0.20 0.20 AS LEFT 7.80 12/29/2004 AS FOUND 7.75 36 -0.05 -0.05 -0.05 AS LEFT 7.75 2/23/2005 AS FOUND 7.65 56 -0.10 -0.10 -04.10 AS LEFT 7.65 5/24/2005 AS FOUND 7.83 g90 0.18 0.18 1 0.18 IN1-245.XLS Page 5

ATTACHMENT 1 PS-2390A/B DRIFT DATA ANALYSIS CALC. NO. IN1-245 A S L E FT 7 .83 -0.53 8/23/2005 AS FOUND 7.30 91 -0.53 -0.53 -0.53 AS LEFT 7.70 11/21/2005 AS FOUND 7.78 90 0.08 0.08 - 0.08 AS LEFT 7.78 2/22/2006 AS FOUND 7.75 93 -0.03 -0.03 -0.03 AS LEFT 7.75 5/23/2006 AS FOUND 7.90 90 0.15 0.15 0.15 AS LEFT 7.90 8/22/2006 AS FOUND 7.60 91 -0.30 -0.30 -0.30 AS LEFT 7.60 _

11/22/2006 AS FOUND 7.65 92 0.05 0.05 0.05 AS LEFT 7.65 i 2/23/2007 AS FOUND 7.80 93 0.15 0.15 0.15 AS LEFT 7.80 5/23/2007 AS FOUND 7.70 89 -0.10 -0.10 -0.10 AS LEFT 7.70 8/22/2007 7.71 91 0.01 0.01 0.01 7.71, AVERAGE 0.00 0.00 0.00

__STANDARD DEVIATION 0.17 0.17 0.17

,COUNT 110 110 110 1%OF ORIGINAL DATA POINTS 100.00% 100.00%

195-/o/95% TOL. INT. PSI 0.38 0.38 INI_245.XLS Page 6

Attachment 2 PS-2390A & B Instrument Leg Head Correction Calculation Number IN1 -245 This attachment shows the method of solution, equations, values, tables and references used to determine "water head correction" for PS-2390A/B.

Method of Solution HP = EL

  • SG HP = hydrostatic head pressure EL = elevation difference SG = specific gravity of fluid EL (elevation bottom of T1 05A and T1 05B) - (elevation midpoint of PS2390A and PS2390B)

Assumptions Elevations from drawings used are assumed to be within +/-1"; calculations will round off towards the conservative value.

Calculation elevation T1 05A/B = 23ft Reference dwg. C338 Miscellaneous Structures Condensate Tank Details elevation PS2390A/B = 8.04ft Reference Attachment "K" SUDDS/RF#92-039 and verified by walkdown 12/17/98 EL = 14.96 ft elevation difference SG = specific gravity of CST water at 40°F (minimum water temperature will create largest head correction)

SG = 62.426 lbs/ft3 Reference ASME Steam Tables HP = (14.96ft)(62.4261bs/ft 3)

HP = 933.891bs/ft 2 or 6.48 psi, conservatively rounded to 6.5 psiq in PNPS 8.M.2-2.5.6 References

  • C-338, Rev. El, Miscellaneous Structures Condensate Tank Details
  • SUDDS/RF92-039, Rev. 0, Setpoint Calc for PS2390A,B from Bechtel Corporation
  • ASME Steam Tables, 5th Edition

"f*.*A'" " 1j'ev. 2 1A1//.1 -24 5- CC I. L. lisher October 19, 2987 02r John Torbeck SUJ1C Seisuic and Ocher tests applcalbe to MIDt 209W12771003 The StAtic-0-3ing pressure witch, MO1209A5127PO02, we@ asesmmiely togged and found to operate normally iaU ". "atsas up to U $S,. /

The accuracy of the devike I :t IX of fua rangel the tauporature offect Is

' 22 per 100'?. Teeting of the prlisure witich wea perform*ed at abtent temperatuze, 15601 and 2120r. (Normal temperature was considered to bo 70"1; 1.546 Ls the Z uam% m uge sed tebperature.)

o The pressure switches are quulified based on XZli Standards 323-1971 and 346-1971.

Tlst Results and qu*GlGficIaon data ato located Is DuI44P ecord f1l. (Don)

ADO-10-4, Index 37.

Material SerVIce. elifteerLuI C

iV

4rMc 1.Ar 4 /M1I2456 ABoston Edison z roov coMePNY NOTES OF TELECON To: Joel Bradley. SOR. Inc. 1913) 881-0767 From: Joseph Tedeschi. BECOIDE&S. (Sa)830-.830S Dale: January 11, 1999(15:001 andJamiaryl 11, 1999(09:15)

Subject. Range of SOR Press, Switches. DECO I0 Nos. PS.2390A. 0 JoG Tede:hh called Mr. Bradley to obtain additional info. on iteadjustabl range of pressure Swltches PS-2390A &6. SOR model no. SN-AA2-XSPP. The adjustable range specified lot the switches in BECO setpaint cat:. E-634.3. Rev. I (SUDDS #92-039) contas references to conflicting ranges at 2 - 25 psig (body of cale.). and 2 - 30 psig (VlSI Spec. shis in Attach. "HI.

A third n=-ft-ng saoure of informaton. the currant SOR catalog .identaies that the adjustable range of the switches should be 7 - 30 psig based on the 8 -2 piston - spring combination identified in tho model no.

Mr. Bradley identified that the adjustable range of these switches has changed over the years for the 6 - 2 Piston. spring combination and that in 1908. lia adjustable range for this type of switch would have been 2 - 25 psig as specified in the body of the referenced BECO caltc. This in Ihe appromxmate vintage of the switches installed at PNPSI.

Mr. Bradley also identified the make-up of the switch model no. as follows:

-. SN.*- X _

Piston Type \ Supplied with paper 10 Tag 5 "Speciar requirements apply Housing: Spring type 2 Weather tight, 314" RH NPT Elec. Conn.,

Similar to current type NO but DPOT swiCth has aluminum housing Details on the 5 -speciar requirements, can probably be obtained i the SOR Nuclear Group were contacted and provided with the specific equipment seral nroe.

C: B. Rancourt - OEM Group M01

ATTACHMENT 9.1 DESIGN VERIFICATION COVER PAGE Sheet I of 1 DESIGN VERIFICATION COVER PAGE Q ANO-1 El ANO-2 E IP-2 El IP-3 [I JAF E] PLP

[]PNPS 0 VY Q GGNS E-RBS E] W3 ENP Document No. IN 1-245 Revision No. Page 1 of 52 2

Title:

Setpoint Calculation for PS-2390A & B Condensate Tank Low Level Transfer

[a Quality Related [] Augmented Quality Related DV Method: Z Design Review El Alternate Calculation El Qualification Testing VERIFICATION REQUIRED DISCIPLINE VERIFICATION COMPLETE AND COMM6ENTS RESOLVED (DV print, sign, and date) i- Electrical El Mechanical Instrument and A-K Barrie /9-25-09 Control Li Civil/Structural r-I Nuclear Originator: b. .v Print/Sign/Date After Comments Have Been Aesolved EN-DC-134, Rev. 2

ATTACHMENT 9.6 DESIGN VERIFICA71ON CHECKLIST ATTACHMENT 9.6 DESIGN VERIFICATiON CHECKLIST Sheet I of 3 IDENTIFICATION: DISCIPLINE:

Document

Title:

Setpoint Calculation for PS-2390A & B Condensate Tank Low Level Transfer -'CivillStructural

'lElectrical IN 1-245 Rev. 2 QA Cat. IMhacal Doc. No.:

A-K Barrie 9-25-09 ENuclear Verifier: Print S Date 0Other Manager authorization for supervisor performing Verification.

C3 N/A Print Sign Date METHOD OF VERIFICATION:

Design Review EL Alternate Calculations 0 Qualification Test Q3 The following basic questions are addressed as applicable, during the performance of any design verification. [ANSI N45.2.11 - 1974] [NP] [QAPD, Part 11, Section 3][NQA-1-1994, Part 11, BR 3, Supplement 3s- 1I].

NOTE The reviewer can use the "Comments/Continuation sheet" at the end for entering any comment/resolution along with the appropriate question number. Additional items with new question numbers can also be entered.

1. Design Inputs - Were the inputs correctly selected and incorporated into the design?

(Design inputs include design bases, plant operational conditions, performance requirements, regulatory requirements and commitments, codes, standards, field data, etc.

All information used as design inputs should have been reviewed and approved by the responsible design organization, as applicable.

All inputs need to be retrievable or excerpts of documents used should be attached.

See site specific design input procedures for guidance in identifying inputs.)

Yes 0 No [3 N/A 13

2. Assumptions - Are assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are assumptions identified for subsequent re-verification when the detailed activities are completed? Are the latest applicable revisionsof design documents utilized?

Yes 0 No [] N/A []

3. Quality Assurance - Are the appropriate quality and quality assurance requirements specified?

Yes 0 No 1 N/A [I EN-DC-134, Rev. 2

ATTACHMENT 9.6 DESIGN VERIFICATION CHECKLIST Sheet 2 of 3

4. Codes, Standards and Regulatory Requirements - Are the applicable codes, standards and regulatory requirements, including issue and addenda properly identified and are their requirements for design met?

Yes 0 No [ N/A []

5. Construction and Operating Experience - Have applicable construction and operating experience been considered?

Yes0* No [ N/A [I

6. Interfaces - Have the design interface requirements been satisfied and documented?

Yes 0 No [I N/A 1]

7. Methods - Was an appropriate design or analytical (for calculations) method used?

Yes 0 No [ N/A El

8. Design Outputs - Is the output reasonable compared to the inputs?

Yes 0 No [I N/A []

9. Parts, Equipment and Processes - Are the specified parts, equipment, and processes suitable for the required application?

Yes 0 No 0 N/A 0l

10. Materials Compatibility - Are the specified materials compatible with each other and the design environmental conditions to which the material will be exposed?

Yes I No 0 N/A 0

11. Maintenance requirements - Have adequate maintenance features and requirements been specified?

Yes [ No [ N/A 0

12. Accessibility for Maintenance - Are accessibility and other design provisions adequate for performance of needed maintenance and repair?

Yes [ No [ N/A 0

13. Accessibility for In-service Inspection - Has adequate accessibility been provided to perform the in-service inspection expected to be required during the plant life?

Yes [ No 0 N/A 0

14. Radiation Exposure - Has the design properly considered radiation exposure to the public and plant personnel?

Yes 0 No 0 N/A [3

15. Acceptance Criteria - Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisfactorily accomplished?

Yes 0 No [ N/A [E

16. Test Requirements - Have adequate pre-operational and subsequent periodic test requirements been appropriately specified?

Yes 0 No 0 N/A 0l EN-DC-134, Rev. 2

ATrTACHMENT 9.6 DESIGN VERIFICATION CHECKLIST Sheet 3 of 3

17. Handling, Storage, Cleaning and Shipping - Are adequate handling, storage, cleaning and shipping requirements specified?

Yes 0 No 0 N/A 0

18. Identification Requirements - Are adequate identification requirements specified?

Yes 0 No 0 N/A 0

19. Records and Documentation - Are requirements for record preparation, review, approval, retention, etc., adequately specified? Are all documents prepared in a clear legible manner suitable for microfilming and/or other documentation storage method? Have all impacted documents been identified for update as necessary?

Yes 0 No [3 N/A [D

20. Software Quality Assurance- ENN sites: For a calculation that utilized software applications (e.g.,

GOTHIC, SYMCORD), was it properly verified and validated in accordance with EN- IT-104 or previous site SQA Program?

ENS sites: This is an EN-IT-104 task. However, per ENS-DC-126, for exempt software, was it verified in the calculation?

Yes E) No [I N/A0

21. Has adverse impact on peripheralcomponents and systems, outside the boundary of the document being verified, been considered?

Yes 0 No 0 N/A 0l EN-DC-134, Rev. 2

ATTACHMENT 9.7 DESIGN VERIFICATION COMMENT SHEET Sheet I of 1 Comments / Continuation Sheet Question Comments Resolution Initial/Date

___ I ___________ I __________ [______

4 4 4 4 4 4 EN-DC-134, Rev. 2 to Enterqy Letter 2.11.040 Marked-Up TS Pages and BASES Pages (7 Pages)

' 14ý PNPS TABLE 3.2.B (Cant)

INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Ivhiiiinurn oiof Opceiavul hrsLrmment Channels Per 46 Trip Syste. il I Trip Function Level Setling Remark S 2 Condenrate Storage Tank Low Level > ove lank zero Provides interlock to HPCI pump suction valves.

I 2 Suppiession Chamber High Level <- 1'1I" below torus zero 1 RCIC Turbine Steam Line High Flow s 300% of rated steam flow (2) 2 RCIC Turbine Compartment Wall 168F IF (2) 2 RCiC Exhaust Duct Torus Cavity .148F (2) 2 RCIC Valve Station Area Wall < 19a'F (2) 4 RCIC Stearn Line Low Pressure 77 > P > 63 psig (2)(5)(6) 1 HPCI Turbine Steam Line High Flow 5 296% of rated flow (3) 2 HPCI Turbine Comparnment Exhaust Duct -<168rF '(3) 2 HPCI Exhaust Duct Torus Cavity < 198°F (3) 2 Hl-PCURIiR Valve Station Area Exhaust Duct - 13 S16 r Aoleindinent No. Qj-"-4. i+9&3I42-1 314.2-16

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.13 INSERVICE CODE TESTING 4.13 INSERVICE CODE TESTING Applicability: Apolicability:

Applies to ASME Code Class 1, 2 and 3 Applies to the periodic testing pumps and valves. requirements of ASME Code Class 1, 2 and 3 pumps and valves.

'-I Objective: Obiective:

To assure the operational readiness of ASME To assess the operational readiness of Code Class 1, 2, and 3 pumps and valves. ASME Code Class 1, 2, and 3 pumps and valves by performance of inservice tests.

Specification: Spoecification:

A. Inservice Code Testina of Pumos and A. Inservice Code Testing of Pump and Valves Valves

1. Based on the Facility Commercial 1. The ASME OM Code terminology for Operation Date, Inservice Code Testing Inservice Test activities is as follows.

of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in Code Frequencies Terminology accordance with the Inservice Code

.1) Testing Program. Weekly Monthly 7 Days 31 Days Quarterly or 92 Days 3 Mths Semiannually/ 184 Days 6 Mths 9 Months 276 Days Yearly/Annually 366 Days Biannual/2 Yrs 732 Days

}JTýLt4I -ý t 7, 4

)

Amendment No. 14-97, 222-1 3/4.1 3-1I

) LIMITING CONDITIONS FOR OPERATION 3.13 INSERVICE CODE TESTIN *G SURVEILLANCE REQUIREMENTS 4.13 INSERVICE CODE TESTING The provisions in Definitions (1.0) for REFUELING INTERVAL, SURVEILLANCE FREQUENCY, and SURVEILLANCE INTERVAL are applicable to Code testing and to the above frequencies for performing Coder7 testing activities.

Performance of Code testing shall be in addition to other specified Surveillance Requirements.

K Nothing in the Inservice Code Testing I'* Program shall supersede the requirements of Technical Specifications.

6~ku

)

Amendment No. t3-4,-2-3-- 3/4.1!3-2

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3 REACTIVITY CONTROL (continued) 4.3 REACTIVITY CONTROL (continued)

) B. Control Rod Ocerability B. Control Rod Operability LCO 3.3.B.1 SR 4.3.B.1.1 Each control rod shall be OPERABLE.

NOTE ---------------------- --..------------------

Not required to be performed until 7 days APPLICABILITY: after the control rod is withdrawn and thermal power is greater than the LPSP of RUN and STARTUP MODES; REFUEL the RWM.

MODE when the reactor vessel head is

( fully tensioned. (See alsoo ACTIONS

ý- Insert each fully withdrawn OPERABLE control rod at least one notch once per 7 days.

NOTE---------------- SR 4.3.B.1.2 Separate condition entry is allowed for NOTE ---------------------- ----------------------

each control rod. Not required to be performed until 31 days after the control rod is withdrawn and thermal power is greater than the LPSP of A. One withdrawn control rod stuck. the RWM.

--.---..-------- NOTE

-- ------------------ Insert each partially withdrawn

/- Rod Worth Minimizer (RWM) may be OPERABLE control rod at least one notch bypassed as allowed by LCO 3.3.F. once per 31 days.

) 1. Verify stuck control rod separation criteria are met SR 4.3.1.1.3 immediately. Verify each withdrawn control rod does not go to the withdrawn overtravel position.

AND a. Each time the control rod is withdrawn to "full out" position.

2. Disarm the associated control rod drive (CRD) AND within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
b. Prior to declaring control rod AND OPERABLE after work on control
3. Perform SR 4.3.B.1.1 and rod or CRD system that could affect SR 4.3.8.1.2 for each coupling.

withdrawn OPERABLE control rod within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 4.3.8.1.4 from discovery of Verify each control rod scram time from condition A concurrent with fully withdrawn to notch position 04 is thermal power greater than  ; 7 seconds in accordance with the Low Power Setpoint SR 4.3.C.1, SR 4.3.C.2, SR 4.3.C.3 or (LPSP) of the RWM. SR 4.3.C.4 AND SR 4.3.1.1.5

4. Verify LCO 3.3.A.1 is met within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Determine the position of each control rod once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

AND 9)

Amendment No. t86, 20, 3/4,3-2

)BASES:

3.10 CORE ALTERATIONS W A. Refuelinq Interlocks

1. Refueling Equipment Interlocks BACKGROUND Refueling equipment interlocks restrict the operation of the refueling equipment or the withdrawal of control rods to reinforce unit procedures that prevent the reactor from achieving criticality during refueling. The refueling interlock circuitry senses the conditions of the refueling equipment and the control rods. Depending on the sensed conditions, interlocks are actuated to prevent the operation of the refueling equipment or the withdrawal of control rods.

One channel of instrumentation is provided to sense the position of the refueling platform, the loading of the refueling platform fuel grapple, and the full insert of all control rods, except control rods withdrawn in accordance with LCO . or fully inserted and disarmed. Additionally, inputs are provided for the loading of the refueling platform frame mounted hoist, the loading of the refueling platform monorail mounted hoist, the full retraction of the fuel grapple, and the loading of the service platform hoist. With the reactor mode switch in the shutdown or refueling position, the indicated conditions are combined in logic circuits to determine if all restrictions on refueling equipment operations and control rod insertion are satisfied.

A control rod not at its full-in position interrupts power to the refueling equipment and prevents operating the equipment over the reactor core when loaded with a fuel assembly. Conversely, the refueling equipment located over the core and loaded WV with fuel inserts a control rod withdrawal block in the Control Rod Drive System to prevent withdrawing a control rod.

The refueling platform has two mechanical switches that open before the platform or any of its hoists are physically located over the reactor vessel. All refueling hoists have switches that open when the hoists are loaded with fuel.

The refueling interlocks use these indications to prevent operation of the refueling equipment with fuel loaded over the core whenever any control rod is withdrawn, or to prevent control rod withdrawal whenever fuel loaded refueling equipment is over the core.

To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling the refueling interlocks, as designed, will prevent inadvertent criticality.

APPLICABLE SAFETY ANALYSES A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. Criticality and, therefore, subsequent prompt reactivity excursions are prevented during the insertion of fuel, Provided all control rods are fully inserted during the fuel insertion. The refueling interlocks accomplish this by preventing loading of fuel into the core with any control rod withdrawn or by preventing withdrawal of a rod from the core during 4fuel loading.

.1* Refueling equipment interlocks satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

R-3vision ~B/~- B3 / 4. 1()- I I

) BASES:

3.10 CORE ALTERATIONS (Cont)

W A. Refueling Interlocks (Cont)

1. Refueling Equipment Interlocks (Cont)

SPECIFICATION 3.10.A.1 REQUIREMENTS To prevent criticality during refueling, the refueling interlocks ensure that fuel assemblies are not loaded with any control rod withdrawn. To prevent these conditions from developing, the all-rods-in, the refueling platform position, the refueling platform fuel grapple fuel loaded, the refueling platform frame mounted hoist fuel loaded, the refueling platform monorail mounted hoist fuel loaded, the refueling platform fuel grapple fully retracted position, and the service platform hoist fuel loaded inputs are required to be operable. These inputs are combined in logic circuits, which provide refueling equipment or control rod blocks to prevent operations that could result in criticality during refueling operations.

The interlocks are required to be operable with the reactor mode switch locked in the "Refuel" position during in-vessel fuel movement with refueling equipment associated with the interlocks.

With one or more of the required refueling equipment interlocks inoperable (does not include the one-rod-out interlock addressed in Specification 3.1O.A.2), the unit must be placed in a condition in which the Specification does not apply or the interlocks are not needed. This can be performed by ensuring fuel assemblies are not moved in the reactor vessel or by ensuring that the control rods are inserted and cannot be withdrawn.

Therefore, 3.1 O.A. 1.a requires that in-vessel fuel movement with the affected refueling equipment must be immediately (i.e., in a time frame consistent with safety) suspended. This action ensures that operations are not performed with equipment that would potentially not be blocked from unacceptable operations (e.g., loading fuel into a cell with a control rod withdrawn). Suspension of in-vessel fuel movement shall not preclude completion of movement of a component to a safe position.

Alternately, 3.l O.A. 1.b requires that a control rod withdrawal block be inserted and that all control rods subsequently verified to be fully inserted. This action ensures that control rods cannot be inappropriately withdrawn because an electrical or hydraulic block to control rod withdrawal is in place. To the extent practicable, in the event of a failure(s) of an individual interlock, the effects of a failed interlock will be isolated to allow refueling activities to continue while the other interlocks are maintained availaole. As a result, the unaffected interlocks will continue to provide partial protection. Like 3.10.A. l.a these actions ensure that unacceptable operations are blocked (e.g., loading fuel into a cell with the control rod withdrawn).

. 4

ý0 petA Rqevision 232 ,93/41. !0-2

BASES:

3.10 CORE ALTERATIONS (Cont)

A. Refueling Interlocks (Cont)

2. Refuel Position One-Rod-Out Interlock BACKGROUND The refuel position one-rod-out interlock restricts the movement of co rol rods to reinforce unit procedures that prevent the reactor from becoming cri al during refueling operations. During refueling operations, no more than on control rod is permitted to be withdrawn except as allowed by Specification .D.

The refuel position one-rod-out interlock prevents the selection of a second control rod for movement when any other control rod is not fully inserted. It is a logic circuit that has redundant channels. It uses the all-rods-in signal (from the control rod full-in position indicators) and a rod selection signal (from the Reactor Manual Control System).

APPLICABLE SAFETY ANALYSES A prompt reactivity excursion durng refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment.

The refuel position one-rod-out interlock and adequate shutdown margin prevent criticality by preventing withdrawal of more than one control rod. With one control rod withdrawn, the core will remain subcritical, thereby preventing any prompt critical excursion.

The refuel position one-rod-out interlock satisfies Criterion 3 of 10CFR50.36(c)(2)(ii).

SPECIFICATION 3.10.A.2 REQUIREMENTS To prevent criticality, the refuel position one-rod-out interlock ensures no more than one control rod may be withdrawn. Therefore, the one-rod-out interlock must be operaby when any control rod is withdrawn (except as allowed by Specification S*. 3 ). The reactor mode switch must be locked in the refuel position to support tthehe operability of the interlock.

With the refueling position one-rod-out interlock inoperable, the refueling interlocks may not be capable of preventing more than one control rod from being withdrawn.

This condition may lead to criticality. Therefore, control rod withdrawal must be immediately suspended, and action must be immediately initiated to fully insert all control rods in core cells containing one or more fuel assemblies. Action must continue until all such control rods are fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted.

B3/4.10-3 I

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