ML20134D373
| ML20134D373 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/20/1996 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20134D254 | List: |
| References | |
| 96-0279, 96-279, NUDOCS 9702050110 | |
| Download: ML20134D373 (63) | |
Text
__
l NUCLEAR POWER. BUSINESS. UNIT CALCULATION REVIE,W Ab) APPROVAL
.. ~. _
. _. Calculation #
N.
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b 'O 9 7 'l Hu,jnber of Pages
.%+ \\ p.4keh m er.f*
Title of Calculation:
hr1Coksnh.$50ctaleWiM Si r v ate M '/h (/Se l i r IT41b7'"
% /v-p t
5 Original Calculation
)D. QA-Scope O Revised Calculation. Revision #
D Superseding Calculation. Supersedes Calculation #
Modification #
==
Description:==
Other
References:
Prepared B Date:
I M 9h his Calculation has been reviewed in accordance with NP 7.2.4. The review was accomplished by one or a combination of the following (as checked):
! A detailed review of the original calculation.
A review of a representative sample of repetitive calculations.
A review of the calculation against a similar A review by an alternate, simplified, or calculation previously performed.
appro timate method of calculation.
Comments:
g g M co m d pu.'fo ( 2 ( M f
- alA W v #* ""* w +.i,,
9702050110 970122 PDR ADOCK 05000301-P PDR s Date:
Date:
T
^
McMLea w it u_"
g>
ll.d..
Reference (s): NP 7.2.4 I
PBF-1608 d!
Revision 1 02/27/95 L_._.._.
a
. _ - - ~-
Review Comments on Calculation 96-0279, Rev. O The following comments were made on this calculation 96-0279 and resolved with the preparer:
1.
Section D I: Per DG-101, Section 3.3.3.15, if sufficient data is not available to perform an as-found/as-left drift analysis and there is no vendor data available, a default value of12% should be used for sensors and 11% should be used for rack components. DG-101, Section 3.3.3.12 allows use of engineeringjudgement where data is not available. The past calibration data discussed in this calculation adequately provides a basis for this engineering judgement.
2.
Section F. General The general methodology in DG-101 is to combine errors as a percent of the calibrated span (see examples in Section 3.3.1 and discussion in Section 3.3.3.5). In this calculation, percentage errors are calculated and then multiplied by the maximum flow rate to get an error in measured flow rate. This technique should yield conservative error values uhen considering flow rates below the maximum.
3.
Section F 1:
FE-128 - Westinghouse data sheet 1.6A lists a il% accuracy in range from 30 - 120 gpm and a differential pressure range of 200" H;O. Use of this orifice error in lieu of the conversions discussed in DG-101 Section 3.3.4.2 is conservative. The data sheet provides a reference for the 140 gpm maximum flow rate used in this calculation.
FT-128 a.
In the calculation of the calibration errors,it should be clarified that the 10-50 mA span is the calibrated span for the transmitter per ICP-04-003.4.
b.
Reference for the IT accuracy should be noted.
c.
The adjustment of the drift to account for differrnt calibration frequency should be done using the guidelines of DG-101, Section 3.3 4.3. It should be noted that calibrated span and the upper range limitare the same d.
The static pressure effects noted here are the random portion of theses errors. As discussed in DG-101 Sections 3.3.2.1 and 3.3.3.18 a non random portion of the error may based on the static head differences between the header and the transmitter, in ICP-04-003.4 a pressure is applied at the transmitter during calibration. As such the impact of any differences between the header and the transmitters is not considered in the calibration. Since we are orly concerned with measurement of differential pressure, this component of the error should be very small and can be excluded per DG-101 Section 3.3.3.9.
e.
The vendor static errors are not treated exactly as noted in DG-101. The difference in handling the static pressure will not significantly affect the results. Since the calibrated span and the upper range limit are the same, the Span Shill Bias calculation is acceptable.
f.
As discussed in the comments on FM-128, conversion of this error to a flow rate is not appropriate at this point.
FM-128 - As shown in DG-101 Appendix D, the output error of a non-linear device should be calculated using the input errors (i c., the SRSS of errors for devices upstream of the non-linear device), the transfer function for the device and the errors of the device. The transfer function from DG-101should be used in the calculation.
-1
Revier. Com. rents on Calculation 96-0279, Rev. 0 1
FI-128 I
a.
In the discussion of the indicator accuracy you reference a telecon with Karen Depodesta. A review of select pages from reference 9 confirms that the information in the telecon is appropriate.
b.
The error value allowed in ICP-04 032.1 is 12.8 gpm and the calibrated span is 4-140 gpm. This gives 12.8/136 = i2.06%
c.
The terms under the square root should oc shown squared.
4.
Sections F.2:
a.
The calibrated range for F1-115 and F1 116 is 20 to 198" H 0 input and 6.32 to 19.90 gpm output.
b.
Errors should be expressed in % rather than gpm.
4 c.
In item a, the indicator readability should be in gpm not psi.
4 d.
Per Westinghouse data sheet accuracy of flow elements is 11%
5.
Section G i
a.
In item a, if errors for FM128 are calculated as noted above, only the error for FM128 l
and Fil28 need to be used in the loop accuracy calculation.
i j
i 6.
Attachment The calculation number and page number should be added to the attachment.
I
)
3 1
4 3
4 d
i a
i 8
,_.~..
TITLE: Uncarttinty Associated CALCULATION * : 96-0279 with Instrumentation Used in Prepared By: EJM IT-21 & IT-22 for Charging Pumps Date;12/20/96 Page iof9 A. Purpose The purpose of this calculation is to determine the uncertainty associated with the instrumentation used in inservice test procedures IT-21 & IT-22, for the charging pumps, (reference 2). The final uncertainty value must include a combination of the uncertainties for all the instrumentation used in the test that will have an impact on the ability to measure the IST acceptance value during the test performance for comparison to the design I
basis acceptance value.
B. Method Instrument uncertainties will be calculated for the instruments used in the l
charging pump Inservice Test procedures, IT-21 and IT-22, (reference 2).
i The charging pump acceptance value will be strictly a total flow value. The charging pump total flow is determined in IT-21 and IT-22 by adjusting the pump speed to 1350 10 rpm using a portable tachometer and adding the flows from three instruments, F1-128, FI-115 and FI-116. The uncertainties for these three instruments will be determined, and then combined, using a square root sum of the squares technique. The result will be added to the design basis pump performance requirement (currently defined in reference
- 1) to establish an IST acceptance criteria for the charging pumps that will ensure they can meet their design basis performance requirement. During the performance of IT-21 and 22, the operator willincrease the charging pump flow to a value that is greater than the design basis acceptance requirement, plus instrument uncertainties. This test will show that the charging pumps are capable of satisfying their design basis requirement.
Since the charging pump design-basis acceptance value is strictly a flow value, and the charging pumps are positive displacement pumps whose flow output is nearly independent of the discharge pressure seen by the pump, it is not necessary to consider any pressure instruments in this uncertainty calculation. it is not necessary to compare the pump to a pump curve, or prove the design basis requirement can be satisfied at a particular pump speed, since the purpose is only to veriN that the pump can satisfy its design basis function, which is strictly a flow value, it is not important what the rpm of the charging pump is at the time of the IST test to determine that the design basis requirement has been catisfied. The instrument uncertainties associated with the charging pump speed, and the instruments used to measure that speed, are not considered in this calculation. The pump speed remains important to ensure ASME Section XI acceptance criteria are satisfied, since these criteria are established to monitor for pump degradation.
This calculation identifies the Unit 1 instrumentation, however the calculation applies equally to the charging pumps on either unit.
TITLE: Uncertrinty Associstsd CALCULATION # : 96-o279 with Instrumentation Used in Prepared By: EJM IT-21 & IT-22 for Charging Pumps Date:12/20/96 Page 2 of 9 C. References 1.
Charging pump operability determination, 10/3/1996.
2.
PBNP Inservice Test, IT-21 and IT-22, " Charging Pumps and Valves Test, (Quarterly)" Unit 1 & 2, Revision 4, April 13,1995.
3.
E-mail, Craig Neuser to Ed Mercier, Subject " Indicators", dated 12/10/1996, attached.
4.
PBNP instrumentation and Control Procedures, ICP-04.003-4, " Charging Flow Transmitter and Indicator Outage Calibration, " Rev 1, August 2, 1995.
5.
PBNP instrumentation and Control Procedures, ICP-04.003-7, "RCP A & B Seal Water And Letdown Flow Instruments Outage Calibration," Rev 3, August 27,1996 6.
Rosemount Instruction Manual MAN 4258, Model 1151HP Alphaline Differential Pressure Transmitters for High Line Pressures, January 1988.
7.
DG-101 " Instrument Setpoint Methodclogy", Revision 1, September 12, 1995.
8.
Westinghouse Specification Sheets. Data Sheet 1.6 dated 7/18/69 9.
VECTRA Calc No. PBNP-IC-07, " Westinghouse 252 Indicator Drift Calculation", Rev 0, 6/9/1995
- 10. Duke Engineering & Services letter to WE, "Si Pump IST Flow Test Uncertainty Evaluation", September 25,1996.
- 11. WE Calculation 96-0191, " Minimum Allowable IST Acceptance Criteria for S1 Pump Performance", dated 9/25/1996.
- 12. Foxboro Component Instruction Manual, Control #00623, Model 66A Square Root Converter, section 18-650, Feb 1969, page 1.
' 13. PBNP instrumentation and Control Procedures, llCP-04.032-1, " Auxiliary Feedwater System and Charging Flow Electronic Outage Calibration," Rev 0, February 22,1995.
- 14. Barton Component instruction Manual, Control #001035, TF18.5, Model 200 Differential Pressure Indicator.
- 15. Westinghouse Specification Sheets for 1FE-128, Data Sheet 1.6A dated 7/30/69.
D. Assumptions 1.
The temperature effect on the instrumentation will be assumed to be negligible as the transmitters are calibrated and used in essentially the same temperature environment.
2.
If manufacturer's data was not located, uncertainties associated with drift of an instrument have been assumed to be the smaller of either 0.5% of full scale, or the instrument calibration tolerance. This value (0.5%) is based on engineering judgment of the maximum expected drift between calibrations for the instrumentation involved. Alternatively, the calibration accuracy is used if smaller, because instrumentation found regularly out of
TITLE: Uncsttrinty Associtted CALCULATION * : 96-0279 with instrumentation Used in Prepared By EJM IT 21 & IT-22 for Charging Pumps Date:12/20/96 Page 3 of 9 calibration are typically either repaired or replaced. Reviews of past instrument calibration sheets for the instruments in this calculation have shown the drift to be less than 0.5% in nearly all cases.
3.
The M&TE error is assumed to be the smaller of either 0.5% of the instrument range, or the calibration tolerance, for all IST instruments. This value (0.590) is conservative based on the research performed for Calculation 96-0191, " Minimum Allowable IST Acceptance Criteria for SI Pump Performance" (reference 11). The calibration accuracy is used if smailer because it is the practice of l&C to use a calibration instrument which is at least as accurate as the instrument being calibrated.
E. Inputs For this calculation, the total uncertainty associated with the instrumentation used to perform the IST test must be accounted for when obtaining the minimum IST acceptance criteria Contributors to this total uncertainty include:
Instrument (transmitter & indicator) accuracy e
Indicator readability Tolerance Drift F. Instrument Uncertainty Determinations 1.
Instrument Uncertainties for F1-128, Charging Flow. The uncertainties for the entire instrument loop, which includes the flow orifice FE-128, the flow transmitter FT-128, the ill square root converter FM-128, and the flow instrument F1-128, will be evaluated and combined using a square root sum of the squares method.
FE-128. Charging Line Flow Element Daniel Orifice Fitting Co., model #520, The accuracy is 1.0% (reference 15).
11.0%
U128 j
=
FT-128. Charging Line Flow Transmitter, Rosemount, Model
- 1151HP5G2201 (The calibrated range is 0-200" H 0,10-50 mAmp).
2 (reference 4)
FT-128 measures differential pressure, and outputs in amps. The square root conversion in the loop is done separately, in FM-128, and thus is not part of the transmitter.
a.
Instrument accuracy (which includes combined effects of linearity, hysteresis, and repeatability) is 10.25% of calibrated scale (reference 6)
i TITLE: Uncsttrinty Associated CALCULATION # : 96 0279 with Instrumentation Used in Prepared By: EJM IT 21 & IT-22 for Charging Pumps Date:12/20/96 Page 4 of 9 b.
Calibration Setting Tolerance; The as left tolerance for the instrument is 0.2 mAdc. This represents 0.5% of the calibrated range (0.2mA / 40 mA), (reference 4)
/
c.
Drift (transmitter stability); 0.25% of upper range limit for 6 months. (reference 6) Based on a yearly calibration, and a 25%
window on the calibration frequency:
1, Drift = 1.25 x 12/6 x 10.25% = 10.625%
/
d.
Static pressure effect Zero Error: 22.0% of upper range limit for 4500 psi i
1 (reference 6).
Span Error: 10.25% of upper range limit per 1000 psi. Assuming a discharge pressure of 2000, this would be 0.5 %
(reference 6).
e.
M&TE Error, Minimum required M&TE tolerance is 1.0" (reference 4), which represents an error of 0.5% (1"/200").
]
Urrus =
(0.25 %)2 + (0.5 %)# + (0.625 %)2 + (2.0%)2 + (0.5 %)# + (0.5 %)#/
Ugr o, =
2.28 %
i l
i input errors to Square Root Converter.
4 i
= h(Urn 28) + (U,7us)#
a (0.01)# + (0.0228)#
2.49 %
/
a =
=
FM-128. l/l Souare RooLConverter for Charging Line Flow, Foxboro, Model 66AC-0. The calibrated range is 12.50 mA - 50 mA input, and 10 - 50 mA output. (reference 13) i a.
Accuracy: The accuracy is 0.5% (reference 12) b.
Calibration Setting Tolerance; The as left tolerance for the instrument is 0.2 mAdc. (reference 13) This represents a 0.5% input error, (0.2mA / 40mA).
c.
Drift; No information was found in the component manual, so 0.5% was assumed. (assumption 2).
d.
M&TE:
0.5% was assumed (assumption 3).
.~
TITLE: Uncarttinty Associttzd CALCULATION # : 96-0279 with Instrumentation Used in Prepared By: EJM IT-21 & IT-22 for Charging Pumps Date:12/20!96 Page 5 of 9 l
(0.5)# + (0.5)2 g(0.5)# + (0.5)#
Urui28 =
+
p Urui28 =
1.0%
Using the transfer function from reference 7 for a square root converter:
Eq. 1 b =g[(a / 28)#
2
+e Where b = Output error from non -linear device a = Input error to non linear device B = Point of Interest (0 - 100% of span = 0 to 1) e = Device Uncertainty from non-linear device Reviewing past IST tests, a typical value for this instrument is 29 gpm.
Based on this, a point of interest of 29 gpm will be used.
9 B
=.2071
=
=
=
Instrument range 140 gpm 100 %
Evaluating Equation 1.
= gf0249 / 2*0.2071)# + (.01)2 7
b i
b =.0609 = 6.09 %
F1-128. Charging Line Flow Indicatur, Westinghouse Model #252, calibrated range of 4 to 140 gpm (reference 13) a.
Indicator readability: Based on plant walkdown by Craig Neuser, the smallest divisions or the meter f ace are at 2 gpm intervals.
(reference 3) Therefore, the instrument is read accurately to within 1 gpm, or 0.74% of calibrated range.
b.
Indicator Accuracy; 1.028% (reference 9)
/
This error also includes M&TE error and drif t, based on telecon on 12/17/1996 with Karen Depodesta, Duke Engineering & Services.
c.
Calibration' Tolerance; 2% of 140 gpm = 2.8 gpm. (reference 13)
Since the calibrated span is 4 gpm to 140 gpm; 2.8 gpm Calibration Tolerance
=.0206 = 2.06%
/
=
136 gpm
TITLE: Uncertzinty Associstsd CALCULATION # : 96-o279 with Instrumentation Used in Prepared By: EJM IT-21 & IT-22 for Charging Pumps Date:12I20196 Page 6 of 9 t
Uni 28 = [(0.0074)2 + (0.01028)* + (0.0206)
Uni 28 = 0.0242 = 2.42 %
2.
Instrument Uncertainties for 1-F1-115,1P-1 A seal injection flow, Barton Instruments Corporation, model 200, calibrated range of 20" H O to 198" 2
H O input, 6.32 gpm to 19.90 gpm output. Ireference 5).
2 Indicator readability; Based on plant walkdown by Craig Neuser, the a.
meter face has divisions of 1 gpm,0.2 gpm, and 0.1 gpm, dependent on the range used. (reference 3) Between 4 gpm and 10 gpm, the indicator range is.2 gpm per division. Based on past completed IT-21
& 22 tests, the reading for this instrument typically f alls between 6.5 gpm and 8.7 gpm. Since this is not a linear meterface, instead of using half of a division for the reading accuracy, it is assumed that the reading accuracy is equal to the smallest division, or 0.2 gpm.
.2 gpm Readability (% of calibrated range) 1.47% /
0.0147
=
=
=
(19.9 gpm - 6.32 gpm)
- b. Instrument accuracy, 0.5% of full scale (20 gpm) or 0.1 gpm (reference 14) 0.5%
- 200" H O 2
Accuracy (% of calibrated range) 1 0.56 /
=. 0.00561
=
=
(198" H O - 20" H O) 2 2
- c. Calibration Tolerance: The as lef t tolerance for the instrument is 10.1 gpm, which represents 0.5% of full scale (reference 5) 0.5%
- 200" H O 2
Calibration (% of calibrated range)
= z0.00561
- 0. 5 6 % -
=
=
(198" H O - 20" H O) 2 2
- d. Drift; assumed to be 20.1 gpm due to calibration tolerance, which represents 0.5% of full scale. (assumption 2) 0.5 % + 200" H O 2
0.56 % /
Drift (% of calibrated scale) 0.00561
=
=
=
{198" H O - 20" H O) 2 2
e.
M&TE (Instrumentation uncertainty due to calibration):
1.0" H 20, (reference 5)
I 1.0" H O 2
0.56% /
M&TE (% of calibrated range) 0.00561
=
=
=
(198" H O - 20" H O) 2 2
f.
FE-115, Daniel Orifice Fitting Co., model #520, (flow orifice associated l'
with FT-115).
i The accuracy is 1.0% (reference 8)
+1.0%
- 200"
=
FE - 115(% of calibrated range) 0.0112 1.12
=
=
(198" H O - 20" H O) 2 2
I
. _ ~
TITLE: Unczrttinty Associttzd CALCULATION # : 96-0279 with instrumentation Used in Prepared By: EJM IT-21 & IT 22 for Charging Pumps Date 12/20/96 Page 7 of 9 Since the bellows in the indicator effectively acts as the square root converter, it is necessary to treat this instrument as a square root converter. All the above errors with the exception of the indicator readability, are treated as input errors, input errors:
= g/(0.0056)2 + (0.0056)2 + (0.0056)2 + (0.0056)# + (0.0112)#
a a=
0.0158 = i 1.58%
/
Using the transfer function for a square root device from Appendix D of reference 8 7
Eq. 1 b =[(a / 28)2 +e Where b = Output error from non -linear device a =loput error to non -linear device B = Point of Interest (0 - 100% of span = 0 to 1) e = Device Uncertainty from non -linear device To determine the point of interest, it is necessary to look at Equation 1 and recognize that the smaller that B is, the greater that the output error will be.
From a review of past IST data, the lowest value for this reading was 6.4 gpm Point of interest 6.4 gpm 32 %
= 0.32 B
=
=
=
Instrument range 20 gpm 100 %
Evaluating Equation 1.
l (0.0147)2 g(0.0158 / 2 0.32)2 b
4
=
/
2.87 %
b = 0.0287
=
This error converted to gpm:
10.57 gpm 2.87%
- 20 gpm b
=
=
/
This error also applies to FI-116, since the instruments and calibration methods are the same.
G. Calculation
i
~.
.~
l TITLE-Uncartainty Associated CALCULATION # : 96-0279 with Instrumentation Used in Prepared By: EJM IT 21 & IT-22 for Charging Pumps Date.12/20/96 Page 8of9 l
l l
The uncertainties of the Inservice test instrumentation has been determined above, and will be combined using a systematic method established in reference 7 and reference 10. This best estimate or realistic approach combines uncertainties using the statistical square root sum of squares (SRSS) method.
i This uncertainty value will be added to the design basis charging pump flow requirement and this will become the IST design basis limit, and will be used as an acceptance value for the charging pumps, a.
Loop Uncertainty associated with Fl-128 (see Section F.1) g[(Urm28) + (Un,2a)
U128
=
fg(0.0609)# + (.0242)2 U178
=
6.55% of 140 gpm range = 9.17 gpm
/
.0655 Ui28
=
=
b.
Uncertainty associated with F1-115 (see Section F.2) 0.57 gpm
/
U115
=
c.
Uncertainty associated with F1-116 (see Section F.2)
U116 =
0.57 gpm
- d. Combining the uncertainties from these three flow instruments gives the following for total uncertainty:
gh128)2 + (U115)2 + (U 116)'
Utotal
=1 f
1 g(9.17 gpm)2 + (0.57 gpm)# + (0.57 gpm)#
Utotal
=
Utotal 9.21 gpm y
=
H.
Results The totalinstrument uncertainty associated with the inservice test procedure for the charging pumps is. 9.21 gpm
_. - ~. __
TITLE: Uncsttiinty AssociltId CALCULATION # : 96-0279 with Instrumentation Used in Prepared By: EJM IT-21 & IT 22 for Charging Pumps Date:12/20/96 Page 9 of 9 s............_.........................&
Printed For:
Date: Tranday, 10 December 1996 9:06am CT To:
From:
subject: Incicators j
Flow indicators FI-115/116:
0 - 4 gpm. 1 gpm divisions 4 - 10 gpm:
.2 gpm divisions 10 - 20 gpm:
.1 gpm divisions Flow indicator FI-128:
0 - 140 gpm: 2 gpm divisions 1
If any more info is needed please let me know.
I
- _. - _ -.. ~. - _. - -. -. - - - - - -. _ - _
j.
i CR 9/.s M
1.
Degraded or potentially nonconforming equipment:
Charging Pumps Ul&2 P 2A P-2B P-2C Safety function (s) performed:
(
internal check valves provide a contamment isolation pressure boundary following a LOCA.
No safety related functions to providing flow.
He charging pumps are needed to bring the unit to cold shutdown with the required shutdown margin at any time during l
core life, assuming that the most conservative control rod is stuck in the fully withdrawn position.
Provide RCP seal flow and makeup for RCS leakage.
I 3.
Circumstances of potential nonconformance, including possible failure mechanisms:
Condition Repott 96-416 identified a potential concern for adequacy of the IST program to ensure that pumps meet design basis as well as ASME Section XI requirements. His evaluation supports determination of operability pending completion i
of detailed analysis 4.
Requirement or commitment established for the equipment, and why it may not be met:
Te::hnical Specifications Section 15.3.2 provides the design basis for CVCS control of RCS Boron inventory. He boration volume available through any flow path is sufficient to provide the required shutdown margin at cold shutdown, Xenon free conditions from any expected operating condition. He maximum volume requirement is associated with boration from just critical, hot zero or full power, peak xenon with control rods at the insertion limit, to xenon-free, cold shutdown with the highest worth control rod assembly fully withdrawn.
l FSAR 14.3.1 states that makeup flow rate from two charging pumps is typically adequate to maintain pressurizer level long enough for the operator to respond without activating the ECCS for a break through a 3/8" diameter hole.
1 Generic Letter 83-28 Required Actions Based On Generic Implications Of Salem ATWS Events 5
IST acceptance criteria may not be conservative when compared to design basis criteria.
5.
How and when the potentially nonconforming equipment was first discovered:
His generic concern was first identified in June 19% as a specific concem for safety injection pump acceptance criteria from ASME Section XI versus design requirements his generic concern was first identified in June 1996 6.
Basis for declaring affected equipment operable:
'l A. During normal operation the charging system is required to provide 32 gpm flow per pump. Normally two pumps are in operation. 32 gpm flow per pump is based on providing sufficient flow for reactor coolant system makeup, based on the sum of the following:
12 gpm reactor coolant pump seal leakage. This is twice the normal leakage of 3 gpm per pump. Also, reactor coolant gmp operation is only allowed if reactor coolant pump seal leakage is <5 gpm per OP-4B (line 2.3.2).
10 gpm labyrmth seal flow. A labyrinth seal flow is necessary to prevent reactor coolant from entering the seal area. 10 gpm i
is the sum of the normal design labyrinth seal flow from each pump (5 gpm each). (E KIM # 102, Section 5, line 5.5.7) 10 gpm reactor coolant system leakage. His is based on Technical Specification Section 15.3.1.D.
B. Technical Specifications Section 15.3.2 provides the design basis for CVCS control of RCS Boron inventory, ne boration volume available through any flow path is sufficient to provide the required shutdown margin at cold shutdown, Xenon free conditions from any expected operating condition. De maximum volume requirement is associated with PBF-1553
{tevision 0 06/24/94
}
l boration from just critical, hot zero or full power, peak xenon with control rods at the insertion limit, to xenon-free, cold I
shutdown with the highest worth control rod assembly fully withdrawn. Calculation P-93-014 was performed to show that for a typical cycle, assuming worst case conditions, the reactor can be maintained suberitical following a reactor trip.
Specifically, the amount of negative reactivity that can be insetted by one charging pump borating at a minimum speed (15gpm) using the refueling water storage tank (RWST) as the suction source is greater than the positive reactivity added from the decay of xenon. Calculation P-93-014 showed that at a rate of 15 gpm from the RWST the intent of tech spec 15.3.2 for maintaining shutdown margin has been satisfied. He calculation did not look at the capability of the boric acid storage tanks ability to provide shutdown margin. He RWST is more limiting for a minimum flow requirement than the Boric acid storage tanks because boric acid concentration is maintained higher in the boric acid storage tanks than the RWST.
The reactor coolant pump seal leakage is not flow that can be considered to be added to the RCS. Therefore the total charging flow is required to be 15 gpm + 12 gpm = 27 gpm. De 32 gpm flow mentioned above is limiting as it envelopes the shutdown margin capability flow of 27 gpm.
C. FSAR 14.3.1 states that makeup flow rate from two charging pumps is typically adequate to maintain pressurizer level long enough for the operator to respond without activating the ECCS for a break through a 3/8" diameter hole. The original basis for capacity of the charging pumps was the ability to makeup to the RCS for normal charging system flow requirements (30 gpm), RCP seal injection flow requirements where the seal injection flow consists of the normal RCP labyrinth flow (5 gpm per pump), maximum seal flow through one RCP seal (75 gpm, assuming pumps equipped with floating ring seals which are not applicable to PBNP), and twice the nominal no. I seal flow in the remaining pump (6 gpm). This total of 121 gpm results in a requirement of 60.5 gpm per charging pump. Neither this requirement nor the 32 gpm limiting flow requirement stated above is related to the FSAR Chapter 14 statement relating charging system performance and a 3/8 inch reactor coolant system hole. The FSAR states that two charging pumps are typically adequate to maintain pressurizer level long enough for the operator to respond without activating the ECCS for a break through a 3/8 inch diameter hole. This statement does not mean that pressurizer level will be maintained constant or that the charging pumps will be able to meet the volume of such a leak. He statement in the FSAR concerning charging system capability is therefore only a general capability statement and not a design basis requirement. A calculation (N-90-015 Response Time for 3/8" Line Break in l
Reactor Coolant System) was perfonned to determine the amount of time available for operator action with two charging pumps available. The calculation concluded that approximately 30 minutes is available for operator action at no load conditions and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is available at full power. This calculation assumed a design flow rate of 60.5 gpm from each of two charging pumps. The maximum flow rate from a 3/8"line was determined to 17.5 lbm/sec or 126 gpm reference cale m l
NCR N-89-187. Note that although the 3/8" piping was upgraded to safety related and eliminated the need to include i
charging flow for 3/8" line break in the design basis, the statement in the FSAR is still true in that two charging pumps operating at their design maximum flow is typically adequate time for operator response. See attached response to Nonconformance Report N-89-187 for additional information.
D. Generic Letter 83-28 (Required Actions Based On Generic implications Of Salem ATWS Events) required each licensee to submit a report describing how it meets the requirements contained within the Generic Letter. A letter from C.W. Fay Vice President Nuclear power to the NRC dtd 11/1/83 contained this report. One of the sections required to be addressed in the Generic Letter concemed post maintenance testing. The report stated that it is current practice to perform PMT on safety related pumps and valves. This testing is performed in accordance with the guidelines of ASME Section XI. The same acceptance criteria used in evaluating performance during periodic testing is used in evaluating performance after maintenance. Current practice at PBNP is to perform periodic testing in accordance with Section XI and to perform PMT to the Section XI requirements after maintenance. The charging pumps are called upon in procedure CSP S.1 " Response to Nuclear Power Generation /ATWS" as a potential boration path for ATWS. A review of the Westinghouse Owners Group Emergency response guideline for FSR S.l does not call out any specific flow requirement for the charging pumps. It provides guidance for various means of borating the unit through emergency boration paths to the charging pumps as one of the titernatives. A review of PBNPs response to GL 83-28 did not find any flow requirements for the charging pumps.
Based on the above and the fact that the charging pumps are tested in accordance with ASME Section XI the commitments conceming ATWS have been met.
E. He IST program tests the charging pump flow against normal system operating pressure. The pumps are tested at 1350 rpm plus or minus 10 rpm. This is below the maximum RPM of the pump. The most limiting pump from attached IST data provided 42.6 gpm at 1350 rpm against normal system operating pressure. This is above the reauired limiting design basis of PBF.1553
. Revisiorf 0 06/24/94
32 gom and thus the charcine numns are considered onerable.
Pump Design IST Required Action Actual Results Basis Range 1P2A 32 gpm 41.39 - 48.95 gpm 42.8 gpni iP2B 32 gpm 39.80 - 47.08 gpm 42.6 gpm 1P2C 32 gpm 41.10 - 48.55 gpm 43.6 gpm 2P2A 32 gpm 39.62 - 46.86 gpm 43.6 gpm 2P2B 32 gpm 40.30 - 47.60 gpm 44.8 gpm 2P2C 32 gpm 39.06 - 46.20 gpm 43.1 gpm Prepared By:
Date: [O /
4
/C! _
Approved By:
Date:
n e.. t-7
^
Reviewed By:
Date: /#
DCS PBF.1553 Revision 0 06/24H
.-Ul Od IUt 11:0V Wl%UfolN ct.cbiluu hrou rnA nu, 919ccicutu
- r. ua i
j Charging Pump Flow Acceptaaec Criteria l
' Westinghouse used the following flows to determine the charging pump capacity (see attached DBD worksheets):
l
- 30 gym (chargmg flow) l
- 10 gpm(labyrinth flow) j
- 75 3pm (marimum seal flow in one reactor coolant pump)
- 6 gym (twice the " norma!" seal flow (3 sym) in the other pump)
This totals 121 gpm, or 60.5 spm per pump based on two operable charges pumps (out of three total). This i
dearmination was made before Me l seals were quahfied for reactor coolant pump applicanons, hence the a
i-very large 75 gpm seal leakage assumption, and the charsmg pumps were bought on this basis. Since the PBNP -
reactor coolantpumps have EA H-1 seals, the following flows can be used to determme the current required j
charging pump capacity:
i
- 30 gpm (chargmg flow)
- 10 gym (labyrmth flow)
- 12 gym (twice the normal seal flow for both reactor coolant pumps) 1 This totals 52 gym. or 26 spm per pump. To arrive at an IST pump flow =+p-e value, it is appropriate to aaneidar only the (1) reactor coolant system inventory control and (2) boric acid addition functions of the CVCS i
(i.e., the " Chemical" and " Volume" control functions of the CVCp Letdown flow does not need to be considered l
because it is related to RCS purification, and this RCS chemistry hmits are addressed by other Technical Specifications (Section 15.3.1.C). This number can be arrived at as follows:
1.
e j
- 12 gpm (makeup for seal leakage - the maximurn allowed by procedure is 10 gpm per pump) +-
i
- 10 gym (labyrmth seal flow - !abyrinth seat flow is required when maintaining seal flow) j
- 10 gpm (Technical Specification 15.3.1.D allowed leakage)
This totals 32 spm per pump. In addition to maintaining RCS inventory, this flow is considered a reasonable acceptance flow doce it also envelopes the following:
(1)The)fric acid transfer pumps are in-<eries with the changing pumps. The BATPs have an IST required flow of@gpm. 32 gpm envelopes this flow.
(2)It envelopes Appendix R evaluations, which assume a 20 gpm flow during fires.
(3)It also envelopes the flow required to accomodate either the assurned RCS contraction during cooldown (30 gpm) and auxiliary spray (30 gpm) flow, but not both together.
See the attachment for why the charging pump capacity does not need to be tied to providing makeup to accommodate 3/8 inch reactor coolant systern line breaks.
s' Con Room Chilled Water Pump Flow Acceptance Criteria,(P-112A, P-1128)
/
Control roont g19 a safety-related function andis y related to any regulatory requirements (i.e. post-TMI).
chilled water kmT gn flow of I16 spm @ 55 feet. His flow was required to accomodate removing a
, lead of Btu /hr. De latest calcidations indicate that the load on the heat exchangers is about 43 116 spm would be a reasonable flow to use, although a flow s
somewhat less would be actuallower-than-design room cooling load.
Cable Spread'
' Chilled Water Pump cceptance Criteria (P-111A,111B)
Cable s
-/l g room cooling is not a safety-related fun ' n is not directly related to any regulatory requir ents. De cabic spreading room chilled waterpump ign flow of 96 gpm @ 53 feet. Bis would be reasonable for to use.
Pump Flow Reference
- FLOW TEST **
Values and Limits Date Established:
3/02/93 Pump #: IP2A IT-021 Entered By: LEH REFERENCE VALUES Pressure:
Flow:
44.50 gpm Vibration:
Point A:
.097 ips l 3fc t /D </3 Point B:
.101 ips Point C:
.087 ips ACCEPTABLE RANGE Flow:
42.28 gpm to 48.95 gpm Vibration:
Point A: s
.243 ips Point B: s
.253 ips Point C: s
.217 ips ALERT RANGE Low Flow:
41.39 gpm to 42.28 gpm High Flow:
48.95 gpm to 48.95 gpm Vibration: Point A:
.243 ips to
.582 ips Point B:
.253 ips to
.606 ips Point C:
.217 ips to
.522 ips REQUIRED ACTION RANGE Low Flow:
41.39 gpm j
High Flow:
48.95 gpm Vibration: Point A: >
.582 ips Point B: >
.606 ips Point C: >
.522 ips 1
i COMMENT: Criteria from ASME OMb-1989, Part 6.
i
_m..m_.
_...__.._.___.....__.m..__-
.._m-
_ _ _ _ - _._... m.
TEST DATA FOR ONE PtMP 9/06/96 Page 1
P.unpi 1P2A Testr 021 Fiow Test Vibrations tipal Test Date Flow A
B C
Int Remarks 10/29/92 45.0
.100
.100
.080 LEH ROUTINE SURVEILLANC 3/01/93 44.5
.097
.101'
.007 6/29/93 46.0
.107
.106
.083 9/03/93 44.9
.109
.151
.077 9/17/93 44.7
.127
.106
.092 12/02/93
.133
.111
.086 3/05/94 45.9
.155
.125
.097 9/07/94 42.7
.123
.096
.086 LEN ROtTTINE SURVEII.IANC
(
12/05/94 42.2
.096 156
.103 LEN ROUTINE, 2P-2A NO 9 3/05/95 43.1
.097
.095
.087 LEX ROUTINE SURVEILLANC 6/02/95
.095
.096
.079 IAD ROUTINE SURVEILIANC 9/06/95 45.8
.103
.099
.007 LEN ROUTINE SURVEILIANC 12/05/95 46.0
.154
.096
.078 BAT ROUTINE SURVEILIANC 3/07/96 45.4
.125
.098
.077 LEH ROUTINE SURVEILLANC 6/05/96 43.7
.094
.086
.083 LRD ROUTINE SURVEILIANC 6/15/96 42.8
.101
.095
.076 IAD POST MAINTENANCE FO 9/05/96 42.8
.105
.088
.084 LEX ROUTINE SURVYILLANC 1
i i
f l
i
- i
Pump Flow Reference
- FLOW TEST **
Values and Limits Date Established:
5/22/96 Pump #: 1P2B IT-021 Entered By: LEH REFERENCE VALUES Pressure: 1975.00 Flow:
42.80 gpm Vibration:
Point A:
.172 ips Point B:
.101 ips Point C:
.087 ips ACCEPTABLE RANGE Flow:
40.70 gpm to 47.10 gpm Vibration:
Point A: s
.325 ips Point B: s
.252 ips Point C: s
.217 ips ALERT RANGE Low Flow:
39.80 gpm to 40.70 gpm Vibration: Point A:
.325 ips to
.700 ips Point B:
.252 ips to
.604 ips Point C:
.217 ips to
.521 ips REQUIRED ACTION RANGE Low Flow:
39.80 gpm High Flow:
47.08 gpm Vibration: Point A: >
.700 ips Point B: >
.604 ips Point C: >
.521 ips COMMENT:
e e
m..
.___.,_.m__..___.
_.._..._,..-__-....m.
m___.m.._.. -..,
...m._..,__._____.s-__
a A
4 i
TEST DATA POR ONE PL94P 9/06/96 Page 1 Pumpr IP2B Testt 021 Flow Test i
Vibrations lips)
Test Date Flow A
B C Int Remarks 10/29/92 44.0
.100
.120
,000 1.EH ROUTINE SURVEIIIJufC 3/01/93 44.0
.100
.106
.088 5/29/93 44.7
.127
.126
.110 9/03/93 44.5
.125
.126
.115 12/02/93
.116
.133
.125 3/05/94 44.5
.128
.123
.117 6/03/94 42.1
.198
.146
.091 9/07/94 42.6
.131
.119
.108 LEH ROUTINE SURVEILLANC 12/05/94 42.1
.252
.196
.087 1.EH ROUTINE, 2P-2A NO 9 2/24/95 41.3
.167
.119
.099 LEN 1P-2B, NO 9407493 2/24/95 41.3
.167
.119
.099 LEH 1P-38, NO 9407493
[
2/25/95 42.9
.128
.126
.100 12H 1P 2B, NO 9502540 r
3/05/95 43.0
.137
.119
.098 IJN ROUTINE SURVEILLANC t
6/02/95 4112
.112
.098 LRD ROUTINE SURVEILIANC 9/06/95 45.5
.177
.120
.105 LEH ROUTINE SURVEIILANC 12/05/95 45.5
.175
.115
.096 BAT ROUTINE SURVEIILANC 3/07/96 45.2
.141
.112
.10 9 12H ROUTINE StmVEILLANC 5/22/96 172
.101
.087 L3H POST MAINT. P28, NO 6/05/96 43.5
.109
.108
.105 1AD ROUTINE SURVEILLANC 9/05/96 42.6
.126
.110
.111 LEH ROUTINE SURVEILIANC 4
. _~
.. -.-.- ~__.
_ --. _.. _ _. - ~ _ _
o Pump Flow Reference
- FLOW TEST **
Values and Limite Date Established:
2/28/96 Pump #: 1P2C IT-021 Entered By: LEH REFERENCE VALUES Pressure: 2000.00 Flow:
44.14 gpm Vibration:
Point A:
.176 ips (7 f 0 f l o ye.
Point B:
.138 ips Point C:
.106 ips ACCEPTABLE RANGE Flow:
41.90 gpm to 48.60 gpm Vibration:
Point A.
s
.325 ips Point 'd: s
.325 ips Point C: s
.264 ips ALERT RANGE Low Flow:
41.10 gpm to 41.90 gpm High Flow:
48.60 gpm to 48.55 gpm Vibration: Point A:
.325 ips to
.700 ips Point B:
.325 ips to
.700 ips Point C:
.264 ips to
.635 ips REQUIRED ACTION RANGE Low Flow:
41.10 gpm High Flow:
48.55 gpm
.700 ips j
Vibration: Point A: >
Point B: >
.700 ips Point C: >
.635 ips COMMENT: Criteria from ASME OMb-1989, Part 6.
WO 9601306 set new reference values.
-.m TEST DATA FOR Oert PUMP 9/06/96 Page 1 Pumps 1P2C Test 021 Flow Test Vibrations (ipe) 1
-Test Date Flow A
B C
Int Remarks 10/29/92 44.9
.130
.090
.090 LEH ActTTINE SURVEILLANC 2/09/93 45.1
.127
.102
.083 3/01/93 45.1
.148
.131
.100 6/29/93 46.0
.116
.101
.004
^9/03/93 44.5
.178
.105
.125 12/02/93-
.128
.102
.086 3/05/94 45.9
.145
.109
.049 5/27/94 42.0
.122
.111
.097 9/07/94 42.7
.119
.097
.089 LEM ROUTINE SURVEILLANC 12/05/94 42.8
.116
.185
.271 IAH ROUTINE, 2P 2A NO 9 1/06/95 41.3
.125
.101
.089 LEM INQtEASED PREQUENCY 3/05/95 42.9
.112
.091
.087 IAH ROUTINE SURVIILIJWIC 6/02/95
.133
.101
.090 IAD ROUTINE SURVEILIJsse 8/04/95 46.3
.141
.098
.083 IJtD POST IRIN. POR 1P2C 9/06/95 46.0
,140
.102
.096 LEM ROUTINE SURVEILLAlec 12/05/95 45.5
.137
.106 4091 RAT ROUTINE SURVEILLA38C 2/02/96 44.3
.128
.094
.089 LEH POST MAINT, 1P2C 2/28/96 44.1
.176
.138
.106 LEH NO 9601306, IP-2C 3/07/96 45.2
.184
.133
.108 LEH ROUTINE SURVEILLANC 4/25/96 43.2
.189
.128
.105 LEH POST MAIIFT. IP2C.NO 5/24/96 43.5
.121
.124
.096 IJtD POST-MAIN. TEST FOR 6/05/96 43.7
.121
.131
.101 IJtD ROUTINE SURVEILIANC 9/05/96 43.6
.153
.122 098 LEH ROUTINE SURVEILLANC
d Pump Flow Reference
- FLOW TEST **
Values and Limits Date Established:
3/01/93 j
Pump #: 2P2A IT-022 Entered By: LEH REFERENCE VALUES Pressure:
Flow:
42.60 gpm Vibration:
Point A:
.112 ips i
Point B:
.138 ips l7f0116 rps Point C:
.115 ips ACCEPTABLE RANGE Flow:
40.47 gpm to 46.86 gpm Vibration:
Point A: s
.280 ips Point B: s
.325 ips Point C: s 4
.288 ips
\\
ALERT RANGE Low Flow:
39.62 gpm to 40.47 gpm High Flow:
46.86 gpm to 46.86 gpm Vibration: Point A:
.280 ips to
.672 ips Point B:
.325 ips to
.700 ips Point C:
.288 ips to
.690 ips REQUIRED ACTION RANGE Low Flow:
39.62 gpm High Flow:
46.86 gpm i
Vibration: Point A: >
.672 ips Point B: >
.700 ips Point C: >
.690 ips COMMENT: Criteria fem ASME OMb-1989, Part 6.
a 4
TEST DATA FOR ONE PUMP 9/09/96 Page 1
Pump: 2P2A
{
Tests 022 Flow Test Vibrations (ips) t Test Date Flow A
B C Int Remarks 11/21/92 ~
44.7
.110'
.14 0
.120 ROUTINE SURVEILIANC j
11/21/92 25.0
.110
.140
.320 SAT ROUTINE SURVEILLANC 4
1/21/93 40.7
.131
.136
.103 3/01/93 42.6
.112 4138
.115 1
6/02/93 42.3
.104
.126
.109 9/t; il 44.2
.126
.143
.107 12/02/93
.108
.138 107 2/18/94 46.1
,106
.140
.120 4
3/05/94 44.8
. 6k's
.137
.112
~
6/05/94 45.8
.098
.134
.111 7/21/94 45.2
.152
.144
.121 EEN ROUTINE SURVEILIANC 9/07/94 44.0
,1,03
.137
.108 LEN ROUTINE SURVEILIANC 11/02/94 46.2
.117
.133
.112 LEN 2P-2A, NO 9407468, 11/02/94 46.2
.117
.133
.112 LEM 2P-2A, MO 9407468, e
4 j
12/06/94 40.9
.111
.140
.110 12M ROUTINE SURVIIIRWC l
~#*0/95 43.0
.101
.126
.116 LEM 2P 2A NO 9501241 2,.F/91 43.1
.092
.127
.121 LEH 2P-2A, NO 9501906 3/05/95
'43.2
.104
.128
.117 LEH ROUTINE SURVEILI.ANC 4/28/95 42.8
.108
.126
.112 BAT POST MLINTENANCE 2P 6/19/95 43.5
.106
.137
.107 LRD ROUTINE,9506129,2P2 9/06/95 42.9
.105
.136
.109 LEN ROUTINE SURVEILIJJfC 9/21/95 44.6
,128
.120
.094 LEM NO 9508942, 2P-2A
]
12/04/95 42.9
.306 1.36
.110 SAT ROUTINE SURVEILLANC 1
3/07/96 43.5
.104
.342
.106 LEN ROUTINE SURVEILLANC 6/C3/96 44.5
.300
.136
.557 IAD ROUTINE SURVEILLANC 6/26/96 44.1
.145
.134
.118 LRD 2P2A, 2P2B INCREASE 9/08/96 43.6
.113
.133
.109 LEX ROUTINE SURVEILLANC j-4 4
...... ~..
l
?
P a
Pump Flow Reference s
- FLOW TEST **
Values and Limits
(
Date Established:
6/08/96 Pump #': 2P2B IT-022 Entered By: LEH-4 t
REFERENCE VALUES d
i I,
j Pressure: 1984.00 Flow:
43.30 gpm l
Vibration:
Point A:
.123 ips i
Point B:
.113 ips Point C:
.114 ips I
1 ACCEPTABLE RANGE 1
Flow:
41.10 gpm to 47.60 gpm l
i Vibra t.i.an :
Point A: s
.308 ips
[
l Point B: s
.283 ips l
j Point C: s
.285 ips j
ALERT RANGE i
j Lcw Flow:
40.30 gpm to 41.10 gpm i.
l Vibration: Point A:
.308 ips to
.700 ips i
Point B:
.283 ips to.
.678 ips j-Point C:
.285 ips to
.684 ips REQUIRED ACTION RANGE i
Low Flow:
40.30 gpm i
High Flow:
47.60 gpm j
Vibration: Point A: >
.700 ips i
Point B: >
.678 ips Point C: >
.684 ips
(
'j-i COMMENT: Criteria from ASME OMb-1989, Part 6.
Changed flow reference only for test dated 06/08/96.
4
)
l l
I l
~.__._......_.....,._-.._-..m____...m.
_m.._._._
.m.
_m.__..._
]
a l
TEST DATA POR ONE PtMP 9/09/96 Page 1
Pump 2P28 i
Test 022 Flow Test j
Vibrations lips)
I l
Test Date Flow A
B C Int Remarks 11/21/92 26.0
.140
.120
.140 BAT ROWINE SURVEILIJWC 11/21/92 45.2 140
.120
.140 BAT ROWINE SURVEILIJWC 1/21/93 40.7
.147
.119
.141-3/01/93 42.0
.166
.159
.152 6/02/93 42.3
.135
.119
.132 9/01/93 43.7
.150
.119
.131 12/02/93
.213
.189
.127 3/05/94
-44.8
.146
.113
.129
{
6/05/94 45.1
.152
.113
- 19 9/07/94 44.4
.151
.111
.132 LAN ROWINE SURVEILIANC 1
12/06/94 41.6
.148
.106
.114 LEH ROWINE SURVEIL 1JWC 3/05/95 43.2
.152
.198
.13 0 12M ROWINE SURVEILIJWC 6/19/95 43.2 150
.087
'142 LRD ROWINE,9506129,2 P2 9/06/95-44.1 165. 118
.160 LEN ROUTINE SURVEILLANC f
12/04/95 43.6
.157
.101
.141 RAT ROWINE SURVEILIJWC 3/07/96 44.1
.149
.104
,123 LEH ROWINE SURVIILLANC 3/27/96 46.6 123
.113
.114 LEN POST MAIFT 2P2B, NO l
6/08/96 43.3
.156
.168
.421 LRD ROWINE SURVEILIANC 6/26/96 44.7
.179
.112
.134 IJtD 2P2A, 2P2B INCRAASE 9/08/96 44.8
.153
.113
.125 LEH ROWINE SURVEILLANC I
j 7
9
__c.
. - - -. ~.. -.
i l
l l
Pump Flow Reference
- FLOW TEST **
Values and-Limits i
Date Established:
9/07/94 l
Pump #: 2P2C IT-022 Entered By: LEH
{
REFERENCE VALUES j
5 t
Pressure:
Flow:
42.00 gpm Vibration:
Point A:
.122 3.9 1
Point B:
.091 ipu l7f#'/8//4 j
Point C:
.097 ips j
3 ACCEPTABLE RANGE l
Flow:
39.90 gpm to 46.20 gpm Vibration:
Point A: s
.305 ips Point B: s
.228 ips Point C: s
.243 ips I,
l ALERT RANGE Low Flow:
39.06 gpm to 39.90 gpm High Flow:
46.20 gpm to 46.20 gpm i
Vibration: Point A:
.305 ips to
.700 ips I
Point B:
.228 ips to
.546 ips Point C:
.243 ips to
.582 ips REQUIRED ACTION RANGE Low Flow:
39.06 gpm High Flow:
46.20 gpm Vibration: Point A: >
.700 ips Point B: >
.546 ips Point C: >
.582 ips COMMENT: Criteria from ASME OMb-1989, Part 6.
1 i
1 l
l l
... - -. ~ ~ -. ~ -.
... -. -.. ~.. -. -. - - - -..
.. ~.
l t
TEST DATA FOR Cert PUMP 9/09/96 Page 1
{
Pump: 2P2C Test 022 l
' Flow Test l
l Vibrations (ips)
Test Date Flow A
B C Int Remarks i
11/21/92 44.7
,140
.110
.090 BAT Rot 7 TINE SURVEILIANC 11/21/92 25.5
.140
.110
.090 BAT ROUTINE SURVEILI.ANC 1/21/93 40.7
.130
.109
.093 3/01/93 41.6
.143
.102
.091 i
6/02/93 42.8
.170
.108
.105 I
i 9/01/93 44.7
.302
.112
.096 12/02/93
.192
.103
.000 l
I 1/27/94
.213
.108
.104 1/27/94 45.0
.213-
.108
.104 i
3/05/94 44.8
.204
.105
.108 i
6/05/94 42.5
.137
.102
.097 i
7/21/94 42.5
.167
.099
.156 ERN ROUTINE SURVEILIANC k
8/10/94 42.2
.159
.100
.099 EEN INCREASED FREQ. TES i
l 9/07/94 42.0
.122
.091
.097 LEH ROUTINE SURVEILLANC l
10/31/94 42.7
.171
.113
.101 LEN POET MMrf. 2P-2C, N 12/06/94 37.2
.183
.105
.101 !.EH RotTTINE SURVEI!4ANC I
12/29/94 43.9
.201
.115
.142 LEH 2P 2C Capacity test I
3/05/95 43.7
.146
.098
.109 LEH RotTTINE SURVEILIANC
}
6/13/95 43.7
.167
.151
.106 LRD ROUTINE,2P2C 950623 6/19/95 43.1
.189
.154
.104 LRD Rot 7 TINE.9506129,2P2 7/07/95 43.6
.158
.104
.117 IJtD R0tTTINE.2P 2C. 95072 9/06/95 43.0
.168
.151
.115 LEH Rotti!NE SURVEILLANC 12/04/95 43.8
.144
.142
.119 RAT ROUTINE SURVEILLANC l
3/07/96 44.0
.155
.143
.118 LEH ROUTINE SURVEILLANC 6/08/96 42.8
,148
.146
.154 IJtD RottrINE SURVE!LLANC 1
6/11/96 44.9
.134
.168
.112 IJtD POST MAINTENANCE FC j
9/08/96 43,1
.162
.153
.126 1.EH ROL? TINE SURVE!LIANC t
l 1
I
POINT BEACH NUCLEAN. PLANT BACKGROUND DOCUMENT I
INSERVICE TESTING PROGRAM APPENDIX F t
THIRD INTERVAL Revision 1 December 10, 1992 CHEMICAL AND VOLUME CONTROL (CVCS)
Safety Function:
The CVCS System serves as an attemate shutdown system by pmviding.for reactor coolt.nt system boration when required. (FSAR 9.2) l J
l Components:
1-P-002 A-C (684J741) 2-P-002 A-C (685J175)
Charging Pumps The Charging Pumps deliver concentrated boric acid solution at the rate required for RCS boration fmm the discharte of the RWST's or the Boric Acid Transfer Pumps to the RCS.
In addition, the charging pumps are used to mitigate the effects of a small break LOCA.
However, per the safety analysis, the safesy injection pumps are the primary means for thing to a small break LOCA but no such credit is taken in the safety analysis.
(FSAR 9.2 and 6.2.2)
Test Requirement: IWP-3000 1-P-004 A&B (684J741) 2-P-004 A&B (685J175)
Boric Acid Transfer Pumps The Boric Acid Transfer Pumps deliver concentrated boric acid solution at the rate required for R.CS boration from the Boric Acid Tanks to the suction of the Chargmg Pumps.
(FSAR 9.2)
Test Required: IWP-3000 l
1-CV-00112B (684J741) 2-CV-00112B (685J175)
RWST To Charging Pump Suction Control Valves These valves open to provide the primary (preferred) source of concentrated borated water for RCS boration.
i l
Test Requirement: BT-O PIT w
e Page 1 of 9 l
_. ~ _ _ _ _ _ _ _
- SEP-30-69ft0H21
- 48 WIS00NSlHEl.ECTRIOHPBU FAX N0. 4142212010 P.20 t
j
, e.....................................
l Printed Fo j
o...............r:
4 1
DIto Monday, 30 September 1996 6: 44nm CT Tot 1
From:
Subjects charging pumps Charging pump flow rate in resolution of Non-Conformance Report N 89 187 1
i Th3 original problem as stated in NCR N-89-187 is a statement on Pag of the FSAR that'one charging pump is capable of maintaining pressuriser e 14.3.1 1 pressure at 2250 psia with a 3/8 in, break in the RCS.
be:n replaced and the FSAR now says that the makeup flow rate from t That statement has i
for the operator to respond without activating the BCC wo 3/8 inch diameter hole.
This is a crua statement s i
a 90-015 assuming minimum charging pump performance. upported by calculation N-design basis is more clearly stated in the definition of the Reactor CoolBut the key i
Pre 2sure Boundary (RCPB).
It makes clear that there is no ant requirement for make-up flow rate from the charging pumps. specific flow rate 4
j Th2 original definition of RCPB in the QA Policy Manual in effect at the time of NCR-89-187 excluded 3 dicmeter and smaller were/4 inch piping from the RCPB.
Connections 3/8 inch requirements because a not considered part of the RCPB and had no QA within the capability of the normal reactor makeup water. system"... failure of
-of a 3/8 inch break by itself withoutth3 normal charging system needed to be cap i
Therefore
(
challenging the ECCS.
es j
part of the QA program as safety related seismic class 1 cospone Documentation shows that this recommendation has been implemented GUIDELINES FOR SYSTEM, COMPONENT AND PART CLASSIFICATION,DG-G06, j
of the QA Policy Manual, no longer excludes 3/8 inch piping from the RCPBthe current vers valvas normally closed during normal reactor operation in s It j
wo does not penetrate primary reactor containment".
i is now that CVCS is capable of make-up due to minor leakageGeneral criteria in DG-G06 brock.
Green line drawing NEST 541F091 shows that the 3/4 inch piping is now
, not a 3/8 inch QA ocope.
CHAMPS lists the components that were non-QA at the time on NCR N.89-187 (RC.500J, RC-500Q and RC-579) as QA scope, safety related, and seismic cla=a 1 components.
3/8 inch break no longer needs to be considered for design of the normalS makaup system.
1 Therafore, etnssquences of a 3/8 inch break no longer exists and the assumpti in calculation N-90-015 for chargin j
requirement for the charging pumps.g pump performance is no longer a design 1
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NONCONFORMANCE REPORT
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2 4
nu u<o4CES (Arfected system equipment. prx-ere. code, drawing..te.)
- NCR BASIS CATECORY O
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NUCLEAR POWER DEPARTMENT m=== w =,
NCR CONTINUATION SHEET N _.g --g; j
THIS PAGE IS A CON 2NUATION 18[ 2.1.2 - condition Description O 2.3.2/2.5.1 - Evoluotfon j
OF THE INDICATED SECTION:
O 2.2 - Reportablity/OperabBity O 2.7 - Corrective Action i
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.PAGE 7
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NUCLEAR POWER DEPARTMENT
= = = =,
l NCR CONTINUATION SHEET N - gq _ g.;t THIS PAGE IS A CONTINUATION O 2.1.2 - condition Descripuon [jZ) 2.3.2/2.5.1 - Ev oluouon i
OF THE !NDICATED SECTION:
O 2.2 - ReportabIlty/Operob!!!y O 2.7 - Corrective Action i
l y
Eraluation i
W 5
i The evaluation consists of three parts.
The first part checks a
i the calculations contained in the condition description section f
of the NCR Snd the numbers reported in the FSAR.
The second part i
g explores the history of the subject paragraph in the FSAR.
The o
third part evaluates the consequences of changing the FSAR.
O p
Part 1 - Calculations
(
E E
Calculations contained in the condition description section of the NCR are correct.
The maximum charging pump flow is approximately 8.3 lbm/s per pump.
In addition, calculations using Crane (1980)l, at normal operating pressure and temperature and with limiting assumptions about break geometry, show that i
break flow through a 3/8 in. diameter break of 17.5 lbm/s is
{
reasonable.
The obvious conclusion is that one charging pump i
cannot maintain pressurizer pressure and level indefinitely.
Maintaining pressurizer level does not necessarily mean that the pressurizer level remains unchanged with a small leak in the RCS.
An alternative interpretation of maintaining pressurizer level is
{
that the charging system is capable of maintaining pressurizer level long enough for the operators to identify a loss of inventory and isolate the leak or perform an orderly shutdown, cooldown and depressurization of the RCS.
Operators should have i
sufficient time to identify and respond to a 3/8 in, diameter break without relying on the ECCS.
1 i
ECCS is initiated when pressurizer pressure falls below 1735 e
psig, when stemn line pressure falls below 530 psig, i
containment or when i
pressure exceeds 5 psig.2 A 3/8 inch pipe break should have little impact on steam line or containment pressures.
{
The setpoint of concern is the pressurizer pressure.
If the i
charging pumps can maintain pressurizer level, then the pressurizer heaters can maintain pressure and the low pressure setpoint is not reached before an orderly cooldown can be initiated.
{
Estimates of thg time required to empty the pressurizer by removing 600 ft of water are shown, as a function of the number of charging pumps running, in the following table.
The table also shows the energy required to maintain normal operating 3
(
l PAGE 3 OF lO Forrn QP 15-1.2 Rev. o n
J
NUCLEAR POWER DEPARTMENT i
ammon w -,
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NCR CONTINUATION SHEET
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?
3 i
THIS PAGE IS A CONTINUATION O 2.1.2 - condition Descripuon 09 2.3.2/2.5.1 - Evoluouon l
OF THE INDICATED SECTION:
O 2.2 - RepxtobDity/OperabDity O 2.7 - Co' rrective Action e
l pressure by producing steam in the pressurizer.
The maximum output of the heaters is 1000 kilnwatts.
5 k
d W
ESTIMATED TIHE TO REMOVE 600 FT3 OF WATER FROM PRESSURIZER 4
i i
o Number of Volume Volume Rate of Time to Energy to i
)
o Pumps Flow Flow Change of Remove Maintain 1
Q Running Rate Out3 Rate In4 Volume 600ft35 Pressure 6 j
q (ft3/s)
(ft3/s)
(ft3/s)
(minutes)
(kwatts) c:
E Zero 0.4137 0.0
-0.4137 24.
1251.
i One 0.4137 0.1967
-0.2170 46.
656.
{,
Two 0.4137 0.3933 0.0204 490.
61.7 i
J Three 0.4137 0.5899 0.1763 1
i The most probable situation is that two charging pumps a e available when the break occurs.
Technical Spec!Cication section 15.3.2.B.1 rvquires that two charging r mos be available
]
when the reactor is taken critical.
A limiting condition of ope ration in section 7 3.3.2.D.1 requires that a second pump be availaule within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if only one pump is available at power.
i The ca.se with zero charging pumps available should never happen j
and the case with one charging pump running should happen infrequently.
j 1'
This calculation should remove any concern that a small diameter i
break is a catastrophic accident which is un-analyzed due to a i
misstatement in the FSAR.
In the most probable condition, the operators should have several hours to identify the break and take corrective action.
i However, the subject paragraph in the FSAR is still misleading.
implies that one charging pump is capable of maintaining It pressurizer level.
4 Corrective action should still be to remove or revise the subject paragraph.
1 i
l Part 2 - History i
l PAGE 4 OF 10
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NUCLEAR POWER DEPARTMENT i
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NCR CONTINUATION SHEET
\\ _ aq _ mg THIS PAGE IS A CONENUAT!ON O 2.1.2 - condition Descripuon E9 2.3.2/2.5.1 - Everuouon OF THE INDICATED SECTION:
O 2.2 - Reportobstty/operobnity O 2.7 - corrective Action j
)
y The subject paragraph was added with Amendment 23 to the FSAR i
EE and documented in a letter to the NRC dated October 14, i
i-1977.
The October 14, 4
1977 letter references an October 27, 1976 letter to NRC reporting a revised small break loss of coolant a4 analysis.
However, neither of these letters contains the basis g
for adding the subject paragraph to the FSAR.
Westinghouse usually provides replacement pages for the FSAR as part of a revised LOCA analysis.
The subject paragraph appears once in our correspondence with Westinghoune in a lettar from J.
- it S. Taylor, if, to G. A. Reed, WE, dated July 2, 1974.
.e July 2 E
letter enclosed a draft write-up of the ECCS analysis in a i
E generic format with specific plant dependent results left blank.
j Four blanks appear in the subject paragraph.
They are for normal j
pressure, sustained pressure, hole diameter, and creak flow rate.
In June of 1975, eleven months after receipt of the subject paragraph in generic format, WE submitted a re-evaluation of ECCS cooling performance and Amendment 16 to the FSAR.
Amendment 16 does not include the subject paragraph.
is not until more than three years after our receipt of the It subject paragraph in generic forinat that it is submitted, with blanks filled in, as part of Amendment 23 to the *.'SAR.
Nothing has been founo in our records to support the add 2 tion of the subject paragraph to section 14.3.1 of the FSAR.
Part 3 - Consequences If the subject FSAR paragraph is removed or revised, then the QA Policy Manual must be revised.
The QA Policy fianual defines the Reactor Coolant Pressure Boundary (RCPB) as the pressure-containing components such as pressure vessels, piping, pumps and valves and connections to the Reactor Coolant System (RCS) greater than 3/8 inches inside diameter.
Connections 3/8 inches or smaller are not considered part of the RCPB.
The basis for not including 3/8 inch or smaller connections in the RCPB is stated in Item 1, Section One of Appendix B to the Quality Assurance Policy Manual as follows:
"...not QA-scope because failure of these conrections results in a leak rate within the capability of the normal
{
reactor makeup water systems."
PAGE 5 OF 10 h,7o# "
i 4
i i
NUCLEAR POWER DEPARTMENT m= = =,
j NCR CONTINUATION SHEET
\\ -- p __re i
THIS PAGE IS A CONTINUATION O 2.1.2 - condition Descripuon (3 2.3.2/2.5.1 - Evoluouon j
OF THE INDICATED SECTION:
O 2.2 - ReportabDity/OperobBity O 2.7 - corrective Action 1
1 EE No reference is cited in Appendix B to support the statement, but
{
Mr. Heiden, of QAS, suspects that the basis for the statement is the subject paragraph from the FSAR.
If no other justification aQ can be found, then th's QA Policy Manual must be revised.
k There are two ways that the OA Policy Manual could be revised.
o The first way is to remove the exclusion of 3/8 inch or smaller i
connections from the RCPB.
One obvious result is that the green-line drawings of the RCS in the QA Policy Manual would need to be Q
revised and appropriate parts upgraded to QA-scope, Information 5:
from the FSAR and fron Westinghouse Systems Engineering personnel EE indicates that adding small diameter piping to the QA-scope equipment list may not be an impossible task.
Section 4 of the FSAR describes the RCS pressure boundary and the j
codes and standards used in the design and maintenance of the
{
boundary.
There is nothing that differentiates 3/8 in, piping j
from the remainder of the pressure boundary.
l Mr. Jim Schlonsky (412-374-4258, spelling uncertain) of Westinghouse is familiar with the subject statement in the FSAR.
However, he was quick to point out that the statement should only be applied to plants with higher capacity centrifugal charging pumps.
The statement is inappropriate for a plant with positive displacement charging pumps bec,use the pump is obviously j
incapable of maintaining RCS inventory.
He stated that j
W
.19 house would never have put a statement like that into our j
FSAR because it is not necassary.
The Point Beach plant wss apparently built prior to the requirem7nt to classify RCS pressure boundary piping as is done today.
Newer plants are required to show that a small diameter pipe break will not challenge the engineered safety features j
because the small diameter piping was not, or could not, be built to the same standards as the large diameter piping.
The second way to revise the QA Policy Manual is to clarify the i
statement J
that a break in a 3/8 inch diameter pipe results in a leak rate within the capability of the normal reactor makeup water system.
Clarification should state that the basis for excluding small diameter piping from the RCPB is the' time j
available for operator response to a break.
l i
t PAGE _(r_ OF 10 Form OP 15-1.2 i
Rev. O 1
l a
i i
1 NUCLEAR POWER DEPARTMENT
===w=,
l NCR CONTINUATION SHEET A
_ sq _. le i
THIS PAGE IS A CONTINUAT!C'N O 2.1.2 - condition osseriptJon O 2.3.2/2.5.1 - Evoluotfon j
OF THE INDICATED SECTION:
O 2.2 - Reportablity/OperobBity C 2.7 - Corrective Actic7 t
l 0
Suggested text for the clarification of Item 1 in the list of E
Systems and Equipment Covered by the Quality Assurance Program is as follows:
j
=.
j d
1.
The Reactor Coolant Pressure Boundary (RCFB) as defined 5
a l
g above.
(Connections to the reactor coolant system (RCS) j greater than 3/8 inch inside diameter are considered part of the RCPB, including branch outlet nozzles or nipples, instrument wells, reservoirs, studs and fasteners in flange T
l joints between pressure parts, traps, strainers, and
{
W orifices.
Connections 3/8 inch inside diameter or smaller l
i E
3 are not considered part of the RCPB and are therefore not QA-scope.
Failure of a connection 3/8 inch diameter or i
smaller results in a loss of RCS inventory to which the j
operator can respond without activating the ECCS.
In the most likely condition, with at least two charging pumps l
available, the operator has several hours to identify and respond to the break.)
The OA Policy Manual should reference a formal calculation of the
^
l time available for operator response under a variety of conditiera as was estimated in part 1 of the evaluation.
-I Corrective Action The recoramended corrective action includes four activities as tabulated below:
i Date D;f GrovD Activity 6[4W/ 1?/1/89 NSEAS Perform a formal calculation of the time available for operrtor response to a tieak in a 3/8 inch inside diameter or smaller hole in l
the RCS.
1 c4 52 7/3. '90 NPERS Revise PSAR section 14.3.1 paragraph 2 as i
follows:
1 1
}
The maximum break size for which the normal j
makeup system can maintain the pressurizer level is obtained by comparing the calculated
]
flow from the reactor coolant system through 4
PAGE ? _OF JD p,7o@ M 2
]
i
-lc
5 4
l NUCLEAR POWER DEPARTMENT
=aaon w m i
NCR CONTINUATION SHEET
\\
_ gq _ J R7 THIS PAGE IS A CONTINUATION O 2.1.2 - condltron oe cription O 2.3.2/2.5.1 - Evaluation j
OF THE INDICATED SECTION:
O 2,2 - Reportabt!!y/OperabElty O 2.7 - corrective Action C
the postulated break against the charging EE pump makeup flow at normal reactor coolant 3
j system pressure, i.e.,2250 psia.
A makeup I
a i
flow rate from two ene-charging pumps is typically adequate to-sustain pressarieer j
g l
pressure-sh-335&-psia-to maintain j
l pressurizer level long enough (i.e several l
hours) for the operator to respond without j
l i
'etivating the ECCS for a break through a 3/8 M j,,sk/c 4e.
iameter hole. --This-break-results-in-a j
W a -e f -approxiinately -14 r5 -lb/see-E P4
)
i E
12/1/89 QAS As an interim corrective action, the basis p$3 for excluding small diameter piping from the 4
RCPB in the QA Policy Manual should be changed as follows:
y
]
Connections 3/8 inch inside diameter or j
smaller are not considered part of the RCPB j
1 and are therefore not QA-scope. because-f i
i l
railure of these connections results in a leak rate within-the-capability-of-the-normal l
reactor-makeup-water-systems to which the l
operator can respond without activating the l
ECCS.
In the most likely condition, with at I
least two charging pumps available, the I
operator has several hgyrs to identify and I
respond to the break.I J Where reference [1] is the calculation created as the first corrective action.
12/1/89 QAS TN Investigate the effort required to remove the exclusion of small diameter connections to the RCS from the definition of RCPB in the QA Policy Manual, add the small diameter connections to the QA-Scope green-line drawings, and upgrade appropriate equipment to QA scope.
Determine final corrective action for QA Policy Manual based on results of investigation.
PAGE._2_OF 1O g,7o# "
m
j NUCLEAR POWER DEPARTMENT m== -,
NCR CONTINUATION SHEET A
-- gi __ i n 2
THIS PAGE IS A CONTINUATION O 2.1.2 - condition Oescripuon O 2.3.2/2.5.1 - Evoluouon i
OF THE INDICATED SEC110N:
O 2.2 - Reportabnity/Operabatty O 2.7 - Corrective Action
!O i
E zwonoTzs 5
1.
Crane, Flow of Fluids, Technical Paper No. 410, 1980 a
y 2.
Point Beach Nuclear Plant, Setooint Document, STPT 2.1, g
.r MAJOR, Revision 1 06-14-89 l
c i
3.
Vout - Mout *v 4
p where Vout = Volume flow rate out of RCS E
v = 0.02364 (f t3/lbm) - specific volume of saturated water at 600 degf.
Mout = 17.5 (1bm/s) - break mass flow rate from FSAR i'
4.
Vin = N
- G
- C1
- C2
- v / v' I
where vin - Volume flow rate into the RCS N = number of operating charging pumps G = 60.5(gpm) - the volume flow rate per pump C1 = 0.13368 (ft3/ gal) - conversion factor C2 - 1(min)/60(sec) - conversion factor v-
.02365(ft3/lbm) - specific volume of sat arated water at 600 degf v' = 0.016204 (f t3/lbm) - specific volume of saturated water at 120 degf.
5.
T = V
- C2 / [Vout - Vin) where T = Time to Remove 600ft3 V = 600(ft3) liquid volume of pressurizer from FSAR 6.
E=
(Vin - Vout)
- hfg
- C3 / vfg where E = energy required to maintain pressure (kwatts) l PAGE 'I OF lO h,7o# "
4 4
NUCLEAR POWER DEPARTMENT amc= = w =,
NCR CONTINUATION SHEET
\\ _a_a THIS PAGE IS A CONllNUATION O 2.1.2 - condition Description S 2.3.2/2.5.1 - EvoluotJon OF THE INDICATED SECTION:
O 2.2 - Reportabilty/OperobDity O 2.7 - corrective Action
{05 hfg = 466.2 (btu /lbm) - heat of vaporization at 5
2000 psia h
C3 - 1.0548 (watts / btu /s) - conversion factor k
vfg = 0.16266 (ft3/lbm) - change in specific o
volume for vaporization f
YY//f d
c::
5
}
l i
i
)
I.
I i
,i PAGE IO OF.10 p.70w u-2 P
Calculation #
f CALCULATION REVIEW AND. APPROVAL NUCLEAR POWER DEPARTMENT N'
~ O/8 Number of pages
/r Title of Calculation:
hJw*r &t /~a2 Y/MCH C /M b Cx32 Nj'$fM"Y
@ Original calculation O Revised calculation.
Revision #
D superseding calculation.
supersedes calculation e Modification 8
==
Description:==
1 Other
References:
g g ggt g. ff./fjL i
i i
Prepared By:
4 Date h:RH ESs C i
i This calculation has been reviewed in accordance with QP 3-6.
l The review was accomplished by one or a combination of the following (as checked):
i A review of a representative A detailed review of the i
sample of repetitive calculca-original calculation 1
tions A review of the calculation A review by an alternate, against a similar calculation simplified or approximate previously performed method of calculation
- l Com:nents :
i i
1 I
t i
i
)
Resiewed By:
Date:
Annroved By:
Date:
W YN
)
90
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= (d.3/,fk' cok Pe c
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i \\ % 7st 0li c ui s..o (o*snH ers R:n.1 - Row bre 77wu. 4 4 " /r6i e $* $R Cf f I $xmVu', -/Mccrh*2r.r.rsot?-ricA ?, -C "fmoorts ?ry '#O)' ' a.) e o 32s a / f frht? 'EF ?. w x> 9 Is K = k r' u. a,. c.- " N r' eir- / \\ 'dde Meeuae z
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- b. DAe en- ~
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p \\ ' [ 720)15 l 4 t..* 1 i I l! 1 i G..ai.<eg_g.. Gsin urs ~ ~ ' ~~ '~ r . Pse-2 - A > w B rr Meouw % A r i E. Owt.5.Sc rw, acaw:rtme now, mm:c w = o. n r /
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- i e=04 fa,,G +o 2 ~ t, 4.n
-setter now carrierm' w tvsnm+c w w = o.wcu)'(o.+) ((2 coo-os)A-o nn >)"' = tr. 3 'o~4 / l' E U *.! M '? e*'e//, {OV4*.In / X ? OM j C2/&Cf M = 6 S.?.S~ l' o!' C ( oPA7, ) Y" l 2e r I, Eon 3-22 w ~? u Y ~ O Q?27/ lY/ der i 'r'l&/J.f]'0, 'W> *D& ) = 0J'] 12cv I, P A-2I 3 r =~ Ia/p, (NJ,a2c)= 6SYS c sP= P, - re di = 20M(/ s>5)= f/o e- (o.orXan)(%) '(o yX"/o.on nY' c a.se /~4 / i 4 NCR # N - S1 -I 9'] n ne 33 aimur,wsw.m
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- YN "kY N
N $ $ {J t,/ / Act Yggy 2 warer At - n roec ev /% n 6-rurws 2. C, c 40.5 SPr/ Z.Z C, = 3 0 d?f I F-mr (?:&rf-/2P*) = 0 0/4 23Y UNt,,, Per C f, - d. t.r.rSe D bi. he& Ca = /nw /to ces. Via CN=/) = 0. / 70 B'/.c / Vw CM 2) c 0 3.ro OVs V,a (x 1) = 0.54'? M % / hCR # N -e9 -t 77 PAGE to 0F 33
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- f 34/o~fh(2OOO) = /3 '/.S
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ll.. - dunlif3 E' ngii,,ec ~~q T,'$' dp. ~ g m P SA/P i NUCLEAR POWER DEPARTMENT m = on vo =, NCR CONTINUATION SHEET \\_p_jg'f l THIS PAGE IS A CONTINUATON O 2.1.2 - Condition DercripUon O 2.3.2/2.5.1 - Evoluodon OF THE 1NDICATED SECTION: O 2.2 - Reportobttty/Operablity CiQ 2.7 - Corrective Action i m l - FSAll Sedh-l'/. 5I fuyey4c2 i' 'A ""J'W l g g gg,,a .f4 < ,1 o,1 cou S~ < c e in a cc ordec e i y cou< dive AclIn N2. bp3 3 ~ g-I ,{ w cla gd FSB pp /WZl-/ / T3 = )s affsf. ?L CM &&"0he rssach ~6J L y '/'0 bb5 7,/ pa S 7/G//o 0 PAGE _2.7 OF _13 $""ooP fp1.2
- - - - - - - - - - - ~ - - ~ ' ' " " - ~ ~ - ~ ~ ~ ~ ~ ~ ~ ~^ i ~ ,7* j 14.3 PRIMARY SYSTEM PIPE RUPTURES j 14.3.1 Loss Of Reactor Coolant From Small Ruptured Pipes Or from j Cracks In large Pipes Which Actuates Emergency Core Cooling ) System _1dentification of Causes and Accident Description i 1 l A loss of coolant accident is defined as a rupture of the reactor coolant k system piping or of any line connected to the system up to the first f closed valve. Ruptures of small cross section will cause loss of the j coolant at a rate which can be accommodated by the charging pumps which 4 would maintain an operational water level in the pressurizer permitting f the operator to execute an orderly shutdown. A moderate quantity of l coolant containing such radioactive impurities as would normally be present in the coolant, would be released to the containment. j The maximum break size for which the normal sakeup system can maintain j the pressurizer level is obtained by comparing the calculated flow from l the reactor coolant system through the p. tulated break against the j charging pump makeup flow at normal reactor coolant system pressure, i.e., 2250 psia. A makeup flow rate from two charging pumps is typically 4 adequate to maintain pressurizer level long enough for the operator to j respond without activating the ECCS for a break through a 3/8 inch dia-j meter hole. Should a larger break occur, depressurization of the reactor coolant system causes fluid to flow to the reactor coolant system from the pres-surizer resulting in a pressure and level decrease in the pressurizer. Reactor trip occurs when the pressurizer low pressure trip setpoint is reached. The consequerces of the accident are Ifmited in two ways: 1. Reactor trip and borated water injection cc.Nlemer*, void formation in causing rapid reduction of nuclear powe-residual level cor-responding to the delayed fission and Vduct decay. 2. Injection of borated water ensures sufft .c flooding of the core to prevent excessive cladding tcaperatures. i i Revision 3 14.3.1-1 June 1990 NCR # N "GE z.s OF'O
4 i .rm l NUCLEAR POWER DEPARTMENT
- njaca, 1
i NONCONFORMANCE REPORT N - 29 - 13 7 THIS PAGE'IS A CONTINUATION [ z.1.2 - condition cescription 2.5.1 - E..iuation 0F THE INDICATED SECTION: r.2 - Reportablitty/ operability 32.7-correctiveAction 1 (Initial and Date all Entries) I ) Recommended corrective actions #3 and #4 on page 8 of NCR N l 187 were assigned to SQA. Upon investigation, it was determined I that the proposed revision to the QA Policy Manual given in j recommended corrective action #3 Vas inappropriate. The issue was discussed with the MSS (MSSM 90-06) and with the NCR evaluator (see attached memo) and a modified revision to the QA Policy Manual was j agreed 1 upon. This revision is contained in QA Policy Manual i Appendix B Rev. 3, dated 5/25/90. The issuance of this revision i on 7/2/90 completer, recommended corrective action #3. As discussed i in the attached memo, an investigation of small diameter connections to the RCS indicates that all existing connections meet j the requirements of QA Policy Manual Appendix B Rev. 3. This i completes recommended corrective action #4. l 7/2/9o l i 7Nk /% RBll 1 i l l 1 2.6.5/2.8!CAREY!EW Form O' 15 1.2 PAGE 2 1 0F O
- - - - ' ^ ^^^ ^^^ " ~ ~ ' ~ 3 j i j WE Internal Correspondence 4 } TO: j FROM: j DATE: APRIL 28, 1990 4 l SUBICCT: NCR N-89-187 I - =.. -. =. - - -... - -... - -.......... - - - _ -. =.. = - - - =... = - =. NCR N-89-187 addresses a discrepancy in the FSAR LOCA analysis i regarding the ability of the CVCS charging system to maintain RCS inventory following a break through a 3/8" diameter hole. 1 i In your evaluation of this NCR (attachment A) you recommended that QAS { perform the following actions: i
- 1) Revise the QA Policy Manual in the interim to give a more l
accurate basis for excluding small diameter piping from being QA-l scope. 2) Investigate the effort required to remove the small j diameter pipinc exclusion from the Policy Manual. I In the course of completing these corrective actions, I have come to the following conclusions: 1
- 1) I do not agree with the proposed Policy Manual wording that
" connections 3/8" ID or smaller are not considered part of the RCPB..." This is not consistent with the 10CFR50'.2 definition of reactor coolant pressure boundary.
- However, I
believe j 10CFR50. 55a (c) (2) (attachment B) provides a legitimate regulatory basis for excluding certain RCPB components from code requirements, and, therefore, from being QA-scope.
- 2) The only instances I identified in which the 3/8" exclusion was used were associated with impulse tubing to the LT-447 and LT-447A reactor vessel water level transmitters. Specifically, tubing and valves beyond the following root isolation valves was installed non-QA:
RC-500J, RC-500Q, and RC-579. Note that all of these valves are normally closed during reactor operation. (Various instrument impulse liner. shown as non-QA on the green line diagrams determined to have been installed and maintained QA-scope. were Corrections will be made to the affected green line diagrams.) I presented the above conclusions to the MSS in March, and based on the response I received (attachment C), I propose to revise the wording in the QA Policy Manual as shown on attachment D. Please i advise me whether you feel this is acceptable. If so, I will j consider QAS' portion of the corrective action for NCR N-89-187 to j be completed upon issuance of the revised QA Policy Manual Appendix B. i k $f NCR e N > 27 PAGE 5D OF 13
i i NUCLEAR POWER DEPARTHENT QA POLICY MANUAL i Wisconsin i \\ Electnc 1 POWER COMPANY 231 W Mctwoort PO Bam 2046. Mih oukee. w; S3201 i (414)221 2345 f l APPENDIX 8 I lRev: 3 l l 1 7/2/90;1 i SYSTEMS AND EQUIPMENT COVERED BY i i THE QUALITY ASSURANCE PROGRAM IDate: l i i [ lPages: 6 l g l Prepared by... f/7/Ta Approved by:... ,hSN4 a l This appendix provides the general criteria for the determination of QA-4 scope hardware at Point Beach Nuclear Plant (PBNP). The PBNP Q-List, ] contained in the CHAMPS equipment data base, identifies all systems, i structures, and components which fall under the scope of these criteria. i Items identified by the Q-List as being QA-scope are assigned QA Codes in CHAMPS. These QA Code numbers correspond to the applicable QA criterion ? i numbers in this appendix. Used in conjunction with the Part III color-coded l diagrams and other Part II appendices, this appendix provides the background used in determining the QA scoping of equipment listed in CHAMPS, as well as a reference for determining QA applicability for new equipment. g l Refer to the CHAMPS equipment data base, or the hard-copy CHAMPS "Q-List" printouts, to determine if particular systems, structures or components j are considered QA-scope. t 4 1 i i h NCR # N 197 PAGE SI 0733 81 usaranson,ar omnenum [gnme
a j m l ~' i i _ SYSTEMS AND EQUIPMENT COVERED BY i THE QUALITY ASSURANCE PROGRAM 1 APPENDIX B ) DEFINITIONS ~ l Safetv-Related - Safety-related structures, systems, and conponents are design basis events to ensure:those that are relied upon to remain functional during th i i 1. The integrity of the reactor conlant boundary, 1 2. The capability to shut down the reactor and maintain it in a safe shutdown condition, and, } 3. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines of 10 CFR Part 100. I { has made a regulatory or design basis commitment; cc, for pl reasons, Wisconsin Electric has ipplemented special controls to assure 1 reliability. QA-Scope - All safety related (SR) or augmented quality (AQ) items are said to be within "QA Scope" and are controlled under the QA program described in section I of this manual (or as modified in other Appendices of this manual). Reactor Coolant Pressure Boundary (RCPB) - Reactor Coolant Pressure i l Boundary (per ICJR50.2) means all those pressure-containing components such as pressure vessels, piping, pumps, and valves, which i are: 1. Part of the reactor coolant system, or i 2. Connected to the reactor coolant system, up to and including i the following: \\ The outermost containment isolation valve in system piping a. which penetrates primary reactor containment; b. The second of two valves normally closed during nors.a1 reactor operation in system piping which does not penetrato primary reactor containment; f The reactor coolant system safety and relief valves. c. i ) NCRgy.gg.7g., } PAGE 3L B* OF 3 h
_. _ _ ~ _ _ _ _ _ _ _. _ _ _. _ _ _. _ _ _ _ _ _ _ i 4 OA-SCOPE ITEM DETERMINATION CRITERIA i The following criteria define those items recuired to be considered " l QA scope". Criteria that define " safety-related" are identified with a "Y" in the left hand column under "SR?". i in this column. Augmented Quality criteria have.an "N" { QA-scope criteria that could apply to either safety-related or augmented quality '. ve an asterisk in the column. l listed in any particular order). (The criteria are not i Y 1. The reactor coolant pressure boundary (RCP8) as defined above. t j (NOTE: 10CFR50.55a(c)(2) exempts certain components ~ connected to the reactor coolant system from code requirements if, following a postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner. Consequently, piping and components of 3/8 inch outer diameter and smaller, beyond a first off isolation valve normally closed during reactor operation, may be exempted from being QA-scope). Y 2. The reactor core and reactor vessel internals (including fuel and fuel assemblies). Y 3. Those items required to function in order to provide overpressure l protection for the reactor coolant system during reactor opera-tion, as required by various safety analyses (i.e., pressurizer safety valves). Y 4. Systems or portions of systems necessary to provide emergency core cooling when required to mitigate the' consequences of an accident. Y 5. Portions of the main steam system required to remain intact and func-tional following a steam generator tube rupture or main steam line break in order to (1) isolate a ruptured steam generator, (2) provide ) redundant protection against blowdown of more than a single steam i generator, or (3) allow continued core residual heat removal using the unaffected steam generator. 6. Systems or portions of systems which provide cooling water for other l QA-scope equipment and components that are required for (1) emergency core cooling, (2) core residual heat removal, (3) post-accjdent I containment heat removal, or (4) spent fuel pool cooling. Y 7. The emergency diesel generators and systems or portions of systems l necessary to support the operation of the emergency diesel generators (fuel oil systems, air starting systems, service water, diesel room ventilation, etc.) NCRfg_pp;97 e,, NAGF 330F 33 -}}