ML17279A167
| ML17279A167 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/31/1987 |
| From: | Collingham R, Krajicek J, Morgan J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17279A161 | List: |
| References | |
| XN-NF-87-24, NUDOCS 8704030086 | |
| Download: ML17279A167 (165) | |
Text
8704030086 6,8703P7 ADOCN'<C'550003q7 PDR ADVANCEDNUCL.EARFUELS CORPORATION ZN-NF-87-24 Issue Date:
3/26/87 WNP-2 CYCLE 3 PLANT TRANSIENT ANALYSIS Prepared By:
E.
ajicek, Sr. Engineer BWR Sa ety Analysis z3'O'H Date Concur:
R.
E.
- lingham, Manager BWR S
ety Analysis Date Concur:
N. Morgan, Ma ager Customer Services Engineering Date Approve:
G. N. Ward, Manager Reload Licensing
+gZ4/d 7 Date Approve:
H.
. Williamson, Manager Licensing and Safety Engineering Date Approve:
G. L. Ritter, Manager Fuel Engineering and Technical Services z/>>/ig Date llh AIIAFFILIATEOF KRAFTWERKUNION Q KLVV.
NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Advanced Nuclear Fuels Corporation. It Is being submitted by Ad-vanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Advanced Nuclear Fuels Corporation-fabricated reload fuel or other technical services provided by Ad-vanced Nuclear Fuels Corporation for lightwater power reactors and it is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, informa-tion, and belief. The information contained herein may be used by the U.S.
Nuclear Regulatory Commission in its review of this report, and under the terms of the respective agreements.
by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Advanced Nuclear Fuels Cor-poration in their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.
Advanced Nuclear Fuels Corporation's warranties and representations concem-Ing the subject matter of this document are those set forth in the agreement bet-ween Advanced Nuclear Fuels Corporation and the customer to which this docu-ment is issued. Accordingly, except as otherwise expressly provided ln such agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf:
A. Makes any warranty, or representation, express or Im-plied, with respect to the accuracy, completeness, or use-fulness of the information contained in this document, or that the use of any information, apparatus, method, or pro.
cess disclosed in this document willnot infringe privately owned rights, or B Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, ap-paratus, method. or process disclosed in this document.
XN NF F00 766 (1/87)
ZN-NF-87-24 TABLE OF CONTENTS
~Sectio
1.0 INTRODUCTION
2.0
SUMMARY
3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1-Design Basis.
3.2 Anticipated Transients.
- 3. 2. 1 Load Re) ection Without Bypass.
3.2.2 Feedwater Controller Failure.....
3.2.3 Loss Of Feedwater Heating.
3.3 Calculational Model...........
3.4 Safety Limit..
4.0 MAXIMUMOVERPRESSURIZATION......
4.1 Design Bases...
4.2 Pressurization Transients 4.3 Closure Of All Main Steam Iso1ation Valves.
5.0 RECIRCULATION FLOW RUN-UP
6.0 REFERENCES
APPENDIX A.
~Pa e
5 9
9 31 31 31 32 33 36 A-1
M
ZN-NF-87-24 LIS OF TABLES Table 2.1 Thermal Margin Summary For Cycle 3...........
~...............
3.1 Design Reactor And Plant Conditions For WNP-2.
3.2 Significant Parameter Values Used In Analysis For WNP-2...'...
3.3 Results Of System Plant Transient Analyses 5.1 Reduced Flow MCPR Operating Limit For WNP-2
~Pa e
10
~
~
~
~
~
1 1 14 34 LIST OF FIGURES
~Fi use
~Pa e
3.1 3.2 3.3 3.4 3.5 3.6 Load Rejection Without Bypass Results, Normal Scram Speed Load Rejection Without Bypass Results, Normal Scram Speed..
Load Rejection Without Bypass Results, Normal Scram Speed..
Load Rejection Without Bypass Results, Normal Scram Speed..
Load Rejection Without Bypass Results, Tech.
Spec.
Scram Speed.
Load Rejection Without Bypass Results, Tech.
Spec.
Scram Speed.
RPT Operable, RPT Operable, RPT Inoperable, RPT Inoperable, RPT Operable, RPT Operable, 15 16 17 18 19 20 3.7 Load Rejection Without Bypass Results, RPT Inoperable,
- Tech, Spec:
Scram Speed.
21 3.8 Load Rejection Without Bypass Results, Tech.
Spec.
Scram Speed.
RPT Inoperable, 22 3.9 Load Rejection Without Bypass Results, MWD/MTU Exposure, RPT Inoperable, Tech 3.10 Load Rejection Without Bypass Results, MWD/MTU Exposure, RPT Inoperable, Tech End-Of-Cycle Minus 2000 Spec.
Scram Speed..........
End-Of-Cycle Minus 2000 Spec.
Scram Speed..........
23 24
iii-XN-NF-87-24 L ST OF FIGURES (Continued)
~Pi are 3..11 Feedwater Controller Failure Results For 47% Power And 106% Flow With Normal Scram Speed...........
~...........
3.12 Feedwater Controller Failure Results For 47% Power And 106% Flow With Normal Scram Speed.
3.13 Feedwater Controller Failure Results, RPT Inoperable, Normal Scram Speed A.l A.2 A.3 3.14 Feedwater Controller Failure Results, RPT Inoperable, Normal Scram Speed 3.15 Feedwater Controller Failure Results For 47% Power And 106% Flow With Tech.
Spec.
Scram Speed.
3.16 Feedwater Controller Failure Results For 47% Power And 106% Flow With Tech.
Spec.
Scram Speed.
5.1 Reduced Flow MCPR Operating Limit WNP-2 Cycle 3 Safety Limit Local Peaking Factors (ANF Fuel).....
WNP-2 Cycle 3 Safety Limit Local Peaking Factors (G.
E. Fuel)...
Radial Power Histogram For 1/4 Core Safety Limit Model..........
~Pa e
25 26 27 28 29 30 35 A-5
... A-6 A-7
XN-NF-87-24
- 1. 0.
INTRODUCTION This report presents the results of the Advanced Nuclear Fuels Corporation (ANF) evaluation, of system transient events for the Supply System Nuclear Project Number 2
(WNP-2) during Cycle 3 operation.
For this analysis the Cycle 3
core was assumed to contain 276 ANF 8x8 and 488 GE P8x8R fuel assemblies.
This evaluation together with core transient events(
)
determines the necessary thermal margin (MCPR limits) to protect against the occurrence of boiling transition during the most limiting anticipated transient.
The evaluation also demonstrates the vessel integrity for the most limiting pressurization event.
This evaluation is applicable to core flows up to the maximum attainable with the recirculation flow control valve in its fully open position which is 106 percent of the rated core flow value at 100$ power.
The methodology for these system transient analyses is detailed in References 2
and 3.
XN-NF-87-24 2.0 SU~KEY The Minimum Critical Power Ratios (MCPR) for potentially limiting plant system transient events at increased core floW are shown in Table 2.1 for powers that bound allowable values (4? to 104% power) at increased core flow.
The system transient MCPR values of Table 2.1 for the load rejectio'n without bypass (LRWB) and feedwater controller failure (FWCF) transients were obtained using a
scram time based on WNP-2 measured values.
The loss of feedwater heating (LOFH) transient results shown in Table 2.1 were obtained from a
generic analysis which is discussed in Section 3.2.3.
The limiting MCPR values for the cases of Table 2.1 are 1.32 for NSSS vendor and 1.30 for ANF fuels.
Also, analyses were performed for LRWB and FWCF events at a cycle exposure of EOC -2000 MWD/MTU when a large number of control blades are still inserted in the core.
These analyses showed that system transients were insignificant relative to the CRWE event (Reference 1).
Thus, plant operating limits can be based on CRWE event for cycle exposures up to EOC
-2000 MWD/MTU.
For exposures beyond EOC -2000 MWD/MTU the limits in Table 2.1 are applicable.
Additional transient analyses were performed
- assuming, the recirculation pump trip (RPT) out of service and assuming technical 'specification scram speed (TSSS).
The delta CPR results for these events are presented in Section 3.
Maximum system pressure was calculated for the containment isolation event, which is a
rapid closure of all main steam isolation valves, using the scenario as specified by the ASME Pressure Uessel Code.
This analysis shows that for WNP-2 Cycle 3 operation the safety valves have sufficient capacity and performance to prevent the pressure from reaching the established
- The Cycle 2
transient events analyzed at the design basis power condition with increased core flow were found to bound the same transients analyzed at the design basis power and flow condition for WNP-2 Cycle 2.
These results are shown in Reference 4.
XN-NF-87-24 transient pressure safety limit of 110% of design pressure.
The maximum system pressures predicted during the event are below the ASME limit of 1375 psig (110% of design pressure) and are shown in Table 2.1.
The analysis conservatively assumed six safety relief valves out of service.
The applicability of the Cycle 2
MCPR safety limit of 1.06 for all fuel types in Cycle 3 was determined using the methodology of Reference 5.
XN-NF-87-24 TABLE 2.1 THERMAL MARGIN
SUMMARY
FOR CYCLE 3 Transient Load Rejection~
Without Bypass 104/106 Delta CPR MCPR*
GE Fuel ANF Fuel 0.25/1.31 0.23/1.29 Feedwater Controller~
Failure 47/106 0.26/1.32 0.24/1.30 Loss of Feedwater***
Heating Not Applicable 0.09/1.15 0.09/1.15 MAXIMUMVESSEL PRESSURE (PSIG)
Transient Vessel Dome Vessel Lower Plenum Steam Line MSIV Closure 1285 1313 1287
+MCPR value using the 1.06 safety limit justified herein.
- These transients were evaluated with normal scram speed.
- Generic analysis bounding value, Reference 9.
~
J
XN-NF-87-24 3.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN 3.1 Desi n Basis System transient analyses to determine the most limiting type of thermal margin transient were performed at the increased core flow condition of 106%.
As shown in Reference 4,
system transients from the increased core flow condition bound thermal margin analyses transients from the nominal (100%)
flow condition.'nalysis of load rejection without bypass (LRWB) was performed at the rated design 104%
power/106%
flow point.
Since feedwater controller failure (FWCF) transients are more severe at reduced power because of the larger change in feedwater flow, the FWCF transient was performed at the minimum power (47%)
allowed for increased core flow.
The initial conditions used in the analysis for transients at the 104%
power/106%
flow point are as shown in Table 3.1.
The most limiting exposure in cycle was determined to be at end of full power capability when control rods are fully withdrawn from the core; the thermal margin limit established for end of full power conditions is conservative in relation to cases where control rods are partially inserted.
The calculational models used to determine thermal margin include the ANF plant transient and core thermal-hydraulic codes as described in previous documentation( >>
> ).
Fuel pellet-to-clad gap conductances used in the analyses are based on calculations with RODEX2( ).
Recirculation pump trip (RPT) coastdown was input based on measured WNP-2 startup test data, and the COTRANSA syst: em transient model for WNP-2 was benchmarked to appropriate WNP-2 startup test data.
The hot channel performance is evaluated with XCOBRA-T(
using COTRANSA supplied boundary conditions.
Table 3.2 summarizes the values used for important parameters in the analysis.
XN-NF-87-24 3 '
Ant ci ated Transients ANF considers eight categories of potential system transient occurrences for Jet Pump BWRs in XN-NF-79-71( ).
The three most limiting transients are described here in detail to show the thermal margin for Cycle 3 of WNP-2.
These transients are:
Load Rejection Without Bypass (LRWB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LOFH)
A summary of the transient analyses is shown in Table 3.3.
Other plant transient events are inherently nonlimiting or clearly bounded by one of the above events.
3.2.1 Load Re ection W thout B
ass This prevent is the most limiting of the class of transients characterized by rapid vessel pressurization.
The generator load rejection causes a turbine control valve trip, which initiates a reactor scram and a recirculation pump trip (RPT).
The compression wave produced by the fast turbine control valve closure travels through the steam lines into the vessel and pressurizes the reactor vessel and core.
Bypass flow to 'the condenser, which would mitigate the pressurization
- effect, is conservatively not allowed.
The excursion of core power due to void collapse is primarily terminated by reactor scram and void growth due to RPT.
Figures 3.1 through 3.10 depict the time variance of critical reactor and plant parameters from the analysis of the load rejection transient from the design basis power and increased core flow point for a matrix of cases which involve normal scram
- speed, technical specification scram speed, and recirculation pump trip (RPT) in service and out of service.
ZN-NF-87-24 Analysis assumptions are:
Control rod insertion time based on WNP-2 measured data (normal scram speed) and technical specification scram speed.
Integral power to the hot channel was increased by 10$ for the pressurization transient, consistent with Reference 8.
Table 3.3 shows delta CPR values for a matrix of LRWB transients with the RPT out of service with both normal scram speed (NSS) and technical specification scram speed (TSSS).
Because a significant number of control rods are inserted into the core at exposures less than end-of-cycle (EOC) minus 2000 MWD/MTU, the system transients are expected to be insignificant for cycle exposures less than this value.
To confirm this, the LRWB was analyzed at the same 104%
power/106%
flow condition point for the end-of-cycle (EOC) minus 2000 MWD/MTU exposure condition for the bounding case of the RPT inoperable with TSSS.
The respective delta CPR values for the NSSS vendor and ANF fuels for this EOC minus 2000 MWD/MTU case are O.ll and 0.12.
These delta CPR values are also shown in Table 3.3 and are about half of the delta CPR values for the control rod withdrawal error (CRWE) event reported in Reference 1.
This shows that the delta CPR for the CRWE bounds plant operation up to EOC minus 2000 MWD/MTU.
For Cycle 3 exposures greater than EOC minus 2000 MWD/MTU, the other MCPR values defined in Table 3.3 are applicable.
3.2.2 Feedwater Controller Failure Failure of the feedwater control system is postulated to lead to a
maximum increase in feedwater flow into the vessel.
As the excessive feedwater flow subcools the recirculating water returning to the reactor core, the core power will rise and attain a
new equilibrium if no other action is taken.
Eventually, t'e inventory of water in the downcomer will rise until the high
ZN-NF-87-24 vessel level setting is exceeded.
To protect against wet steam entering the
- turbine, the turbine trips upon reaching the high level setting, closing the turbine stop valves.
The compression wave that is created, though mitigated by bypass flow, pressurizes the core and causes a power excursion.
The power increase is terminated by reactor
The evaluation of this event was performed using the scram and integral power assumptions discussed in 3.2.1.
Sensitivity results have shown that the calculated delta CPR is insensitive to the rate of feedwater flow increase, that EOC conditions are bounding because rods are inserted for lower cycle exposure, and that high flows are bounding because of higher axials in the core.
Because the total change in feedwater flow is the greatest from reduced power condition, the FWCF is more severe from reduced power conditions.
The FWCF transient event was analyzed from the lowest allowed power (47%) at increased core flow.
Figures 3.11 through 3.16 present key variables.
The delta CPR values for the co-resident fuel types for these three 47%
power/106$
flow transients are shown in Table 3.3.
A FWCF transient (47% power/106% flow) was also performed at EOC -2000 MWD/MTU which confirmed that the CRWE event is limiting for cycle exposures less than EOC
-2000 MWD/MTU.
As with the
- LRWB, partial insertion of the rods substantially reduced the change in critical power during this transient.
3.2.3 Loss Of Feedwater Heatin The Loss of Feedwater Heating (LOFH) transient has been analyzed on a generic basis for a wide cross section of BWR configurations.
This generic analysis is documented in Reference 9.
The Reference 9 analysis provides a statistical evaluation of the consequences of the LOFH transient for BWR/4, BWR/5, and BWR/6 plant configurations under conditions which cover the operating power flow map including increased core
ZN-NF-87-24 flow conditions.
Rather than use the 95:95 value in the reference
- report, a
bounding value was used.
The generic conclusions, when using a bounding value, support a MCPR operating limit of 1.15 for plants with a MCPR safety limit of 1.06, indicating a delta CPR of 0.09.
As noted in Section" 2.0 of this report, the WNP-2 MCPR safety limit for Cycle 3 continues to be 1.06; hence the LOFH transient requires a
MCPR operating limit of 1.15 for WNP-2.
3.3 Calculational Model The plant transient code used to evaluate the pressurization transients (generator load rejection and feedwater flow increase) was the ANF advanced code COTRANSA
. and XCOBRA-T( ).
This axial one-dimensional model predicted (2) reactor power shifts toward the core middle and top as pressurization occurred.
This was accounted for explicitly in determining thermal margin changes in the transient.
All pressurization transients were analyzed on a
bounding basis using COTRANSA in conjunction with the XCOBRA-T hot channel model.
3.4 Safet Limit The MCPR safety limit is the minimum value of the critical power ratio (CPR) at which the fuel could be operated where the expected number of rods in boiling transition would not exceed 0.1% of the fuel rods in the core.
The operating limit MCPR is established such that in the event the most limiting anticipated operational transient
- occurs, the safety limit will not be violated.
The safety limit for methodology presented input parameters and presented in Appendix all fuel types in VHP-2 Cycle 3
was confirmed by the in Reference 4 to have the Cycle 2 value of 1.06.
The uncertainties used to establish the safety limit are A of this report.
10 ZN-NF-87-24 TABLE 3.1 DESIGN REACTOR AND PLANT CONDITIONS FOR WNP-2 Reactor Thermal Power (104$ )
Total Recirculating Flow (106%)
Core Channel Flow Core Bypass Flow Core Inlet Enthalpy Vessel Pressures Steam Dome Upper Plenum Core Lower Plenum Turbine Pressure Feedwater/Steam Flow Feedwater Enthalpy Recirculating Pump Flow (per pump) 3464 MWt 115.0 Mlb/hr 101.8 Mlb/hr 13.2 Mlb/hr 529.15 BTU/ibm 1035. psia 1045. psia 1052. psia 1068. psia 974. psia 15.0 Mlb/hr 403.5 BTU/ibm 17.26 Mlb/hr
ZN-NF-87-24 TABLE 3;2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 High Neutron Flux Trip Void Reactivity Feedback Time to Deenergized Pilot Scram Solenoid Valves Time to Sense Fast Turbine Control Valve Closure Time from High Neutron Flux Time to Control Rod Motion Scram Insertion Times (normal)~
Turbine Stop Valve Stroke Time Turbine Stop Valve Position Trip Turbine Control Valve Stroke Time (Total)
Fuel/Cladding Gap Conductance Core Average (Constant)
Safety/Relief Valve Performance Settings Relief Valve Capacity Pilot Operated Valve Delay/Stroke 126.2%
10$ above nominal*
200 msec 80 msec 290 msec 0.404 sec to Notch 45 0.660 sec to Notch 39 1.504 sec to Notch 25 2.624 sec to Notch 5
100 msec 90% open 150 msec 556 'TU/hr-ft2-F Technical Specifications 228.2 ibm/sec (1091 psig) 400/100 msec
- For rapid pressurization transients a
108 multiplier on integral power is used; see Reference 8 for methodology description.
- Slowest measured average control rod insertion time to specified notches for each group of 4 control rods arranged in a 2x2 array.
12 XN-NF-87-24 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)
MSIV Stroke Time MSIV Position Trip Setpoint'ondenser Bypass Valve Performance Total Capacity Delay to Opening (80% open)
Fraction of Energy Generated in Fuel Vessel Water Level (above Separator Skirt)
High Level Trip (L8)
Normal Low Level Trip (L3)
Maximum Feedwater Runout Flow Two Pumps Recirculating Pump Trip Setpoint 3.0 sec 85% open 990.
ibm/sec 300 msec 0.965 73 in 49.5 in 21 in 5799.
ibm/sec 1170 psig Vessel Pressure
13 ZN-NF-87-24 TABLE 3.2 SIGNIFICANT PARAMETER VALUES USED IN ANALYSIS FOR WNP-2 (Continued)
Control Characteristics Sensor Time Constants Steam Flow Pressure Others Feedwater Control Mode Feedwater 1008 Mismatch Water Level Error Steam Flow Equiv.
Flow Control Mode Pressure Regulator Settings Lead Lag Gain 1.0 sec 500 msec 250 msec Three-Element 48 in 100%
Manual 3.0 sec 7.0 sec 3.3S/psid
14 XN-NF-87-24 TABLE 3.3 RESULTS OF SYSTEM PLANT TRANSIENT ANALYSES Event LRWB RPT Operable, NSS*
Maximum Neutron Flux Rated 295 Maximum Core Average Heat Flux Rated 115 Maximum System Pressure Q)~si ~
~Fue Fuel 0.25 0.23 LRWB RPT Inoperable, NSS LRWB RPT Operable, TSSS**
LRWB RPT Inoperable, TSSS LRWB EOC -2000 MWD/MTU RPT Inoperable, TSSS 390 370 440 304 121 121 127 112 1175 0.31 0.28 1170 0.33 0.29 1183 0.37 0.33 1167 0.11 0'2 FWCF (47% Power/106%
156 Flow),
NSS RPT Operable 54 1015 0.26 0.24 FWCF (47% Power/106%
Flow),
NSS RPT Inoperable FQCF (47% Power/106%
Flow),
TSSS RPT Operable 205 172 57 56 1020 0.31 0.29 1020 0.30 0.27 MSIV Closure With Flux Scram 668 130 1313 N/A NOTE:
All results are for the design power and increased flow point (104%
power/106% flow) unless otherwise noted.
'**Technical Specification Scram, Speed (TSSS)
30 25 HEA 3.
FLOM FLOM 20 Cl LLJ I
tt:1 5 O
2 5
50 5.0 0 '
0 '
0-6 0 '
1 '
1 '
- TIvE, SEc 1
~ 4 1 '
1
~ 8 I
I 2 '
I Figure 3.1 Load Rejection Without Bypass Results, RPT Operable, Normal Scram Speed
2.
VES EL WAT LEVEL (1N) 12 10 80 60 40 20 1
0 0.0 0 '
0 '
0 '
0 '
I
~ 0 1 '
TINE.
SEC 1 '
1
~ 8 Figure 3.2 Load Rejection Without Bypass Results, RPT Operable, Normal Scram Speed
60 50 2
HEA FLUX 3.
REC RCULAT1 N FLOW VES EI STE FLOM 40 I
~30 a
~20 100 1
3 5
'8.0 0 '
0 '
1
~ 2 1 '
2 '
2 '
T I t1E. - SEC 2 '
3
~ 2 4
~ 0 Figure 3.3 Load Rejection Without Bypass Results, RPT Inoperable, Normal Scram Speed
14 2 ~
VES EL MAT LFVEL 120 10 80 60 40 20 1
0 0.0 0 '
0 '
1
~ 2 1.6 2.0 2 '
2 '
Tlt1E.
SEC 3
~ 2 3
~ 6 Figure 3.4 Load Rejection Without Bypass Results, RPT Inoperable, Normal Scram Speed
60 50 2.
HEA FLUX 3 ~
REC RCULATl H
FLOM YES EL STE FLOM 40 I
w20 10 3
5 2
0 '
0-8 1
~ 2 1-6 2 '
2 '
- T1NE, SEC 2 '
3 '
3 ~ 6 I
0 I
4 ~
0o I
C Figure 3.5 Load Rejection Without Bypass Results, RPT Operable, Tech.
Spec.
Scram Speed
14 2.
VES EL MAT LEVEL 12 10
.80 60 40 20 1
0 0.0 0,4 0-8 1
~ 2 1 '
2 '
2 '
2 '
TIME.
SEC 3
~ 2 3 ~ 6
.0 0o I
Figure 3.6 Load Rejection Without Bypass Results, RPT Operable, Tech.
Spec.
Scram Speed
60 50 2.
MEA FLUX REC RCULATl H FLOM 4.
YES EL*STE FLOW 40
~30 o
~~20 10 3
5
'8.0 0-4 Figure 3.7 3
~ 6 3
~ 2 2 ~ 8 2 '
1
~ 2 1 '
2 '
TINE ~
SEC Load Rejection Without Bypass Results, RPT Inoperable, Tech.
Spec.
Scram Speed
14 2.
S EL MAT LEVEL l1H) 12 10 80 60 40 20 0 0.0 0 '
0 '
1
~ 2 1 '
2 '
2 '
Tlt1E.
SEC 2 '
3
~ 2 Figure 3.8 Load Rejection Without Bypass Results, RPT Inoperable, Tech.
Spec.
Scram Speed
60 50 2
HEA FLUX 3 ~
REC RCULATI YES EL STE N FLOM FLOM 40 a
LLJ IC~30 a
I w20 IK LLJ 10 2 3
~ 0 0 '
0 '
1
~ 2 1 '
2 '
2 '
T I t1E.
SEC 2 '
3
~ 2 3 ~ 6 Figure 3.9 Load Rejection Without Bypass Results, End-Of-Cycle Minus 2000 MWD/MTU Exposure, RPT Inoperable, Tech.
Spec.
Scram Speed
2.
YES EL MhT LEVEL (1H) 12 10 80 60 40 20 0 0.0 0.4 0 '
1
~ 2 1
~ 6 20 2'4 28 T I ME.
SEC 3 ~ 2 3
~ 6 4 '
Figure 3.10 Load Rejection Without Bypass Results, End-Of-Cycle Minus 2000 MWD/MTU Exposure, RPT Inoperable, Tech.
Spec.
Scram Speed
2I20'.
HEA 3.
REC I ~
YES FLUX RCULAT1 EL STE H
FLO'M FLO'N'6 A
~12 C) 80 40 2
10 12 T I t1E.
SEC 16 18 20 Figure 3.11 Feedwater Controller Failure Results For 47X Power And 106X Flow With Normal Scram Speed
20 2.
VES EL MAT LEVEL (1N) 12 80 40 10 12 T 1 ME SEC 14 18
~20 Figure 3.12 Feedwater Controller Failure Results For 47% Power And 106X Flow With Normal Scram Speed
24 200 2 ~
HEA FLUX 3
~
RFC RCULATI N
FLOW 4
~
VES EL STE I1 FLOW 160 CI LIJ I
~120 o
~80 40 10 12 TIME SEC 16 18 20 Figure 3.13 Feedwater Controller Failure Results, RPT Inoperable, Normal Scram Speed
160 2 ~
VES EL MAT LEVEL 120 80 40
-40
-80 10 12 T I [ATE, SEC 18 I
I 20 I
Figure 3.14 Feedwater Controller Failure Results, RPT Inoperable, Normal Scram Speed
20 2.
HEA REC YES FLUX RCULATl N FLOM EL STE FLOM Cl
~12 o
I
~80 LLJ 40 10 12 TlME.
SEC 16 Figure 3.15 Feedwater Controller Failure Results For 47% Power And 106% Flow With Tech.
Spec.
Scram Speed
12 2
VKS EL NAT LEVEL (JN>
10 75 50 25 10 12 T I t1E.
SEC 18 20 Figure 3.16 Feedwater Controller Failure Results For 47% Power And 106% Flow With Tech.
Spec.
Scram Speed
31 XN-NF-87-24 4.0 MAXIMUMOVERPRESSURIZATION Maximum system pressure has been calculated for the containment isolation event (rapid closure of all main steam isolation valves) with an adverse scenario as specified by the ASME Pressure Vessel Code.
This analysis showed that the safety valves of VNP-2 have sufficient capacity and performance to prevent pressure from reaching the established transient pressure safety limit of 110% of the design pressure.
The maximum system pressures predicted during the event are shown in Table 2.1.
This analysis also assumed six safety relief valves out of service.
4.1 Desi Bases The reactor conditions used in the evaluation of the maximum pressurization event are those shown in Table 3.1.
The most critical active component (scram on MSIV closure) was assumed to fail during the transient.
The calculation was performed with t'e ANF advanced plant simulator code COTRANSA(2),
which includes an axial one-dimensional neutronics model.
4.2 Pressurization Transients ANF has evaluated several pressurization events and has determined that closure of all main steam isolation valves (MSIVs) without direct scram is the most limiting.
Since the MSIVs are closer to the reactor vessel than the turbine stop or turbine control valves, significantly less volume is available to absorb the pressurization phenomena when the MSIVs are closed than when turbine valves are closed.
The closure rate of the MSIVs is substantially slower than the turbine stop valves or turbine control valves.
The impact of this smaller volume is more important to this event than the slower closure speed of the MSIV valves relative to turbine valves.
Calculations have determined that the overall result is to cause MSIV closures to be more limiting than turbine isolations.
32 ZN-NF-87-24 4.3 Closu e 0 All Main Steam solation Valves This calculation also assumed that six relief valves were out of service and that all four main steam isolation valves were isolated at the containment boundary within 3
seconds.
At about 3.3
- seconds, the reactor scram is initiated by reaching the high flux trip setpoints.
Pressures reach the recirculation pump trip setpoint (1170 psig) before the pressurization has been reversed.
Loss of coolant flow leads to enhanced steam production as less subcooled water's available to absorb core thermal power.
The maximum pressure calculated in the steam lines was 1287 psig occurring near the vessel at about 5 seconds.
The maximum vessel pressure was 1313 psig occurring in the lower plenum at about 5 seconds.
These results are presented in Table 2.1 and 3.3 for the design basis point.
33 XN-NF-87-24 5.0 RECIRCULATION FLOW RUN-UP The MCPR full flow operating limit is established through evaluation of anticipated transients at the design basis state.
Due to the potential for large reactor power increases should an uncontrolled recirculation flow increase occur from a less than rated core flow state, the need exists for. an augmentation of the operating limit MCPR (full flow) for operation at lower flow conditions.
Advanced Nuclear Fuels Corporation determined the required reduced flow MCPR operating limit by evaluating a
bounding slow flow increase event.
The calculations assume the event was initiated from the 104% rod line at minimum flow and terminate at 1208 power at 103% flow (flow control valve wide open).
This power flow relationship bounds that calculated for a constant xenon assumption.
It was conservatively assumed that the event was quasi-steady and a flow biased scram does not occur.
The power distribution was chosen such that the MCPR equals the safety limit at the final power/flow run-up point.
The reduced flow MCPRs were then calculated by XCOBRA( ) at discrete flow points.
The recirculation flow run-up analysis performed for WNP-2 Cycle 2
was
- reviewed, and the assumptions and conditions used for Cycle 2 are applicable to Cycle 3.
- Thus, the reduced flow MCPR operating limit for WNP-2 Cycle 2 is applicable to Cycle 3.
This reduced flow MCPR operating limit is presented in Figure 5.1 and tabulated in Table 5.1.
The MCPR operating limit for WNP-2 shall be the maximum of this reduced flow MCPR operating limit and the full flow MCPR operating limit as summarized in Reference 1.
34 XN-NF-87-24 TABLE 5 '
REDUCED FLOW MCPR OPERATING LIMIT FOR WNP-2 Core Flow
~Raced 100 Reduced Flow MCPR 0
e at n Limit 90 1.12 80 1.17 70 1.23 60 1.32 50 1.42 40
1.6
- 1. 5
~ l.i I
CC~1.3 0
K3 NOTE:
The HCPR operating limit shall be the maximum of this curve or the rated condition MCPR operating limit.
1020 30 40 50 60 70 BO 90 100 TQTAL CORE RECIRCULFIT ING FLOH
(%
RATED)
Figure 5.1 Reduced Flow MCPR Operating Limit
36 XN-NF-87-24
6.0 REFERENCES
1.
J.
E. Kraj icek, "Supply System Nuclear Project Number 2
(WNP-2) Cycle 3
Reload Analysis,"
XN-NF-87-25, Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, March 1987.
2.
R.
H.
- Kelley, "Exxon Nuclear Plant Transient Methodology for Boiling Nuclear Company, Inc., Richland, WA 99352, November 1981.
3.
M.
J.
- Ades, "XCOBRA-T:
A Comput'r Code for BWR Transient Thermal-
~*~
Volume 1 Supplement 2,
Advanced Nuclear Fuels Corporation,
- Richland, WA 99352, February 1987.
4.
J.
B.
- Edgar, Letter to
- WPPSS, Supplemental Licensing Analysis Results, ENWP-86-0067, Exxon Nuclear Company, Inc., Richland, WA 99352, April 15, 1986.
5.
T.
W. Patten, "Exxon Nuclear Critical Power Methodology for Boiling Water
- Richland, WA 99352, November 1983.
T.
L.
Krysinski and J.
C.
- Chandler, "Exxon Nuclear Methodology for Boiling Water-Reactors; THERMEX Thermal. Limits Methodology; Summary Company, Inc.,'ichland, WA 99352, January 1987.
7.
K. R. Merckx, "RODEX2 Fuel Rod Mechanical Response Evaluation Model," XN-March 1984.
S.
E.
- Jensen, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors:
'evised Methodology for Including Code Uncertainties in Determining Operating Limits for Rapid Pressurization Transients in Company, Inc., Richland, WA 99352, March 1986.
R.
G.
- Grummer, "A Generic Analysis of the Loss of Feedwater Heating Company, Inc., Richland, WA 99352, February 1986.
P r
~
~ II A-1 ZN-NF-87-24 APPENDIX A MCPR SAFETY LIMIT A.l I
RODUCTION Bundle power limits in a boiling water reactor (BWR) are determined through evaluation of critical heat flux phenomena.
The basic criterion used in establishing critical power ratio (CPR) limits is that at least 99.9% of the fuel rods in the core will be expected to avoid boiling transition (critical heat flux) during normal operation and anticipated operational occurrences.
Operating margins are defined by establishing a minimum margin to the 'onset of boiling transition condition for steady state operation and calculating a
transient effects allowance, thereby assuring that the steady state limit is protected during anticipated off-normal conditions.
This appendix addresses the calculation of the minimum margin to the steady state boiling transition condition, which is implemented as the MCPR safety limit in the plant technical specifications.
The transient effects allowance, or the limiting transient change in CPR (i.e.,
delta CPR),
is treated in the body of this report.
The MCPR safety limit is established through statistical consideration of measurement and calculational uncertainties associated with the thermal hydraulic state of the reactor using design basis
- radial, axial, and local power distributions.
Some of the calculational uncertainties, including those introduced by the critical power correlation, power peaking, and core coolant distribution, are fuel related.
When ANF fuel is introduced into a core where it will reside with another supplier's fuel types, the appropriate value of the MCPR safety limit is calculated based on fuel-dependent parameters associated with the mixed core.
Similarly, when an ANF-fabricated reload batch is used to replace a
group of dissimilar fuel assemblies, the core average fuel dependent parameters change because of the difference in the
A-2 XN-NF-87-24 relative number of each type of bundle in the core, and the MCPR safety limit is again reevaluated.
The design basis power distribution is made up of components corresponding to representative radial, axial, and local peaking factors.
Where such data are appropriately available from previous
- cycles, these factors are determined through examination of operating data for previous cycles and predictions of operating conditions during the cycle being evaluated for the MCPR safety limit. If operating data are not available, either because the reactor has not been operated or because appropriate data cannot be supplied to ANF, the safety limit power.distribution is determined strictly from the predicted operating conditions during the cycle being evaluated.
Operating data for WNP-2 during Cycle 1 operation was not evaluated because it is not considered typical of later cycle operation.
Operating data for WNP-2 during Cycle 2 and the predicted operating conditions for Cycle 3 were evaluated to identify the design basis power distributions used in the Cycle 3
MCPR safety limit analysis.
A-3 XN-NF-87-24 A.2 ASSUMPTIONS A.2.1 Desi n Basis Powe Dist ibutio The local, radial, and axial power distr'ibutions which were'etermined to. be conservative for use in the safety limit analysis are shown in Figures A-1 through A-3.
A.2.2 H draulic Demand Cu e
Hydraulic demand curves based on calculations with XCOBRA were used in.the safety limit analysis.
The XCOBRA calculation is described in ANF topical reports XN-NF-79-59(A),
"Methodology for Calculation of Pressure Drop in BWR Fuel Assemblies,"
and XN-NF-512(A), "The XN-3 Critical Power Correlation."
A.2.3 S stem Uncertaint es System measurement uncertainties are not fuel dependent.
The values reported by the NSSS supplier for these parameters remain valid for the insertion of ANF fuel.
The values used in the safety limit analysis are tabulated in the topical report XN-NF-524(A),
"Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."
A.2 '
Fuel Re ated Uncertainties Fuel related uncertainties include power measurement uncertainty and core flow distribution uncertainty.
The values used in the safety limit analysis are tabulated in the topical report, XN-NF-524(A), "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."
Power measurement uncertainties are established in the topical report XN-NF-80-19(A),
Volume 1,
"Exxon Nuclear Methodology for Boiling Water Rectors; Neutronics Methods for Design and Analysis."
4'
~
~, ~
A-4 XN-NF-87-24 A.3 SA ETY LIMIT CALCULATION A statistical analysis for the number of fuel rods in boiling transition was performed using the methodology described in ANF topical report XN-NF-524(A),
"Exxon Nuclear Critical Power Methodology for Boiling Water Reactors."
With 500 Monte Carlo trials it was determined that for a minimum CPR value of 1.06 at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition with a confidence level of 95%.
A-5 XN-NF-87-24 LL 0.91 L
0.95 ML 1.01 M
1.05 M
1.05 ML 1.01 L
0.95 LL 0.91 L
0.95 ML 0.97 H
1.07 ML'.89 H
1.04 H
1.07 M
1.03 L
0.96 ML 1.01 H
1.07 H
1.02 H
1.01 H
0.99 H
1.05 ML'.89 H
1.01 W
0.00 M
0.91 H
0.99 H
1.04 M
1.05 M
1.05 H
1.04 H
0.99 M
0.91 W
0.00 H
~1.00 M
0.95 M
1.04 ML 1.01 H
1.07 H
1.01 H
0.99 H
1.00 H
1.01 L
0.95 M
1.03 ML 0.91 H
1.04 M
0.95 H
1.07 ML 0.97 ML 1.06 LL 0.91 L
0.95 ML 1.01 M
1.05 M
1.04 M
1.07 ML 1.05 L
1.01 XlbCW)551 Figure A.l WNP-2 Cycle 3 Safety Limit Local Peaking Factors (ANF Fuel)
A-6 XN-NF-87-24 LL 1.03 ML 0.99 M
0.99 M
0.99 Ml 0.99 L
1.00 LL 1.03 L
1.00 M
0.99 H
1.03 H
1.02 MH 0.99 MH 0.99 ML 0.97 L
1.00 ML 0.99 H
1.03 L
0.91 H
1.02 H
1.01 MH O.SS MH 0.99 ML 0.99 M
0.99 H
1.02 W
0.00 H
1.02 H
1.01 MH 0.99 M
0;99 I
M 0.99 H
1.02 H
1.01 Le 0.91 W
0.00 H
1.02 H
1.02 M
0.99 ML 0.99 MH 0.99 H
1.02 H
1.01 Lo 0.91 H
1.03 ML 0.99 ML 0.97 MH 0.99 H
1.02 H
1.03 H
1.03 M
0.99 L
0.99 M
0.99 ML 0.99 L
1.00 LL 1.03
)OACH4523 Figure A.2 WNP-2 Cycle 3 Safety Limit Local Peaking Factors (G. E. Fuel)
WNP-2 CYCLE 3 DESIGN BRSIS RRDIRL POWER 12 10 Vl 43 C3 8
CQ 4C) 6 43 0.2
.0.4 0.6 0.8 1
1.2 BUNDLE POWER f.ACTOR 1.6 I
I 1.6 I
C Figure A.3 Radial Power Histogram For 1/4 Core Safety Limit Model
ZN-NF-87-24 Issue Date:
3/26/87 WNP-2 CYCLE 3 PLANT TRANSIENT ANALYSIS Distribution:
R.
E. Collingham J
~
G'.
Ingham S.
E. Jensen T. H. Keheley J.
E. Krajicek T. L. Krysinski J.
L. Maryott J.
N. Morgan G. L. Ritter G.
N. Ward H. E. Williamson J.
B.
Edgar/WPPSS (50)
Document Control (5)
~
~
8705120286 ORIGNN
.,~of ~
Washington Public Power Supply System PLANT DEFICIENCY REPORT/NONCONFORMANCE R~RT i
Np Numoer aa'o/I R Nuffder Sa(ety Related, QC x I Non-safetv Related K ii G
Non-nstalled E.cIpt.
Security System Other Than Non-inst.
I I Fire Protection sseniia Radwaste Originator Da C
//o Qg System No.
0
.0 Yalida Date
//'iJ/m PO/ pec Procedureq g7gg(
Full Desc"iption of Problem
( o( Hold Taga Vital HiiR (
E(iviroH~t~fJI 9vaf~ P~t~fiawi.o j: fat
$ (fa(')-(ol /rt/ i'fa~
d'or o/arafior 4'I f~oclk f (Ic f't(fP)"1U(o off.o7rf'q f."oo/CP Ig go/
J Qf P 1/y RP ly (foPcJQ~7c'c C
1 g ea1pI/'~t~ j'g ho'F
/1'fA ~y
) o oaf'/
go jof, P/oP w i 3's,
$ p (e'fy QLr+C $ 101
~
I Not Reportable I Reportable
+
~q r
az~l ava 1 uaIfPr
///2 g Date ASbK PC'I Other
, +Zanediate Disposition Use-As-Is Reject Rework Repair l) !A~l~wlloh3 la(1tiaot'/(a/aa Haat~)
~<<~~~ ~"-"~"~ b"'" ~-4 ot.-sg-osis-e.
~s-I
~o emntirhg Oman zat'on f-.,~~ (-'f1(<ff'~~ AS 14 /~hips((t/ @aspic~ a(c'uf cM~~~f TSS~~
gy O
use/Effect:
ZH/(beau((t~ Eocu~f(<77((v gl(J ~44' (~
f JJ Qwgh(70&/(vs('-c cl4vig.ohkHcA w(
QtS(f(f 1 01 C((aTl O+
I I
Corrective Acticn:
i ~
P~IfP(fair( W 1/I/C//PM~C tPU+f.f/ICW77d~
gJ+
c 5-M-Niff'g d neo~ us~ o/= 77~(ov(f c~?5.
D'oositicneW S-'ZO'-g6 Oa" e 4 Pic.
415.'o em ntin Oman'zation Approval:
+R ~>ed fo-
+PTM g+
Signature
+Wi S gnature QA'an-ture /
a hr or ss epor~
es pr or 5 ~i I F4, Date Date/
a, ta to return to service Sagnature Signature ANi xonature Data Date
'ffien I
i ~
~ a SQ,~ 'e."iienta lla uGCumafit PROCCOURC (aUaaSCR 1 3 1'7
~ & ~ e
~ a R 1 1'S 4S)
~in'ent'a bianace / at RcvisloN NUaascR pAcE (aUMScil ev'ew aflak ur /Da 7 (/a+
l.~.3.2-I5 of
>M
CMS-RA-27C and -27D are preamplifiers mounted on thermoelectric
- coolers, which are disconnected.
The equipment must be capable of operating in the event of a LOCA during which it would experience an environmental temperature of 128 F.
gualification documentation does not demonstrate capability o performing at 128 F, although an examination of the subcomponents provides a basis for judging that the equipment is not likely to fail to perform its safety func.ion, the deficiency is a documentation deficiency.
K. R.
Wsse
WASCCIFCCTOIa FOILIC OOW!SC 43 SUPPLY SYSTE1VI g ~
~
~
FROM'UBJECT:
REFERENCE:
June 3,
1986 K. D.
Co an, t naaer, HtlP-2'Technical
- 988U L. T. Harrold, Assis ant Direc or, Genera.ion Engineering - 944E NCR 286<<011, OUALIFICATION OF PREA'IPLIFIEPS CtiS-RA-27C AND -27D l!ITHOUT THER.'<OELECTRIC COOLERS A tes.
was run to verify the environmental quali ication of preamnlifiers CNS-RA-27C and
-27D without their hermoelectric coolers.
The test results verified the oreamplifiers are qualified without their thermoelectric coolers.
The test results are documented in QID 270101E.
This comoletes the resolution of NCR 286-011.
RAC/ssm
. INTEROFFICE MEMORANDUM SS2-PE-86-715 0
DISTRIBUTION: MAILDROP WNP-I FILE WNP.2 FILE~2 1 WNP-3 FILE WNPW FILE WNP.S FILE HGP FILE PKWD FILE LEGAL FILE 9 ADMINFILE RA Call F~
981F DM Porter 520 NS'orter 981C JE Rhoads 981F KR iiiSS E~
981F RAG/Lb 981i LTH/Lb 994E 1
I'C 1
\\
~ ~ <<Lilt I<1L Cnl ~.~
I NCR Nupcer Washington Pubic Power Sucply System PI~
OE.=~r"AGENCY R~ORT/NONMNFORMAN - R=ORT,l<
I rDR.Nurser I
Sas ety
- ected, 4
I Non-safetv Related ~ ii I
I Non-installed c.~z. I, Secu nzy yszem I
lcssent'ia' I
I Other Than Non-inst.
I I Fi=e Protec 'on I
Raowas.e 0" gxnato" c I-'W Date-$ C I Yalj.dated By I
- 0 r.L.=~
.Date L.-/9 -g5 olenoso alves/PSR-Y-X73 2, -X80/Q System No.
I 8/Spec/
roc au X83 1
5 2, -X84/1 8 2/Cntmt, Isolation Full Desc-ipticn of Problem S of IIo o Taos Vi ad. IeIR S
The subject solenoid valves are heat traced with thereal insulation around the valves'olenoid coil assembly.
fhe solenoid operator assembly consists of an electrical housina assembly (i.ceo rectifier, terminal block, position switches, ezc.)
and a solenoid coil assembly (i.e., coil, magnet, bobbin, etc.).
Conversations with a manufacturer representative and field inspecti ons have de termi ned:
(Continued on Paae 2 of 2)
I vi Not Reportable I
I Reportable
+
=va uato" Date Event Daze plan-lance
" pwver RRD I
I ves I
I hlo irrmediate Disposition lg Use-As-Ls I
Reject I
Reworx I
I Repai I
APE I
I Ott er Use as is, EQE has qualified the subject valves until R2 based actual field data
{TP 8.3 ~37), Valcor test
{QR 526-6042-1A)
)I the Leak Rate Testing Data
{PPM 7.4.6.1.2.4).
Cause/
I feca.:
it was determined by EQE that the subject valves were not qualified to perform there R.G. 1.97, Cat.
1 function, based on information from a manafacturer representative and field inspections.
The valves are still qualified to perform their isolation functions.
However the R.G. 1.97, Cat.
1 function qualification can not be demonstrated.
p ~ ~-d//(( tw<q Rework thermal insulation and re ve heat trace from the coil section or" the valves.
Due to the attached letter neat trace 6 insulation is no longer required.
The only corrective action will be to revise the Leak Testing Program to reflect yearly testing.
Ef leakage exceeds
{approx.
200 SCCM); the EPR o-ring near. the valve seat shouLd be replaced with a
~ silicone o-ring i repair is not d
ff g(
~I ts n<:c:e 1
n ff<ffffff<<>~
. IUy UVCff a S Qpazure
'b/t. gj g~g
~'a w<C U p ior o se v
~ <ffib inc s.U Daze Cn su sr~M C
AC I a,ff r
'en<c1 aJ s<Q
~
~
ffl <C I cs UCi ff
-," emen=nc
.anac r(~ate I I.ci u
@ca PIROCC>VaC NVaeeca RCVISIOfd fevs<sSCJt pa-c favaseea
~
n
~ 8 fff
~
~
ff
/
~ a
~
'W wl
~<aaa $ $ 4 St1 v'$4S1
Page 2'of-2
~
MBiiAL Washington Pub~'c Powe" Sucply Sys.em PUK DE."ICIB/CY RK."CRT/NONCOhFORMANC:- R" RT N& Nugae"
~> E ~ - AQ ~l PM Name" (Can-'nuat'on Shet 1)
Insulation of the coil assembly results in insufficient coil heat dissipation, which results in high coil operating temperatures with coil burn-out and/or rectifier failure probable after operating for any significant duration.
2)
Because installed solenoid valves have occasionally been operated in the present configuration for greater than a two year period, the likelihood that the valves can be reasonably expected to operate throughout a postulated six-month DBE accident is
~..remote.
If opened during accident conditions, the valve is expected to close due to the spring return-to-close featur e of the valve.
3)
Thermal degradation due to constant eneraization of the heat tracing strips and/or occasional energization of the solenoid coil has accelerate aged the valve seat (ethylene polypropylene)
"0B-ring seals.
Continued operation in this manner will increase the likelihood of valve leakage.
Post accident safety function (Regulatory Guide 1.97, Category
- 1) is to provide valve position indication.
Lead wires (tefzel) from internal position/reed switches are being heated beyond their design rating.
Field wiring to the upper valve internals may also be degrading (Ref.
IEN 84-68) due to the heat retaining aspects of these insulated valves Assurance of post accident position indication can not be demonstrated.
>"~refore, it is concluded that the subject valves are likely to fail to remain operable and
)tion indication could be lost.
They are also likely to leak through degraded seat "0"-rin<
'1 s.
The valves are still qualifiable to perform their containment isolation function.
However, the Regulatory Guide 1.97, Category 1 function qualification can not be demonstrated.
Corrective action to rework the thermal insulation and assess degradation of the limit switch lead wire and field hook up wire is required to establish Regulatory Guide 1.97 safety functio qua 1 ificati on.
PROCEDURE NUMBER RKVISIOIIteUMBKR PACE NUMBER
In ernal DistÃbu".ion Hl. Aeschliman - 956B JP Burn - 580 KD Cowan - 988U MS Davison 988U LT Harrold - 994E CR Hexum - 994E JF Peters - 927S PL Powell - 956B JE Rhoads - 981F SI Stevens - 956B Docket File - 956B PL2/LB - 956B CMP/LB 927M GCS/LB - 520 WNP-2 Files - 964Y MR 'Wuestefeld - 988U Plant Files 1300.2 - 927S bcc:
WG Conn -
BSR - 994i RG Graybeal - 927S LD Sham - 981C KR Wise - 981F RJ Barbee - 988U May 8, 1986 G02-86-411 Docket 30-397 Director of Nuclear Reactor Regula ion Attn:
Ms E.
G. Adensam, Project Direc.or BWR Project Directorate No.
3 Division of BMR Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dear Ns Adensam
Subject:
Reference 2}
NUCLEAR PLANT NO. 2 OPERATING LICENSE NPF-21, REQUEST FOR AMENDMENT TO OPERATING LICENSE, LICENSE CONDITION 3.6, ATTACHMENi 2, ITiM 3(a),
PASS VALVE DEFrERMENT, MITHDRAWAL Or
- Letter, G02-86-282, G.C. Sorensen (SS) to E.G.
Adensam (NRC), same subject, dated March 28, 2986 2}
- Letter, D.B-Vasallo (HRC) to R.W.
Caps ick (VY HPC),
NUREt'737, Item II.B.3, dated 'January l4, 1985 3)
Le..er, J.R. Miller (HRC) to W.G. Counsil (Northwest Nuclear Energy Co.),
NUREG-0737, Item II.B.3 Evalua-tion of Pos.-Acciden Sampling Capabili-ies, dated June 14, 1984 Re erence
- 1) reouested a oeferral of a licensing condition requiring
=ully quali ried COmpOnentS mee.ino RegulatOry Guiae 1.97, ReViSiOn
?
reouirements for six (6)
PSR valves u ilized in ohtainino containment aerosphere
- samoles, post accident.
The col-.ponents theoretical ly fail environmental oualifica-ion when exoosed to post acciden environmental conditions added to service condi ions resulting
-"rom hea. tracing and insulating o
the valves.
g aAg's/1<
~
~ D'OR:
Dl b%,>61 I ~~+r~
S/irr'C l PoR slGN
~ VRE oPr GC Sorensen I SEC
~ lON I
I l
l FoR *MR~4l.DP I RJr Ba>> 66 J> '
Krl Cowan I
MR WU6stef6 ld I
.Z~ Gl av6e6a I
l LDr.Sna. o
) I APPROVED MP% ~ r I S-2-e. A
'~
Ppb l.r;, Z/~ WI r;: D{W6 S
IG~
+I=I 077~
.E.
G'.
Adensam
~
Page Two May 8; 1986 REVEST FOR AMENDMENT TO LICENSE, CONDITION 16, ATTACH.2, ITEM 3(a),
PASS VALVE DEFFRMENT, WITHDRAWAL OF In a subsequent phone conversation between Messrs J.O.
Bradfute and F.
Witt of your staff and P.L. Powell, H.L. Aeschliman, R. Baric, and L.
Sharp of the Supply System on April 24, 1986, it was stated by the NRC that heat tracing and-insulating bf the lines and valves related to obtaining containment atmosphere samples for es imating core damage are not necessary.
This is based on the act that the basic requirement is to obtain a core damage es imate by measuring the noble gases in the containment atmosphere, which are not susceptible to plating out in the sample lines.
The insulating and hea tracing was installed to prevent plateou By the nature of the targeted isotopes to be analyzed, plateout canno. occur and there is no longer a need
.o provide insula-tion and heat tracing.
Accordingly, the Supply System is revising
.he Post Accident Sample System to eliminate the requirements for heat tracing and insulating on the subject sample lines and valves.
References 2 and 3 were provided by the sta f to document this position.
Additionally, reference
- 1) identi ied that a reanalysis e
or. and in-spection program would be conducted to verify the valves had not a
11 suffer d
ed oegradatson suf,scient to render them not capable of surviving f
cuay a potential accident exposure.
That program has been completed and the actual service condition resulting from heat tracing is signi,ican ly lower than the conserva ive values assumed in the original qual-i,-ica ion assessme".
Coupled with the removal o, the requirement or heat tracing and the lower actual to date service conditions the valves contain su,,icient estimated li,e to remain inservice.
With this revision the requested defermen. is no longer necessary; renoval o
heat tracing and insulation requirements resulting in elimi-na ion of the higher service temoerature will allow the valves to meet qualifica ion reouiremen.s.
Hence the reoues o defer qualifica ion of the six PSR valves until the second re.ueling outage is withdrawn and
.he prooosed amendment to License Condition I6 is o be revised.accor-dinaiy.
. E.
G.
Adensam Page Three May 8, 1986 REOUEST FOR AM""NOMENT TO LICEHSE, LICENSE CONDITION 16, ATTACH. 2,
'ITEM 3(a),
PASS VALVE DEFERMENT, VITHDRAWAL OF It is requested that the sta.f provide confirmation of this oosi ion to the Supply System similar to that provided bv references 2 and 3.
Should you have any further ques ions please contac Mr. P.
L. Powell
- Manager, MNP-2 Licensing.
, I Very truly yours, I
G.
C. Sorensen, Manager
.Regul atory Programs PLP/bk cc:
JO Bradfute -
NRC C
Eschel s EFSEC DB Martin NRC RV E
Revell - BPA NS Reynolds - BLCPKR NRC Site Inspector
~
~
WASHIMCTOPC tl,'dl.IC O'OWtt 4P SUPPLY SYSTEM
.. INTEROFFICE MEMORANDUM
"""-PE-8o-606 DISTRIBi1~.GN:
MAILDROP DATE:
TO."
FROM:
SUBJECT Hay 2, 1986
-- ow'a ='"Hanaoer 'NP-2-"Technical'-
988U arro d, Assistant Director, Generation Engineering - 994E NCR 4286-042 (INSUFFICIENT INSULATION ON VALCOR SOLENOID VALVES)
WNP-I FILE WNP-2 FILE~~
WNP-3 FILE WNP4 FILE WNP.S FILE HGP FILE PKWD FILE LEGAI FILE ADMINFILE
REFERENCE:
1)
Valcor Engineering Corporation., Test Report No. QR526-6042-1A, dated 9/83 2)
Procedure Number TP 8.3.37, "Temperature Testing of PSR Valves" 3)
PPH 7.4.6.1.2.4, "Containment Isolation Valve and Penetration Leak Test Program" 4)
Supply System Calculations:
EQ-02-86-02, EQ-02-86-03, and EQ-02-86-12 HJ Heyer DW Porter NS Porter~~~
JE Rhoads KR Wise KRW/Lb LTH/Lb EQE Files C
Hexum QI0 361014E 981F 520.
981C 981F 981F 981F 994E 981F 994E 981F 5t4 I
I The purpose of this letter is to close out NCR 8286-042.
This NCR was issued on six Valcor solenoid valves because insulation on the coil housing did not provide sufficient heat dissipation.
EQE has qualified the subject valves until R-2 based on current vendor test reports (Reference 1), field measured temperatures (Reference 2),
and leak testing (Reference 3).
Supply System calculations (Reference
- 4) have been prepared to support a preliminary qualification.
QID 361014E will require revision to finalize qualification and determine qualified life beyond R-2.
In support of this oualification, the plant leak testing program shall be revised to reflect yearly tests and the insulation shall be removed from the coil regi on on al 1 si x val ves.
No hardware changes are required.
- However, when valve leakage becomes excessive (approximately 200 SCCH), the EPR 0-ring near the valve seat should be replaced with a silicone O-ring.
A PHR is attached for your approval to authorize Engineering to revise the design data base to reflect the above.
HJH/ssm WP 102 R4 P tt)
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- 5) ASSIGNED PLAUDIT SYSTEM ENGINEER
- 5) TECHNICALMERIT YES HO PLANT SYSTEM EHCWEER/DATE
- 7) COHCKPTl!ALDESIGN AUTHOIUKATIOH TKCHHICALSVPERVISOR/DATE
>>S) PIHALDESIGN AUT)IORIZATIOIL TECH%GAL SUP ERVISORIDATE
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- 10) PLAHT SYSTEM EHQIHEER/DATE
- 11) TECHNICALGROUP SVPERVISORIDATE
- 12) PLANTTECHHICALMANAGER RKVIEW REQUIRKD YES NO
- 13) PLANTTECHKICAI.MANAGER/DATE
PROPOSKD IMPLEMENTATION DATE
- 15) PMR APPROVED YES MO PRIORITY PLANT MANACERIDATE id) aWR i?Of<?
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- 17) fCR'S IKITIATEDT YES HO tlfYES LLST ON CONllKUATIOHSHEET)
- 18) INSTALLATIONAND STATIC TESTING COMPLETE PLANT SYSTEM ENGINEER/DATE
- 19) COHTROL ROOM OPERATING PROCEDURE AND TOP TIER DRAWINGS UPDATED, MODIRCATIOHCOMPLETE.
SYSTEM OPERABILITYTESTING COMPLETED.
SHIFi MANAGER/DATE
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/
21a) COMTINUATIONSHEET YES NO PLAHT ADMINISTRATIVEMAKAGER/DATE
AMENDMENT NO.
17 July 1981 APPENDIX B NNP-2
RESPONSE
TO REGULATORv ISSUES RESULTING FROM TMI-2
NNP-2 AMENDMENT VO.
23 February 1982 II.B.3 POST-ACCIDENT SAHPLING CAPABILITY Position.
A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be per-formed to determine the capability of personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, respectively.
Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products.
If the review indicates that personnel could not promptly 'and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.
A, design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capa-bility to promptly quantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionuclides that are indicators of the degree of core damage.
Such radionuclides are noble gases (which indicate cladding. failure), iodines and cesiums (which indicate high fuel temperatures),
and nonvolatile isotopes (which indicate fuel melting).
The initia3. reactor=~1'ant spectrum should
- correspond to a Regulatory Guide 1.3 -'or 1.4 release.
The review should also consider the effects of direct radiation from piping and components in the auxiliary bui.lding and possible contamination and direct radiation from 'airborne" effluents.
If the review indicates that the analyses requi'red cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procuremeat shall be undertaken to meet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.
Procedures shall be provided to perform boron and chloride
'hemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4.source term)'..Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis with'n an hour and the chloride sample analy-sis within a shi
).
Clar'fication The following items are clarifications of recuirements iden-tified in NUREG-0578, NUREG-0660, or he September 13 and October 30, 1979 clarification letters.
a.
The l'censee shall have the caoability to promptly obtain reactor coolant samples and containment B.2-12
AHENDMENT NO.
17 July 1981 b.
atmosphere samples.
The combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a
sample The licensee shall establish an onsite radiologi-cal and chemical analysis capability to provide, within the 3-hour time frame established
- above, quantification of the following:
1 certain radionuclides in the reactor coolant and containment atmosphere that may be indi-cators of the degree of core damage (e.g.<
noble gases, iodines and cesiums, and non-volatile isotopes);
2.
hydrogen levels in the containment atmosphere; 3.
dissolved gases (e.g.,
Hp), chloride (time allotted for analysis subject to discussion below),
and boron concentration of li.quids.
4.
alternatively, have inline monitoring capa-bilities to perform all or part of the above analyses.
c.
Reactor coolant and containment atmosphere sampling during post-accident conditions shall not reauire an isolated auxiliary system (e.g.,
the letdown system, reactor wate cleanup svstem (RES)) to be placed in operation in order to use the sampling system.
d.
P essurized reactor coolant samples are not requi ed if the licensee can quantify the amount of dissolved gases with unpressurized reacto coolant samples.
The measurement of either "otal dissolved gases or Bp gas in reac or coolant samples is conside ed adequate.
Measuring the 02 concentration is recommenaed, but 's not mandatory.
e.
The time for a chloride analys's to be per ormed is dependent-,
upon two factors:
(1 ) if the plant's coolant water is seawate or brackish
- water, and (2) if "here is only z single barrier between prima v containment systems and the cooling water.
Unaer both of the above con-
WNP-2 AMENDMENT NO.
17 July 1981 ditions the licensee shall provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being taken.
For all other cases, the licensee shall provide for the analysi's to be completed within 4 days.
The chloride analysis does not have to be done onsite.
The design basis fox plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that it is "possible 'to obtain and analyze a sample without radiation exposures to any individu'al exceedin'g the cri-teria of GDC 19 (Appendix A, 10 CFR Part 50)
(i.e.<
5 rem whole body, 75 rem extremities).
(Note that the design and operational review cri-terion was changed from the operational limits of 10 CFR Part 20 (NUREG-0578) to the GDC 19 cri-terion (October 30, 1979 letter from H. R. Denton to all licensees.))
The analysis of primary coolant samples for boron is required for PWRs.
(Note that Revision 2 of Regulatory Guide 1.97, when issued, will likely specify the need for primary coolant boron analy-sis capability at BWR plants.)
If inline monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide backup sampling'hrough grab samples, and shall demonstrate the capabil-ity of analyzing the samples.
Established plannin'g for analysis at offsite facilities is acceptable.
Equipment provided for backup sampling shall be capable of providing at least one sample per day for '7 days following onset of the accident and at least one sample per week until the accident condition no longer exists.
The licensee's radiological and chemica1 sample analysis capability shall include prov'sions to:
1.
Zaentify and cuantify the isotopes of the nuclide categories discussed above to levels corresponding to the source terms civen in Reculatory Guides 1.3 or 1.4 ana 1.7.
Where necessary and practicable, the ability to ailute samples" "o provide capabili"y for measurement and reduction of personnel expo-sure should be provided.
Sensitiv ty of onsite lieu'd sample analysis capabil'ty 3.2-14
WNP-2 AMENDMENT NO.
17 July 1981
'hould be such as to permit measurement of nuclide concentration in the range from approximately 1 ACi/g to 10 Ci/g.
2.
Restrict background levels of radiation in the radiological and chemical analysis faci-lity from sources such that the sample analy-sis will provide results with an acceptably small error (approximately a factor of 2).
This can be accomplished through the use of.
sufficient shielding around samples and out-side sources, and by the use of ventilation system design which will control the presence of airborne radioactivity.
j.
Accuracy, range, and sensitivity shall be ade-quate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coo3.ant systems.
k.
Zn the design of the post-accident sampling and analysis capability, consideration should be given to the following items:
1.
Provisions for purging sample lines, for reducinng plateout in sample lines, for mini-mizing sample loss or distortion, for pre-venting blockage of sample, lines by loose material in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions to limit reactor coolant loss from a rupture of the sample line.
The post-accident reactor coolant and containment atmosphere samples should be'epresentative of the reactor coolant in the core area and the containment atmosphere fo3.lowing a transient or accident.
The sample lines should be as short as possible to minimize the volume of fluid to be taken from containment.
The residues of sample collection should be returned to containment or to a closed system.
2.
The ventilation exhaust from the samp3.ing station should be filtered with charcoal adsorbe s and high-effic'ency p'a ticulate ai" BEPA) <<i3.ters.
Guidelines for analytical or i..strumentation range are g'ven 'n Table
~ i.B.3-1.
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i
~
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~ g AW WNP-2 AMENDMENT NO.
35 November 3.984 WNP-2 Position WNP-2 is using a General Elect. ic post-accident sampling sys-tem which will be capable of sampling the primary containment and reactor building atmosphere and of obtaining liauid sam-ples from the reactor, RHR loops, and various reactor build-ing sumps.
This system is d signed to obtain grab samples
~
'hich may be analyzed on site or transported to offsite facilities for more detailed analysis if necessary.
The sam-ple station is located in the radwaste building and is shielded to reduce radiation exposure rates to the operator.
All remote-operated valves are controlled from this area.
Lead pigs are provided for radiation protection when trans-porting samples either to onsite facilities or off site.
A more'etailed description follows.
Gas samples will be obtained from locations in the drywell.
the suppression pool atmosphere, and from the secondary con-tainment atmosphere.:.-The samp3.e system is designed to oper-ate at pressures ranging from subatmospheric to maximum design pressu es of the primary and secondary containment.
Heat-traced sample 3.ines are used outside the primary con-tainment to prevent precipita"ion of moistu e and resultant loss of particulates and iodines in the sample lines.
The gas samples may be passed through a particulate filter and silver zeolite cartridge for determination of particulate activity and iodine activity by subseauent analysis of the samples on a gamma spectrometer system.
'Alternately, the sample flow bypasses the particulate/iodine
- sampler, is chilled to remove moisture, and a LS-milliliter grab sample can be taken for determination of gaseous radioactivity and for gas composition by gas chromatography.
This size sample vial has been adopted for all gas samples to be consistent with present. off-gas sample via3. counting actors.
Reactor coolant samples will be obtained from two points in the jet pump pressu e instrument system when the reac or is at pressure, The jet pum'p pressure system has been dete mined to be an optimum sample point for accident conditions.
The pressure taps are we3.1 protected from damage and debris.
Xf the recirculation pumps are secured, the water level will be raised about 18" above normal.
This provides natura1 ci"culation of the bulk coolant past the taps.
Also, the pressure aps are located sufficientlv low to permit sampling at a reactor water level even below the lowe core support plate.
NNP>>2 AMENDMENT 'NO.
35 November 1984
" A single sample line is also connected to both loops in the RHR system.
This provides a means of obtaining a reactor coolant sample when the reactor is depressurized and at least, one of the RHR loops is operated in the shutdown cooling mode.
Similarly, a suppression pool liquid sample can be obtained from the RHR loop lined up in the suppression pool cooling mode.
Samples from the five drain sumps in the reactor building aze also available.
The sample system isolation valves are controlled from the local control panel.
The sample system is designed for a purge flow of one gpm, which is su ficient to maintain turbu-lent flow in the sample line.
Purge flow is returned to the suppression pool.
The high flush flow also serves to 'allevi-ate cross-contamination of the samples when switching from one sample point to another.
'll liauid samples are taken into septum bottles mounted on sampling needles.
The sample station is basically a bypass loop on the sample purge line.
Zn the normal lineup, the sample flows through a conductivity cell (readable range 0.1 to 1000 micromhos/cm) and then through a ball valve bored out to 0.10-milliliter volume.
Plow through the sample panel is established, the valve is rotated 90',
and a syringe is used to flush the sample plus a measured volume of diluent (gen-erally 100 milliliters) through the valve and into the sample bottle.
This provides a dilution of 100:3. to the sample.
Alternately, the valve sampling sequence can be repeated 10.
times to provide a
1 ml sample diluted 10:1.
The sample is transported to the laboratory for furthez dilution and sub-sequent analysis.
Alternately; the sample f:ow can be di-ve"ted through a 70-milliliter bomb to obtain a large pres-surized volume.
This 70-milliliter volume can be circulated and depressurized into a known volume gas expansion chamber.
The pressuze change in this chamber will be used to calculate
~the Total Dissolved Gases in the reactor coolant.
A grab sample of 'these gases may be taken through a septum port for subsequent analysis.
Ten-millili ez aliquots of this degassed liquid can also be taken for on or of site chemicaL analyses ecrui ing a relatively Large samp3.e.
A radia ion monitor in the liquid sample enclosure monitors Liquid flow from the sample s ation to provide immediate assessement of the samp3.e ac" ivity level.
This monitor also provides infor-mation as to the effectiveness of the demineral'"ed water flush'ng of the sample sys em following sample opera"ion.
The cont ol ins"rumenta ion is ins a'3.ed in two 2' 2'
6'igh standard cabinet control panels.
One panel contains the conduc"'v'ty and radiation level readouts.
Another contxoL panel conta'ns he flow, pressure and tempera.ure ind. cato"s, and the various cont oL valves and switches..
B.2-16a
AMENDMMT NO.
35 November 1984 A graphic display panel, installed directly below the main control panel, shows the staus of the pumps and va3.ves at all times.
The panel also indicates the relative position of the pressure gauges and other items of concex'n to the operator.
The use of this panel will improve operator comprehension and assist in trouble-shooting opex'ation.
. Appropriate sample handling tools, a gas sampler vial posi-tioner and gas vial cask are available to the operator at the sampling station.
The gas vial is installed and removed by use of the vial positioner through the front of the gas sampler.
The vial is then manually placed down in the cask with the positioner which allows the vial to be maintained about 3 feet from the individual performing the operation.
The small-volume (10 ml) liquid sample is remote3.y obtained thxough the botto~ of the sample station by use of the small-volume cask and cask positioner.
The cask positioner holds the cask and positions the cas'k directly under the liquid sampler.
The 'sample vial is manually raised within the cask to engage the hypodermic needles.
When the sample vial has been filled, the bottle is manually withdrawn into the cask.
'The sample vial is always contained within lead shielding during this ope ation.
The cask is then lowered and sealed prior to transport to the laborato~.
A large-volume cask and cask positioner 'is 'available or transpoz ing large licuid samples.
A 27 milliliter'bottle is contained within a lead shielded cask.
This sample bottle is raised from its location in the cask to the sample station needles for bottle filling.
The sample station vill only deliver 10 milliliters to this sample bottle.
When filled.
the bottle is withdrawn into the cask.
The sample bottle is always shielded by 5 to 6 inches of lead when in position under the sample station and during the ill and wi~Ddraw
- cycles, thus preventing opex'ator exposure.
The cask is transported to the required position under the sample station by a dolly cask positioner.
When in position this cask is hydraulically elevated approximately 1.5 inches by a small hand pump for contact with t'e sample station shielding under the liquid sample enclosu e f3.oor.
Tne sam-ple bot le is raised, held, and lowered bv a simple push/
pull cable.
The cask is sealed by a threaded top plug tha"
'nserts above the sample bottle.
The weight of this la ge-volume cask is appr oximately 700 pounds.
B. 2-3.6b
AMENDMENT NO.
35 November 3.984 The particulate filters and iodine cartridges aze removed via a drawer arrangement.
The quantity of activity which is ac-cumulated on the cartridges is controlled by a combination of flow orificing and time sequence contzol of the flow valve opening.
In addition, he deposition of iodine is monitored during sampling using a radiation detector installed adjacent to the cartridge.
These samples will hence be limited to activity levels which will normally not require shielded sample carriers to transport the samples to the laboratory.
The power supply to the sample station and all associated equipment will not be shed during accident conditions.
The system design is such that a sample can be dragon and analy ed
,within the required 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, aftez a one hou-preparation time.
The post-accident sampling station will provide conductivity measurements in line as an indicator of liquid chemical con-
'centrations and changing chemical conditions.
The system allows collection of g ab samples for gas analysis of 02, N2, H2, and di"ect gamma spectrometric determination of aliquots of gas samples.
The system also allows collection of iodine samples on a silver zeoli 'e car"ridge to minimize noble gas interference. in the determination of iodine iso-topic content.
Liquid samples will be analyzed fo" pH using a semi micro pH electrode and additionally analyzed for boron and chloride using ion chromotography..
An aliquot of the sample may also be analyzed for gross ac ivity or isotopic content, by gamma ray spectromet"y.
All laboratory analysis-meet Regulatory Guide 1.97 requirements for sensitivity and
- zange, with the exception of the range foz dissolved gases.
.'however, the ana ytical capability for dissolved gases is consistent with the maximum dissolved gas concentrations expec ed for BNRs.
The post-accident sampling system will be used to perform
.periodic reactor coolant sample analyses for gamma isotopic
- content, chlor ide, conductiv 'y, pH, and total dissolved gas.
Every six months, for training and operability testing; a liquid grab sample will be drawn, transported, and analyzed in ~De Hot Lab or gamma isotopic content.
This sample will be handled as a pos accident highly radioactive sample.
In
- addition, eve y six months, a containment air sample will be ana3.yoked fo
Classroom training will also be provided on system operat'on and proper handl'ng of high'v radioactive samples:.
B. 2-16c
HHP <<2 A~~B',T N. 36 Decenher 1985 Based on information developed by General Electric, the Supply System has developed plant specific procedures for the determination of the extent of core damage under accident conditions.
The procedures provide for distinguishing be-tween fuel cladding fai1ure and fuel melt based on isotopes present and concentration.
The extent of damage is based on concentrations pzesent of isotopic mixture of Xe, Kr, I, and Cs.
The estimated maximum potential whole body dose to retrieve a
reactor coolant sample under worst case accident conditions is 0.36 rem; the source being airborne noble gas activity in the radwaste building from effluent releases.
Lapsed time is about one hour.
The maximum dose rate from a 0.1 ml reactor coolant sample (1
hour decay) in a 4'hick lead transport cask is less than 5 mR/hr at one foot.
Exposure to analyze a sample is expected to be less than 100 mR.
All valves used are fully qualified for the environment in which they are located inside and outside reactor containment.
Power for. the post accident sampling equipment is supplied from ~Fjr-. Division 1 or Division 2 czitical po~er sources and will be available during accident conditions.
The staff review of this position in NUREG-0892,.dated December
- 1982, recognized several issues requiring resolution and consolidated them in Licensing Gotjdition 9.
Subsequent Supply System submittals, primarily Amendment 23 to the
- FSAp, resulted in the staff finding the Post Accident Sampling System acceptable in Supplement 4 NUREG-0892, secton 9.3.2.4.
A requirement to have the system completed and operable prior to exceeding 5% power was made a condition to the license (NPF-21 issued December 20, 1983).
Supply System lettez'02-84-272 dated April 27, 1984 reported the system completed and operable thus satisfying the licensing condition.
B.2-16d
"PAGE 1
OF 2
Vl'JLS~GTON PUBIC PO~ SUPPLY SYSTLbf 10(1CO I FOR ~lAilCE REPORT (SKK INSTRVC IONS ON RKVCRSK SIDE)
I HCA NQ DATE February 11, 1986 QUALITYCLASS/A5ME CLASS
- a. SUPPL,Y SYSTEM PROJECT/PLANT/DEPT, ViNP-2 IL QAICINATOR 2 iy 'St+ JIy M.
Me er/J.
Costello 5
VALIDATEDSY 2-if%
So PHYSICAL I QCATION QF NONCONFORMANCE Reactor Buildin le ORGANIZATION/OEPARTMKNT Eauioment ualification Enaineerina Se ORGANIZATIQNlOEPARTMCNT 10, REQUIREMENT sOVRcC ggP 2
FSARr a
1 I, SUPPLIER HAME/P,Q, NQ /CQNTRACT NQ Valcor Enq.
Corp./P19576/220 IJ 4
Iz4l Q
III Q
C 12, ARDWARK/SOPTWARC IT M NQ /DCSCRIPTIOtl System Solenoid Valves/PSR-V-X73/2,
-X80/2, -X83/1 8 /2, -X84/1 8 2/Cntmt. Isolation 12 FVLI DESCRIPTION QP. NQNCONFQRMANCEl The subject solenoid valves are heat traced with thermal insulation around the valises'olenoid coil assembly.
The solenoid operator assembly consists of an electrical housing assembly (i.e., rectifier, terminal block, position switches, etc.)
and a solenoid coil assembly (i.e., coil 1 magnet, bobbin, etc.).
Conversa-tions with a manufacturer representative and field inspections have determined; 1)
Insulation of the coil assembly results in insufficient coil "heat dissipation, which results in high coil operating temperatures with coil burn-out and/or
~ rectifier failure probable after operating for any significant duration.
Continued on paae 2 of 2.
14, KKVIKWEO FOR RKPQKTASILITYPKR IOCFR21 ANO/OR SOBS( ~ )
)
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NVMSER PAGE QF
NRC No.
February 14, 1986 Page 2 of 2 2)
Because installed solenoid valves have occasionally been operated in the present configuration for greater than a two year period, the likelihood that the valves can be reasonably expected to operate through-out a postulated six-month OBE accident is remote.
If opened during accident conditions,'h'e'Valve 'is expected to close due to the spring return-to-close feature of the valve.
3)
Thermal degradation due to constant energization of the heat tracing strips and/or occasional energization of the solenoid coil has accelerate aged the valve seat (ethylene polypropylene) "0"-ring seals.
Continued operation in this manner will increase the likelihood of valve leakage.
4)
Post accident safety function (Regulatory Guide 1.97, Category
- 1) is to provide valve position indication.
Lead wires (tefzel) from internal
~ position/reed switches are being heated beyond their design rating.
Field wiring to the upper valve internals may also be degrading (Ref.
IEN 84-68) due to the heat retaining aspects of these insulated valves.
Assurance of post accident position indication can not be demonstrated.
Therefore, it is concluded that the subject valves are likely to ail to remain operable and position indication -could be lost.
They are also likely to leak through degraded seat "0"-ring seals.
The valves are still qualifiable to perform their containment isolation function.
However, the Regulatory Guide 1.97, Catagory 1 function qualification can not be demonstrated.
Corrective action to rework the thermal 'insulation and assess degradation of the limit switch lead wire and field hook up wire is required to establish Regulatory Guide 1.97 safety function qualification.
oasoasoror rsssts rOas ~
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!'slitted cosssercial grade ASCO solenoid pilot valves tagged as ROA-SPY-)O, -ll, ->E 17 1Iere qualified by means of engineering analyses supported by separate ma<<ri>>s test data.
Th~~~ ~~mp~~~~t~
op~~~t~ the ROA ~y~t~~
H~AC air dampers t Control Center Rooms in the.Reactor.Building during postulated accidents so as to ensure that he rooms'mild environment" classi ication status is not compromised.
- However, these (continued on page 2 of 2)
QD I
BOG PIBPO' cS B
I,R po="Bt)
+
Ptegl'1-BiiBiPi.
"va'UcT.
Da.e tso>>
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I Reject
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I I Repa'=
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mTio m
S s!O i~ Gan dat On s >1o>>o>>
. n>>
i
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Cause:
Equipment Class uporade
=rom "passive" to "active'"
'ff'+Ac'<<f sfrfcstsz'o~ iv Jc>>fc.ocr(gsv'y39cs on!I -) Jo'cha defs!fsM P <<gg E-.,ec-.':
Ouali-,ication of;he equipment to tte n w required sa,ety cia-ssi-,ica ion I par~m~te~ cannot be demonstra ed! o Jwc J O4' "F I/ QSC SPA'oyer dye Prnble~. ~~
S
~V>> f1v
~ C s~~ ~
gRV')
I..Replac installed ASCO HBX8320A1 and 81 mocels with ASCO Huclear Service "Hp Series"
- models.ji ASCO Iiodel HP832GA172E has.
been procured via Supolv Sys-em P.O.
-:.071.""=0 for,he sPeci ic aPPlicction stated (Attach!sent 2).
Attacrmers 3 indicates that the rePlacement (continued on page 2 of 2)
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. Mes%in~on I"-P~'c Pose" Sv"p1y System PI&i D=-ZC O'DY FI~VTi/NONKGRHAN~RRRT Gcn='nva-ion Snee=)
Full Description of. Problem (Continued)
'components have had their Equipment Classification upgraded from "passive" to "active" via DCP 84-1390-0AR, pages 015 and 016 (Attachment 1), invalidating their current qualification.
As these components are exposed to steam environments during postulated HELB accidents'for which they are required to actively operate (i.e., isolate when
~ de-energized as a result of manual initiation or receipt of a F, A, or Z signal),
qualification of these components:can no longer be assured in the absence of test da La ~
Although qualification o the equipment cannot be demonstrated, an engineering review and evaluation has determined that the equipment is "not likely to fail" to perform their required safety func.ion or ail to mitigate a postulated design basis accident.
This determination is primarily based upon the fact'hat the subjec.
solenoid pilot valves, when de-energized, return to "fail safe" position due to their spring return feature to the closed position..This results in isolation a-,"the POA system HYAC air dampers.
Addi.ionally, in accordance with the provisions of 10 CFR 50, Section 50.49, any replacemen.s for the subject components are required to satisfy the requirements oI NURE6-0588,'ategory I (refer to NCR 286-137).
~<, ~g
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Wreg 8'CP g)g nQ Chaiipg Ag ) e5AZJJ<SP P ~Crn+
d@4r)P~Ion7 e Correcti ve Action
( Conti nued q "NP Series" components pre readily available for installation in the plan.
during the current refueling/maintenance outage.
2)
For the EPNs and application stated, establish qualifica.ion documentation
=or the replacement models.
P R C C K~ iJ R E H IJM8 E R
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- 2) OUALI17 CLASS
- 3) SAPET7 RELATED TES ~ HD QD-A= 5'5-3d ~'sf '
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4O) DRIGMATOR/DATE
- 5) ASSIGNED PLANT S~~EI4 ENGIHEER
- 4) TKCIIHICALNERIT YES NO T) COttCEPTUAL DESIGN AUTttORZATIOH
./PMALDESIGH AUTIIORZATIOH
- 44) ORIGNATOR QNPRVISOR/DATE C(ZA(;b PLANT 5YSTEN ENGINEEIVDATE a
TKCIIHICALSUPERVISOR/DATE TECttH~ SUPERVISOR/DATE PLART $KYKTY/AUTHOC2AT)O)I Pa) DCP NO.
- 00) ~ ALARAREVIEW IIANAGKR~/CtIENISTRY 1I/) PLANT SYSTEN KNGMEER/DATE
- 11) TECHNICALGROUI SPPETlVISOIVDATK
- 12) PLANTTKCttNICALlEAHAGKRRE~ RKOVlfLU)
YES NO
- 12) PLANT TECHNICAL NANAGKIVDATE
- 14) POC RKOCSIRED YKS NO IITG. NO.
PROPOSED INPLKNEHTATIOtl DATE 1S) PIIR APPROVED TES HO PLANT NAIIAGKILYIATE
- 14) IrWR NO{5}
YES NO DF YES L)ST OH CONTINUATIONSHEET) 1S) 4HSTALLATIOHAND STATIC YXSTMG COI4PLETE PLANT SYSTEIl EttG/NEER/DATE If) CONTROL ROON OPERATING PROCEDURE ANDTOP TIER DRAWIHGS UPDATED. QODIFICATIOHCO54"LETE, YSTKN OPERAbVdTY TKSTMG CONPLETED.
SHIFT NA/tACER/DATE iMREVISED. PLANTDESIGN DOCUNEHTS UPDATED AHD PROCEDURE REVISIONS ltlITlATED.PTL ~TED. IIEL~ i SHEETS TIIJJcS~iED AND QEL UPDATED.
)
214) COi/TIIRIATIONSIIEKT P~i ADA/NISTRATIYKNANAGKIL'DAK
II~ '>> I'I IIIII I IIIII II 4
~ ~ rA tev rerrl>>'I ~ Sr>>reer Ie>>eeerrr>>eo
~Irer r~
1Kl.
shot thor'.
t 7 SYS'1'liP1 pg-/SF r
IJ Edd n uerv I',I'N aud lutorfnnllon to htl'.I, M
hlodlly eilsllffg I',I'N tnforfffntlon nn JltI:L Q
l)clcle estsf lug I.I'N frofn htl',L (cqfflpfncnt no longer exists tff the ptfffft)
U l)clrte cxlsflug I.I'N fsoshf I:ItJSIIht flchls (cqffttuucnt no'longer safety-rctatcdh but still extsts Note: I'lace a "E" ln field No. 30 and llchls 31 Ihrongh 45 niff) Ihe'ifotntton "I)clcte fflldate ln llchls 30 through 45")
Nt)TIB:
I. I'tace n P syfubol ln auy tlcld Ihnl Is not aplhllcatfte. I,eave btantf If data b unavailable.
- 2. Inslructtons fot confplcllng lhe nuufbered tlchls belovr are glvcn ln Et 2.3S.
.Adr'$:-
draft a line through
+r A AFFECT?.D Pr 6, 5'FE Qf E or &
Or rgrS PC4
- l. 1:.I'
- 2. NSSS
- 3. SYS
- 5. htl'OI
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Ii. CVI4~
lh. Ctt<<r/EXr
)0. EI.Ehl 2.I. FI'I LI
- 26. I:I'2
- 6. htOI)EL
- 17. CONN/ASht
- 21. SEIS~
2I. IIANUALLLj
- 27. CErH
- 29. I'Sl
- 22. CLEANLJ I
LLJ.
LOC. I)ETAIL IO. SLI/O U II. ELEY/ZONE
- 13. rIVII/OUT Ih. YENI)
Is.qC4 IS. IEEELI I II.COUELL
>5. ISIOIQ
.IO. I IAI;tC,~,/y,t)II q> U 3t. CON t IIA(:r
- 35. IIOUIISl~~
- 36. SAIIETYFUNCTION ~
SL LEYEL~
II.Equir CLASS+
- 31. ACCUIIACY
- 43. 'Iht II.ANALYSIS~
II.YIIEO~
- 45. QIU
'Ãe/l.frffae fl? f10/hyf UAra y
7
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49 SUPPLY SYSTEM D~P PLATE
- 3. DES RIPTION OF CHANCE gPtV S 8FEFCFED Pf C'484r Zg de PHJ-. E 015 0+ i Hi'S PC'P iAJ Ech-SP~-io pog-spa ll goy-SPV-fP Roy- ~~v gp~ -st'V->V R'D ~-sF'~-~~
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a ~ sUpsRscQE$ pREvlovsLY 4 ppRov ro DCP Page:
IDCP Pag:
f'P t39O IId8 rib oe'i>es IDrawing / Document No.:
I
%IO
&6c M999 l6-8',
i l*c Drawing Ih CHE KS SY 2one.'
CVi Dccumen:
Rerision:
SCA' i I QRAWIOln 0I ~ va
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- WhSHING'TO'N PV IC PQWEX 43 Sue VSTEM P.Q Box 968
~ Richland, WA 99352 - 0968 P.O, DA G2 (20/P."I A('I F'>
URCH E ORDER ND.
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uomlt Intoices in oupircere to:
Attn: ACCOUNTS PAYABLE MaDa DES-~,
P.O. Box SSB, Richrenth WA'B~I D968 va'I
~
I NO EXCESS OUANTITIcES OR ADVANCE SHIPMEh.S OTHER THAN
'HOSE AU i HOREEO WILLBE ACCEPTED.
~ I I
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PURCHASE ORDER NUMBER 'MUS i APP" AR ON ALL INVOICES,
.,PACKAGES DOCUMENTS AND CORRESPONDENC PA.~
a I
BUYERS REPRESENTATIVE 1%5'n scheciuie oueiity oetf otttlence is not e
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tenuiteftlefrt
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~ Richland, WA 99352 - 0968
<, DA
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ORIGINAL ORDER CONFIRMING ORDER ICHANGE e ORDER Cj(Ci:.":.L"l;".."..FiihT LP(eriii 1:i'TTEHTIOHe an A 9B '02 SHlr's rise t&saU r t/la arUr /
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/(iS: / tta UI PL ie" /lie /l ~
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Submn invoices ln oupllcate to:
Attn: ACCOUNTS PAYABLE M.D. 055 P.O. Box 968. RichiantL WA, 99~~2-0966 NO EXCESS OUANTI ~ IES OR ADVANCE SHIPMEN aS 0
> H R THAN THOSE AUTHORI2ED WILLBE ACCEPTED.
PURCHASE ORDER NUMBER MUST APP" AR ON ALL INVOICES, PACKAGES, DOCUM"N s S AND CORRESPONDENC 10(P/ on scheduf ~ quality oerf ormanoe is not ~
oesirable goal it is a requirement BUYERS REPRESEIeTATIV:
WASHING. ON PUBLIC POaao SUPPLY SYSTE.'66-15551 Io~I PAGE 4 PUR HASING Fl'"
I'l
'AEHrN'GTON FG
.1C RDWERt
'6IIPS. SUPPLY SYSTEM.-
P.O Box 968
~ Richland, O'A 99352 - 0968 G2/
P P.l APL6'<
RCI5 NO Q($ OC'P r
-: "PUB~ASS 'ORDER PACES ERMS
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IA i voices in duohcete tol Attn: ACCOUNTS PAYABLE M.D. 055 P.O. Box 968. RichlencL WA, 99352 - 0968 t40 EXCESS QUANTITIES OR ADVANCE SHIPME/ETS OTHER THAN THOSE AUTHORIZEDWILLBE ACCEPT"D.
PURCHASE ORDER NUMBER MUST APPEAR ON ALL INVOICES, PA KAGoS DO Uh cN S AND CORRcSPO!4Dct4CG 100K on scnedul ~ awlity perforinena is not e oesireble yoel it is e reouirernent.
BUYERS REPRESENTATIVE WASHINGTON PUBLIC POWE SUPPLY SYSTEM
&So t856t f744) PAGE C PEJRQHASING 15 E
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CIA}.INSTRUCTIONS:
Suotntt tnros(sat in ouolsoata tot Attn: ACCOUNTS PAYABLE M.D. 055 P.D. Bos( QSR, RIO}stand, WA QQ252
~ OQ58 tesv EXCESS OUANTITIES OR AOVAN E SHIPMENTS OTHER THAN THOSE AVTHORIZED WILLBE ACCEPTED>>
PVRCHAS ORDEP.
NUMBER MVSi APPEAR ON ALL INVOICES>
PA KAGES, DO UMENTS AND CORRESPONDENCE.
100'n aetaduta ouaIItY oattottnanoa
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tpouttatnant BUYERS REPRESEhTA t lVE 7se' S M I }e G T0 }e P I' LI 0 P 0 ees/ S I SUPPLY SYST'~ V.
QBQ-1B551>> I/~I PAGE ste PURCHASING."-It E
'W AS H 1'N O'T OT>l P 4 1 C P 0 W B R>.
SUPPLY SYSTEM.
P.Q BDx 968
~ Richland, WA 99352
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IALIhlS 1 sn auol>cere ta:
Attn: AC OUNTS PAYABLE M.D. 055.
P.O, Bor 966, Riahiand, WA II353 ~ OII68 N
EXCESS OUANT>ITIES OR ADVANCE SHIPMENTS OTHER THAN THOSE AU HORI2ED~WILLBE A EPTED.
PURCHASE ORDER NUMBER h<UE I APPEAR ON ALL INVOICES>
PACKAGES, DOCUMENTS AND ORRESPONDENCE.
100K on aanedule auailty perrormanoe it not a
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.IAI. INSTRU T In OP ICete to, Attn: ACCDUNTS PAYABLE M.D, 056
- P.O. Bax 968. Rilchbgnd. WA 99352
~ 0968
~
1 NQ EXCESS QUANTITIES OR ADVANCE SHIPMEh i S OTHER THAN THOSE AUTHQRI2EDWILLBE A EP ED..
PURCHASE ORDER NUMBER MUST APPEAR ON ALL INVQI EcS.
PACKAGES. DQCUMEhTS AND CORRESPONDENCE.
10tXC on acheOLIIe CMellty oegtoglnenoe is not e oeairable yoel
}tls e legs U I gegn cot BUYERS REPRESENTATIVE WASHING t ON PCHI.IC'POV'E SUPPLY SYSTEM'66-18561 f7~I PAGE tl
,PURCHASING PILE
%'AS HIN CT0% PU;1 C P Q t<r)/E 8 4B SUPPLY SYSTEM P.O Box 968 o Riband, WA 99352- 0968
---PURC';IASE ORDER Po g /hh /pg, G7> o~t'I P r1/,Il()
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CHaNGE:
I oices in ouohcste to:
.1 Attn( ACCOUNTS PAYABI.E M.D. 055 P.O. Box 966. Rich4nd, WA 99352
~ 0&68 NO EXCESS QUANTITIKS OR ADVANC"SHIPMENTS OTHER THAN THOSE AUTHORIZH)WILLBE ACCEP i ED PURCHASE ORD" R NUMBER MUSi APPEAR ON ALI. INVOI cS, PACKAGES. DQvUMKNTSAND CORRESPONDENCE.
1ONC on ache(fufe ouallty perf ormance is not a oeairaOI ~ aoaf lt is a reouirernerlt BUYERS'REPRESENT I ATlYE WhSH II(6
> ON PUB LI C PO%>f E SUPP';Z SYSTEM 658-16561 f74<I PAGE a PUR HOSING.-fLE
..: 43', SUPPLY SYSTEM fQa A L
~
u
() '/20/9 PURCtt E, ROE'G...
r PU'R<.EASE ORDER P.O Box 968 ~ Richland, WA 99852 - 0968
$.06PQ 07lo5G etc Ms 1'.-
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'NSPECTION REQ'D OQA R.I.P. RE Q'D
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CIALINSTRUCTIONS:
Sublnit inroicns in auplicsts to:
Attn: AC OUNTS PAYABI E W). 055 P.O. Box 868. Richland. WA 96352
~ 0968 Nlu cXCESS QUANTITIES OR ADVANCE SHIPMENTS OTHER THAN THOSE AVa HORIZED WII.LBE ACCEPTED s*
PURCHASE ORDER NUMBER MUS APPEAR ON A'
'INVOI S,
PACKAGES, DOCUMENTS AND CORRESPONDENCE.,
100)'n scneduio ousllty Oertormsncn is not s oesatsb4 Icosi -tt is s tecuirelnena BUYERS REPRESENTATIVE WASHlta CTOh PUBLIC PO REF.
SUPPLY SYSTEM 56-t856t I'1~I PAGE C PURCHASING FI' 1
C
~
KhSH1NGTON FZ;1C FOIBLE&'W LN SUPPLY SYSTEM P.O Box 968
~ Richland, WA 99352 - 0968 I'0>
L 02 I )/PC, Pt'I OC>
ND, 071 <~yC
- PURCHASE ORDEB Ca rAQCS crtecs 0
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.. CONFIRMING ORDER ORDER CHANGE ORDER cite r a.. ~ >A J. CI>> la> 1
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L INSTRUCTIONSI Sttomit invoices in oooiicata tot Attn: ACCOUNTS PAYABLE M.D. 055
'.O.
Box 968. Richlencc WA 99352 - 0968 NO EXCESS QUANTITIES OR ADVANCE SHIPMENTS OTHER THAN THOSE AUi HORIZED WILLBE ACCEP i ED.
100~ on scneotc4 owlity perl olrnence ls not e clesirable peel it is e leooiretnent BUYERS REPRESENi ATlVH ss'ASH IN CTON PUBLIC POwEi SUPPLY SYSTEV.
W>85v
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FLORHAM PARK. NEW JERSEY 07932
~ N~.-f2oiloaa-oooo'r.v.-faaf w-seas Manufa rural of DEPENDA8LE CONTROL Sinca 1888 CERTIFICATE OF COMPLIANC CiCV "+gk<< ~k~g
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June 42.
1985 107053 Date Customer P.O. No.
Wi%
ASCO Shop Order No.
6 459~
ASCO Pa< No,
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Quamizy-
'I 3 Consi ee: 4ashin ton Public Pave" Suoolv System Consirnee P.O.
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07" 950 This is to certify tha. the work has been completed in accordance with the requirements on the purchase order and referenced arawings.
We further certify that the material has been manu actured free of mercury contamination.
Pzte=ial supplied meets o>> exceeds tne auality zeauizements established by the zefe=ences, spec~rications in the above pu=cnase o de".
"riri '
r'icu R.l. Inspector Dat.
S i AT" OF NEW JERSEY)
COUh i Y OF MORRIS
) *'UTOMATIC SWITCH COViPANY Company SWORN TO AND SUBSCRIBED BEFORE ViE THiS >'-='" n*VO.= ~"'"',>a iii.
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FLORHAM PARK, NEW JERSEY 07832
~ Ply. [2chl gree 2c00/N.Y. l212) & 375$
CERTIFICATE OF COMPLIANCE Customer Name BLACKS-INDUSTRIAL, IN Customer P 0 N.o.
Consignee RASHINGTOH PUBLIC POVZR Consignee P.O. No, 07t o ASCO Shop Order No.
644>~N ASCO Par No.
NP83MA" 72E Voltage
).20 0
Quantity Eng. 3ob No.
This is to certify that the subject va)ve(s) meet the performanc requiments ofI EE-323-1974, IEEE-344-1975, IEEE-382-1980 (Revision ofIEEE-382-1972) and IEEE-627-1980, as substantiated by testing valves ofgenerical)y equal design in accordance v, ith ASCO Qualification Specification AQS-2) 680/Rev. C, dated 3u)y 13, 1981, The fol)owing test levels were included in this qua)ificauon test program:
I. Aging Simulation Phases:
A. Thermal Aging Simulation - 250'F for 18'l4 days. These aging parameters were determined by Arrhenius calculations to simulate a minimum of 8 yean in a 140'F continuous zmbient. Refer to Figure 1 for additional information regarding service periods for elastomeric components and Figure 2 for additiona) information regarding service periods for solenoid coi)s.
B. Wear Aging Simu)zt)on - 20,000 operations at mzximum operating pressure differential and nominal voltage. Ten percent ofthe v, ear aging simu)ation:(2,000 cyc)es) was conducted concurrently with the th rmal aging simu)ation.
C. Pressurization Aging Simulation-15 ambient pressure excursions from atmospneric pressure to 80psig to simulate the expected periodic pressure zation ofthe containment forleak t sting. during the lifeofthe n)ant.
~
Raoiation Aging Simulation - 20 m gzrads of gamma radiation at z rate not exceeding 1 me arad p r hour to simu)ate expected non-accidem radiation exposure.
Vibration Aging Simulztioni - Continuo s sinusoidal sweeps from 5 to 200 to 5 Hz at a rate of2 octaves per minute. with a minimum peal; acceleration level of 0.75g (except at low frequencies where the acce)eration level ivzs7educed puch that tne displacement did not exceed 0.025'ouble amp))tudor i. for, z minimum of90 minutes'in each ofthree orthogonal. axis. The test valves were alternately de-energized or energized even 15 minutes duHng this exposure. Th va)ves were attacned to tne shaker table by rigid test t>xtures using the standard vzlve mounting provisions iviththe solenoids (Solenoid 'A forNP8323 va) ves) verucal znd upright. Flexible hoses were used on all ports: therefore. tne set-up did not a~iect the rigidity or mass of the va)t es being tested.
Seismic Apng (OBE) Sim )ation and Resonznce Testing - The valves vt ere mount d to tne shaker tab)e as described for the vibration aging simu)ation and wer exposed to two sinusoidal sweeps from 1 ta 35 to 1 Hz.. with z peak acceleration level (witnin machine
)imits)of3g. in each ofthree omogona) axes at a rzte of not more than
) octave per minut
. One sweep in each axis wzs conducted with the valves energized and th.other v )th the i'sa) v'es de-energized. These sinusoidal swee ps are corsidered 3o provide the equivalent dynamic efiect of 5 OBE's. During tnis t sting. acce)eroaeters were zttacned to tne soienoids oftn test vz)vesto det rmine iftne vaives exhibited any resonance. Resonan" is destned as a r sponse gAgith a wz~mitude of acc )eration at least twi" as great as the input acceieration. No valve r sonances were detected.
ASCO SHOP ORDER NO.-
II. Design Basis Event (DBE) Phases:
A. Seismic DBE (SSE) Simulation - The valves were mounted to the shaker table is described for the vibration aging simulation and were exposed to a series ofsingle-frequency, single-axis sine-b at tests at 37 test frequencies between 1 and 35 Hz. The excitation was in the form of a continuous series ofsine beats, with 12-15 oscillations per beat, for a minimum duration of 15 seconds at each test frequency.
The successive beats were phased such that any supeiposition of response motion was additive. At each test frequency, the peak input acceleration was increased (up to 15g maximum) and the g-levels were recorded at which the cylinder port pressure (zero when de-energized and fullinlet pressure when energized for a norma))y closed valve. opposite for a normally open valve) differed from the nominal by 09o, 59o and 109o ofinlet pressure.'The valves are considered to function proper)y up to a 109o change in cylinder port pressure. This )eve) was selected as being sufiiciently lowto prevent spurious shifting of the customer's main valve or other equipment.
Motion was applied at the same frequency and acceleration limits in each of the three orthogonal axes separately.
Based on this testing and/or additional testing conducted by ASCO usin single-frequency continuous sinusoidal inputs (after con-sideration ofmargin as suggested in IEEE-323-1974), the followingacceptable maximum acceleration levels have been determined:
9.0g
~,g
~ sa B. Radiation DB" Simulation - 180 megzrads oi gamma raoiation a'. a rat not exceeoing 1 m garad per hour to simulato (after consioerztion ofmargin as sugg sted in IEEE-3 3-1974) at )east 163 megarads of ac "iden'. radiation exposure.
C. Environmental DBE Simulation - The va)ves were installed in a pressur vessel znd suojected to a 30-dzy exposure to stezn:. chemical spray znd clear water spray simulzting a corn'oined )oss-oi-coo)znt accident/high-energy'-line-break event and post event cool-down. Tne peak ambient temperature oftne simulation was 420 r and the peak ambient pressure N zs 70 psig,. The valves were pressurized to r."axi-mum operating pressure and continuously en rgiz d for 4 nou.s prior to the tirst transient (to produc tnermal saturation of the solenoid coils). They were de-energized when the temperature of tne first transient reached 420'F (to demonstrate th a'oility to perform a typical safety function) and wer normally de-energized out were cycled periodicz))y during the 30-day exposure to demonstrate the abilityto operate on demandi. The oualiried temperzture profile demonstrated by tnis simulation (after consideration ofmargin as sugg sted in I:"EE-323-1974) is shown in."igure 3.
Tesi Report AQR-67368 is on file at Automatic Su itch Compzny in F)orhzm Park. Y.J., and is avaiizble for customer perusal.
Dated.
~Jl~=, 1-. '85....,
Autnorizec Signature
0
40 A:
30 26 Q0 20 WO
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GO NP-1 VAI.VEbSIIOULl)IIL flLI3UILTUSING I'IIL.
APPI3OPI3IA I L. SPAHI= PAATS Wl IENEVEI3 I TED I3Y TlIE I'EfllODICINSI'ECTION OF VALVE NENTS OI3 WlIENEVEII'ANY6F TIIE FOLLOWING I.E LS SIMLILATLDDUIIING'YlUAI.II.ICATIONTES I ING, AAE IIEACIIED:
1.
WEAI3 ACING 20,000 CYCLES 2.
IIADIATIONAGINC 2 x 10 IIAD TlIEIIMAI.AGING TlIE MAXIMIJM SEI3VICE PEI3IOD INDICATED FOB TIIE APPI.ICAI3LE SEA VICE AMI3IENTTEMPEAATUI3E.
P
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(122" F)
SEI3VICE AMBIENTTEMPEI3 ATUAE Gn"- C (140" F) 70~C-(168 F)
. RGUBE
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FOIIIi V6 3239A3
6a 4a 30 IX 25 Q
20 0
Llj 15 ul0 la 0
D 8
X 7
NOTE'N OfIDEATO MAINTAINQUALIFICATION,CATAI.OG NP-1 VALVESSIIOULD OE fIEDUILTUSING Tl!E APPfIOPfIIATESPAI)E PAATS WIIENEVEfIINDICATED OY TIIE I'EfIIODICINSI'ECTION OF VALVECOMI'0-NENTS OA WlIENEVEfIANYOF TIIE FOI.I.OWING LEVELS, SIMIjLATEDDUfIINGQUAI.IFICATION TESTING, AflE I1EACIIED:
l.
WCAA AGING 20,000 CYCLES 2.
fIADIATIONAGING 2 x 'la~ fIAD 3.
TlIEfIMALAGING TlIE MAXIMUM SEfIVICE PERIOD INDICATEDFOfI TlIE APPLICABLE SEI)VICE AMBIENT TEMP E AATUI1E 20'C (68 F) 30' (86~ F) aa'c (104 Fl 60"c (122" f-')
60'C (140" F)
Va'C.'158')
SEAVICE AMBIENTTEMPEfIATI)fIE FIGURE 2 MAXIMUMSERVICE PERIODS FOB SOI ENQID COILS IN ASCO CATALOG NP-I VALVES Form V till
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NATCODE DETAILED DATA.
ACTION 8 PRINTER QQP HATCODE: 54501580 VALVE SOLENOID 3-MAY 1/4" X 1/ib" OAFICE WITI( 120VAC COIL HATERT It'NT ENCLOSURE ASCO NP8320A172E CATEGORY CODE OB STATUS CODE A
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TECHNICALSUPERVLSOR/DATE 6
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- 12) PLANTTECHHICAL, MANAGER REVIEW REQUIRED YES NO
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- 15) PMR APPROVED YES HO PRIORITY PLANT MANAGER/DATE
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- 14) INSTALLATIONAND STATIC TESTING COMPLETE PLANT SYSTEM EHGtNEERlDATE
- 19) CONTROL ROOM OPERATING PROCEDURE AND TOP TIER DRAWINGS LIPDATED. LIODIFICATIOHCOMPLETE.
SYSTEM QPERASII.ITY TESTING COMPLETED.
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To:
K. D.
Cowan - 988U 0
~I FROM:
L. T. Harrold -
94 sag~:
iNGIN"":"RING CAPITAL WORK ORD:-R 0 ALUATION FOR PMR 86 0156
{R""PLACE ROA-SPO-10"15, 17)
DISTRIBUTION: DIAILDROP
'NP.I FKZ
<~ WNP 3 FILE WNPr3 FILE
~Q WNPw FILE WNP.3 FILE
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~ ADMIaai FiLE In accordance with he Capital PMR Work Order approval
- process,
.he subject PMR has been evaluated and is disposi.ioned as ollows:
x
- s ense
=or implementation viill ha 1
ss han 510,000 Drawing Change/
Maintain Configure:ion Control only.
Further approval for enoineerino to proceel is not required.
Genera-ion Engineering is continuing wi"h preparation of a Design Change Package (DCP).
XXX. r pense for implemenztion will be less than S10,000 may require material/,
, labor expenditures.
rurther approval "or engineering to proceed is not
.,required.
Generation ingineering is continuing with evalua.ion and vill.
disposi4ion
.he PtR with appropriate documenation DCP, memo, etc.
B<pense for implenentation will be grea er.han S10,000 "=naineerina es imate attached.
Further CWO approval required.
=ngineering is no-proceeding with preparation of a Design Change Package
{DCP) until an approved work order is returned (a desian need da e will have to be re-established at that time).
Other-JGT:ch
Attachment:
as stated cr.:
wo/a:-
=ngrg i:;gr -
NS Porter - 981C
<ngineer T. Miles - 981C Tech SQ
="nor -
Kippes-988U preoarer -
(Tellefson)
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- 994:"
Wtt Waddel - 982B w/at RJ Barbee
"-P Divincenzo RL Koenigs
- 5. Scamon
.=. Walten M. Wuestefeld PMR Coordinator
~'lR - le
- 9881J 1C"5 988U 988U 9275 988U 98BU r ~
C.sss
{Cowan'3 Copy)
Desi~ ~~~e Pamage A.
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Xndus~ial Sat eely review......
6.
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CaapLete da fc~l'ng cu s~ns:
<<~
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Te tnica2. S~ ioatian Mange recu ~
(if yes, a ~ na~)....
~
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6.
Soe:~ tmwra ~go'~
('t yes, a mW a l's ard/oe cLs~m"~on) 7.
Any P~ ~M~s a Con~i Ream Cocumnw a~ }e~mf (a aN blent Mv).
8.
Natal ~l~~ ta te pu=<<has&
(ii yes, a~>zN '~sv).......
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A mm To PhR 5<54.ail ~ m~/Da A
a"linen" 2 Pag 2 o 2
~O~UKC NUMOCR 1
1
<<o<<o<<
wmssh R7 (~}
RCVlSJOI1 NUMOCR PACK }tUMOCR t
t(} <<p C
<<o <<o<<<<M vl
I 0
'I
~
~
Rev'an p/0 Ivy t ~l Ream comms N Plant m"af~ ass" be nad'~ ~a-w Jm e ~
GKft~ cz'st l't be'g E~icct aA) ~~~en B.
Xns an.rrt Has" D-~ Sh C.
VexfCam~~ W Ha'n"~m rLanum D
Has-
- au per L.s oM c'cLi.~~
r AfTi. SVSi%>>i%~Re
~e~e A a~
1.
What does modi ica.ion accomplish:
/
s8 t7 Qp'CMlcrcr s
F I SC v~s PIP BY-~~
0':-~
2.
Description o
modi icz.ior.: (enter "same" if descrip.ion on PhtR is correc.
and ful adequate):
I,~~ pv-lc Ji x
I~
r 3.
Specific just-'ca".ion for modification: (e.g. reliabili.y,. e ficiency, labor/cos
- savings, safety, regulatory, et").
Cost-benefit statement::
SC'.
4.
Safety significance (include
~ech.
spec.
involved or affected):
r s-Z tlodification a, Feet on system/plant:
a.
plant outage required:
Yes Qtlol h.
ysst er owuge reqrired:
~es no c.
other systsnLs affected:
Yes
~to d.
special limitations requiring at en ion to Pr Hw~ o4 (if "yes" iden.ify)
(iden i y) perform modi ication:
e.
integrate/work this RS with o her related work:
6.
Order of Magnitude Cost "=stimate:
Design-hf 1/:qpt-Ensulla ion-Total -
c')0 ewe Caoi ul iiork Order Iter,",:
Yes 'JrO CMO 8-Schedule info: (include procure@ n., ins:allation):
Recarmended plant modificzt".'on implemen~tion date:
(bzsis) v
Reference:
o relat d docum n-~:
c&/poR:
i
~ ~
~ ~\\ ~
O ner:
S'>
N'kR:
P P
4p sUpp6 HsrzM 26-Ql z&-OP, DESIGN CHANGE PACKAGE i>> REASON FOR OCF 7HZ PlJRPCISE, Ã "iiti5 Di~F'S TO PROVIDE E)IICIJERINC DIRECi IQQ TO 8" MOUE EXI5TIM& (QOQ.CIUAUFIED )'&~0 MODELQo. HBXG>ZDAI 1 BI ) SOLEUOI05 VALVE,&
RQFt-SPV>>e IOI I I I IZ(GI I II I 5 41'7 AQD Fc'.PLACE WITH QUALIFIt D (<4SCQ MODEL ISIo. k)F'ESZOAI'72E ) VA'r=S.
THIS 15 Mi=CM&RYTO t)KFADt THE EQUIPhlt":h) i FROM "PASSIUE " TQ "ACTIUE"STATUS.
R':-r I:="M =: k)CR'I)O. ZS6-ISB Se OESCRIFTION Of WORK SFCCIAa OIS 8 RISUTION i.L.Mll~5 i/D 95I~
~ F geg RIFC R N CgJSO U RCL 25-0156 0 SUSIE EC i 50t ""QQID t/ALV:" iKFLhC=MENT Se LOCATION i'OP ELD&. VARICIOS SYSTEM NO>>
8I.O R.. EL'. =" Vil HUA~
~ ~ QumTYC~ T E Field work required Rt=rEJ>>,'TO PAloE
+~'t OF 1Hlcu Dgf fQF>>
lAlSTirUCTIQMQ.
Q Drawing chsoye only Fc-CICIZv-QO-CI65 Q 'I/oi*DC':
vsoRK oN TNIs ocf ssgoULo ec cooRQINph B5-002Z.QD 4>>
IS OCF OEFENOS ON f'RIOR INSTAI>>uhoATIONOf OCF
'a'-'"-""<<-'"l0-'DKQQk;CXANKPACXAGE'WDEXIANDlPAGENJMBERlNG;SEQUENCE DCP Approval Form:
10CFRSLE9 Safety Evaluation:
DCS input Sheets:
DCP Plates.
Pa9e 001 Pa9e 002 MEL input Dace Sheetst IrtrtalletiontTest Req'Inta:.
ASME XI Section XI PQL New Drawings:
Th>>ough Th>>ough PSagt/Tech Spec Chg P>>op>>oah Thtaugh New Docunlents:
Ptocu>>e>>ooo SpeoT>>oet>>o>>>>o Th>>el>>gh Billof Material:
D 9
Desiyn Backup/Rev. Appt Th>>ough hogg RA N/R.
~
~ IO TT DERQS REM L COOL COMFaekANCL v"
I'Sic 'r/.R~ /<-"-
C FR A
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TECH. SPE REFEREN 5
SECTION I
PAGE
>>Do<< this DE>>uen Chan9E.
Proceoure Revision, and/or Soecsal T<<t consv.
tuse a
cnanse as oeacnoed in tne Final Saresy Anaiyas Report?
I YES M~NO R REFERENCE VOLUME I SECT>>ION I
PAGE I
I I
I
~os Adore<<ed in Tech. Specs.
I lYES WNO
'ls
~
cnanea in
> ecnnlcal Soecill ~
cioons ineoieed?
I I
I i
@Not Adorecaed in FSAR IF YES IS NO UNREVIEWED SAFETY OUESTION EVALUATION:
Answer tne followinII ouesoons with s "yes" or "no", and proeioa soeciRic reasons iusdfyinIItne occision:
A Can tne probability of occurrence or tne conseouences of an accident or mattuncoon of eouiplnent unoortant to safety previously eeaiuated in tne safety analysis report be incraasa8
@YES tPNO gecausal Iaat OKCIGs'HhstGF trf>>Ch Doc 50 LUCIO VALVGI>> FMIw Ph55 ISI E IC "Aa iuE" STATIlS K PI,ACE'INOIOT 'AITR ~ QIIPILIFIECI UaIITS
~
Complete Stoctc No. 8 ot this torm Prooaeo to Stooc No>> 7 L Can a possibility for an acsooent or malfuncoon of a difIKeil, type than, any eoaiuated preeiousiy in tne satety analysis reoort be creetecg
@YES
~O Because; SHOIIH F~SULT Iht IMFRDIIo D I l~>>Y.
Reowst and receive Nuclear Ree uiatory Commission autnorszaoon tor chance onor to lmolemeo sasloh ot sll~
luotec.
cnanoe eser so ls tne marIF'n of safety as oelined in tne oasis tor any tecrmicsl spec.
Ificacion reowacfl ROT C>>PK'Cl i-l(ALLYAODF-<~ E D tM H. c FI=<.
ISA>>S>>CII I'yES I
>>AIIAnswers ISA>>S>>CI NO T
AI.HORI2>> ION IIECSIVED Iemese>>>>ee>>en ~,
>loeeeel 1 Ctu>>e<<aeei<<aeecse>>
Te<<
e>>eclat>>ee tt~ ln aloa No. 4 is YES. o h olE ctlol+
is reoortsow unoer;C FRSLSEO ano o<<criouon ot tne'mosqo willoe inciuoeo in tne Annwi Repen.
The inoielouai inioanne sne Design ~OE. Pro-caoure RerlQon, anclor Soecsal Tees ls rrsoonsiole tor suomissing F SAR chances to ow Plant ~cenung Mana<<c>>
aoe ~ >>o>>e
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f.
RZMQVE EXI5TIQC SOLENOID VALtJKS (A5CQ H"EB3Z~AI 5EJ) AQD F.'-TUFhJ TO VIihRE HQLSE A5 $1'AR&t.. 2""TAIlU tlhRDlVAIC!EFM Ljtt "R U5c.
2, FABRICATE 5:-V"4 (7) "Uh)IVEPSAL MOUQT)hl& BIPACK"T/ ADAP t =rJ.'5 ITS SHONM OM PACE OD~
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DRILL MbOhlTIhl6 BQLEKI IhJ L<IDE OF SKhCK:"TS TO NIAKH ZXETIhJC MPPQRT IHAhl&~I'QLr CQUF I60kAIIOh) rrRTtt.'= IItJOIUIDUAL SOL~M&ID VALVF Ihl5>ALLAt IOAK.
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S. RETFRMIh)ATE WIaltu&. -tG THE los-IPI SOLEMDID VALVE5 18 ACCbROAQCE Milt H FPH JD. Z'5.19.
- 6. KECMAlcCT TUBIhJ & 'TO THE'QEMf SOLrIIJOID VALVE'SI US)M& SNh~MK F'RACTIDAJAlRME TO "'t"QF i COh)h)ECTQK5 r COIvJFREKIQIO TUEIIIJ& UILIJOh)S Ah)0 OTHER SWAGLOK QUAL) t Y ClA5S r
COIvIPR'RZIQh)S J=ITtlh)&5 R5 NECESSARY, SEE EXAMPLE DN FA5E OOE OF 7HIS DCP.
7:
PERFORM VISIIAL INSPc~IOIJi U!IRIN& CHECKS AND OF:RAiIOh.iALl FIIIICYIINALYESY AS INDICATED OIII FA&E DI~
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0 Add a new EPN and Information to MEI.
- I MEl INPUT MTSIIEET EI 2.36 nct tAoE rto 86-0156 -
h = Ota sUtEllsEDEs toEvloUst.Y Attnovf0 OCt tAOES t8 Modifyextstlng Ef'N lnforrnatlon on MEI.
tl Delete exhtlng EPN from MEL fequlpmsnt no longer exists fn the plant)
~
~
I I Illsnh nut CI6/SRM lie'lde Iequfpm ant no longer safety related, but still exlstsl Note: Pface ~ "¹ " symbol or a 0 tn Field No. 30:
See El 2 36 Instructions for F Iaid No. 30
~ '
I Tag Change Only:. Enter former EPN here 2
Enter new EPN In Field ¹ I OENEIIALINSTRUCTIONS t.
Place a ¹ symbol In any field that ls not appllcaMe 2,. Place a symbol ln any'field fot whfcfr data ts to b dilated 6.
Show numeric sero as Jf, and alphabetic "nh". as 0, to avoid confusing the computer.
3.
I.eave field blank lldata Is to be entered, but ls currently unavailable 8.
left.hend fustily all entr les.
Colnplet ~ the numlrered tl~'lds befow per El 2.36 0:-."00':LF;--,.":;".-w::"':.:,.YIe'CMMrOLIOIOSew~Y.':;!<<I,':0/IECY'>>.':;=.:;:-':A3.'.OE O I
'.INrN
. ATE iL'E5T:-.EF: 03"'!"O>>D:;.'.'.::IT,EL.Sail'.,,",.N ~-;:III.II:'L."../:.I: /.-.'.2/-
I.EFNL22-pl~on 5~f'v -
I
.4. DESC
- 2. NSSS
- 6. hlFO
- 7. S/N Q. REF IjWO/ZONE 12.CVI
- 18. CONT/EXP 20.ELEM LLJ I
- 23. FPI LI 20.FFSLL>> ~
2/I. PS1 L2I-S.MDDEL [NNP8 Z O n l l Z Ei I. I.OC. DETAII.
/
- 10. OLDS LLj II.ELEVIZONE
/
L~ IS.FTTDIDDT
- 17. CONN/ASht 2l. SEIS LLJ
- 22. CLEAN IJ
- 24. EOUIP DESCll, ~
- 26. PSTOT Lj 2l. CEPN
- 20. PS2 2 I~. VEND~~
IS. OC~
- 18. IEEE LLj 1L CODE LLJ
""FSg'"FNJY&0'2")""Ti %co,'INS Int::-:pt'TFGI'"r"'"'L'f>>lsM's ~ -TSI"S"CS>>t"O'
/Sf1 IE ON
.,t'F>>~~OF&'l:S.'A4R.'Cist.'N.; I '~~I.S.s."0 t.>> LIM~">>~.cget>>.b>>4~".>>02~~%4>>02.".'~.L'/S qo. FI.Ao (c,s.¹,OR o) LiJ
- 31. coNT RACT
- %. IIOIIIIS~JSS. SAFETY FIINCTIONQ~
S/ISMIC OIIAlr 39. IIYDRO I.OADSLj>>IO.TEST lJ
>>12:IIUAI.IF ICATION STATUS - SEISh'IIC Lj ENV LI
- 32. I.EVELLj
- 33. EQUIP CI.ASS IJ
- 46. OID
- 31. ACCURACY Lj Q
- 41. ANALYSIS
- 03. TM Q SS. FIIEO I
- 34. USE IJ Ll tIIEPAflEUBY lish Ill1th 04 tthlhl DAlE nt Rtc fo, i>8s
~ Mh! ~ 3 ~ 9 sr
~ 1111 ~ ~
~:.>
'I Y. SYSTl'.M rtlECII OtiEs l
1 Add ~ naw El'NandlnformatlontohlEL MKLINPUI' r~1IEEy El 2.35 Dcp PArtg No.
BE-Ql5 I'o.Ol -all SDs'KttSEDgg PitfVlOUSt.y ArI'IlDVED DCP PArtgr 8l htl>ltity axlstlng EPN tnformallon onhlEL t ) I)~I ~ I~ exhllng EPN from MEI. (equipment no longer exIsts In lhe ptantl l ) tllants o<<l CIE/Snhl flclds (equipment no tonqsr safely rataled,bul still exlslsl Nol~: place ~ "g "symbol or ~ 0tn Ftatd No.30:
Sea El 235 Instructions for Ftetd tto.30 l..l 1ap Change Only; Knlar tormsr EPN har ~
2 t
~
Enter new El'tJ ln Ftsfd ff 1
\\
OENEUAL IttsfflUC1tONS '.
Place a ffsynshot fn any field that ls not eppllcaMe 2.
Place ~ 'ymbol fn any Ilshl tor vvhtch date ls to defiled 3.
Leave Iield Mank lfdata 4 lo be entered, bul ts currently unavaltaMe 4.
Coinptat ~ the numbered ftatsts below par El 2.35 6.
Show numeric sero as/, and ~IPhabetlc "oh" as as to avoid confusing the computer.
8.
Lait hand lustlfy all ants les.
..;. ".2'7'."" "i~"'" 'VKQI>ihf22<srgirr@t:sEts<'I':Will " 'g';:Qgtsg'."'h".':@'FI.IfccOE II Kl'l p
FA~@QPi" )".!E'ggikgwtt'~~$$ jyY>'"Etj'o;)ll:.r.:"j'jgec>'j'I"""32%jlx"'x~~x39'.2-jSI I. Ere)I!-IB~oh~5~f'v-1 ~l~~ ~~E.Nsss~
- 4. I)Esc Lj3. 696L~~
- 6. hll lli 3~
- s. MQI)ELhr 2.S/JJ (~
9.11EF OIsc/ZONE
- 12. CIIILIE~
ls, corlllsxr L~
- 29. EI.EMl~~~
1g. I SI L1I~
85Z l7 E
ls. Ilhcc LLj
- 11. El EV/EOIIE
- 13. 9299/ocr~
Il.CONN/ASM I2 J
- 24. EQUIP DEscn. ~
- 25. PstoT Li
- 21. CKI'N
- 20. PS2+2-
- 14. VKtJD 16.aC~
ls. IEEE LLJ 19, CONELLj ri I*'"""'"6'!."i-'.~"""!'"9'+""Fcs"9'"62!I"Io'!"O'"Ix 0 cl isllKI7IE oN
. 29!:" "*'ll"-h"."'"Iis'ssliu:."..i!r,'::.ih':*'Ir'I'*9'.'III ';":."9':.2!I'."-.i!his'o::.-
30.rLA0{C.S,Jt.anat LLJ 3t.cotJtnAcr L~<
3
. ImiiilsLL> L.lss. shrsrv Fuxcrialil~~~L>
sElshtlf: ottAL: 39. IIYDnoI.DAns lJ 3lo. 1Esr lJ Lj IJ e2. OUAI.IIIcAIloN SFAIUs. sElshllc Lj EtJV Lj
- 43. TM 3l. ACCUflACY
- 41. AtJALYSIS Es.roso ~
Is.om LL~
- 31. LEVELLl
- 33. EQUIP CLASS U
- 34. USE,LI LI 39.Ilhl~
r'llSI'AtlgDDV lseh le tie nl IS/hhl Rf'Hlf.Zor l'386,
~
rr LZ8~EPc
I. I <In F I ~ I s ~
F 11 11 2 N ti'll I'I'I.YSYS'I'I'.M CIIFCK ONE:
f.) Ashl ~ nsw EPN and Information to MEL MEL INPUT DATKSIIEETEI 2.35 ncr rhag rro 86-0!56.. Oh -Ol7 suransenss rnavioust.y arrnnvstr ocr rhogr l.
Place a ¹ symbol ln any flefd that lr not appllcsbfe 11 2.
Place a 'ynIbof In any field for whlcfr data It to deleted confusing the comput ~ r.
0.
I.~It hend justify all enlrler.
3.
I.esvs llefdblsnk II date ls to be entered, bui Ir currently unsvslfsble h.
Complete tha msmbsred fleMrbelow per El 2.35 Lf J Mozf)fy exlrtlng EpN Informatfon onhiEL I
l.) Dcl~ t ~ exlrllng EPN from htEL (equipment no longer exists In the pfantl
[3 I)fsnk out CIEISIIMfields lsqulpmsnt no longer rslety related, but still exfrtrl Note) Place a "¹ "symbol or i 0fn Field No. 30 See El 2 35lnrtructfonr for Flefd No.30 U lag Change Only; Enter former EPNher ~
2 Enter new EPII ln Field ¹ 1 I
t)ENEflALIIISTIIUCTIOIJS 6.
Show numeric aero ar JI, aml ~Iphsbetlc "oh4S as 0, to avo'Id I"".;':rl.. iX'.:-lsd.:-fl)'5'iZ!'2':>:'<<"":Inlha'EoÃ'-"i""'~-".8'.":3Z"lÃUCX'QM%hti:-
E EIIA ME
~ fffso,n TA> "nanIIE~IE;;::".'z",~;-.:~I'8-"-':"'i.I'."-":;A'.9.".2".4<'s)Yi~/IIIN'i'"'~'lfs t"-".n.n I. ~ FN Q?~~RO~IA - ~SP V - I Z~LLLL
- h. DESC
- 2. NSSS
- 6. htFO 0.MODEL P 83 ZO I 7 2 E S. LOC. DETAIL S. TIE F I)WO/ZONE
- 10. OLDO QJ
- 11. ELEVIZONE
- 12. CVI L)f
- 13. PylflIOUT IS. CO/I I/EXF
- 17. CONNIASM
~l
,< F.N <;N..<(.,
-A.E.I Wa.t):.q:q<p I'- grI):SF44A;. NIY)-".~s <9."x<>Hr<XKY~A'". 0 Cl Sll IE
)
ON lh. VEND 15.ac LI9
as. IIUUIIILLII jas. 5AFETYTU/IOIIUNI~LLLI~LI sflshttc <<tfhll 30. IIYDlloLDAI)sQ
- 40. 1EsT LI LI IJ
- 42. OUALIFIC)gJ.IONSTATUS - SEISMIC LL FNV Q h3.'IM
- 34. USE Lj Lj
<S.OIULj~
32, LEvFLLj
- 33. Eaulr CI.Ass Lj SI Acean.ncYLLLLL hl. ANALYS'lS LLj Lj
<I. F.OEO LLL) rntrAngn ay J dilly PATE CI IKC APRIL Z/1,1986,
<IN Pia OA) E I YAA/f/.'
~ ~IB 66 ~ SS ~
~ MM ~ B L.i st Jl'I'IY SYS'I'I'.lI MELINPUl'ATASIIEET El 2.35 rlCl'AGE HO Blt -Otta( -OA -QIE CI IE K (1NE:
t ) A'dd a naw EPN and lnformatton to MEL star EnsEpEs rnEvtot)sLY Arrnovtp PCt PAGES 0
htostify ~Illrttnp EPIl lnfornratlon on MEL 0 Drlete eatlttng Ef'N troln hlEL tequtpnrant no longer erlbtt ln I'ha plant)
~
6 13 ttlanII out Clf/Stlht tleldt (equ'Ipment no longer tafaiy related, but still arlfrtr) Note'. I'tace a "It"symbol or a 0 ln Ftatd tlo. 30:
Sea E12.351nrtructlont for Field No. 30
~
~
I I 'fag Olange Only; Enter lorlnar EPN tiara L22t Enter naw EPkln Field I/1 t)ENEIIALIIISlAUCTIOIIS 1.
Place a 9 aymbolln any field that tr not appltcat>te 2.
Place ~ 'ynd>>ot tn any lie'fd tor which data lt to deleted 5.
Show numeric aero ar Jf. and alphabetic "oh" ~ r 0, lo arold conturtng the conIputar.
3.
Leare flehl tltanII lf rlata tr to be entered. but Ir currently unaraltabt ~
8.
Left hand 12rtlty all entrlar.
4.
Cnnrplet ~ tha numl>ared tlaldr below per E12.35 6
I;.
': 5'i"..""-..i'.)7'!>>"'.;3'a 457:%K
/1H'f124.".'l->>'l'rl;.4rIE>'y.-',".'; go":fr@:4gz>>",rp.';
E ]E A ME f,tjF A p,.jazzy::'1":,>>s.:tl92>>p'E';,';,4;.";I ""his.:,.'.'s}4r,r;">h~~'.:."2."-.t,"~;;. 'F}...".-'>>Sl.
- 2. NSSS
- e. LOC. DETAIL ls. BlDG'Lj II. ELEV/Snlls~/
- 13. Pyyn/OUT
- 11. COHIf/ASIH
~~l~
Sl. SEIS LLj
- 22. CLEAN LI
- 21. CEPN 29 PS2 2-
- i. EFN I~S~R
/I - 5 r II.
I 5"
- 4. DESC
- 5. hlFO 6MODEI. Mf 03Z h I f ZE
- 9. tlEF t)IVtt/2ONE
/I LNrrrrrr
- 15. CONT/EXP 2>>.rriLi~~ll~
- 26. r>2Ll~~
- 3. EYs LI M.VEND~~
is. nc~
IB. IEEE LLj
- 19. CODE~
LLLIJ
- ~ ",'.- '"- ".4-2
'"t'>4)S'A'4% ".Wgt;"';i "."4>>."; "
'NO ESVPP@<"VY>>43W'D' I'1 Sf1 IE ON 1>I~~BE)t~>::APAEYY'2>",a'S>~~,".'...'.." I.r.".'.::S')4F:...EF'>BMX N'.Pre~~'A'<tel~~ r.i>3->>
- 30. FI.AO tC.S,A',t)tt t)I 4J
- 31. CONTI)ACT L~J 32, LEVELLj
- 33. EQUIP CI.ASS IJ
~:
34.'USE Ll LI 3, IIIIIIBSLI I Lj36 SAFEIYFIIIICYIDIILI~~I ~ l&~~
32.ACCUBAI:Y
- 36. IIMLI~IQ SFIShllC OIIAL: 39. IIYDIIOLOADS Lj
- 40. TEST Lj Lj Lj 4t. ANALYSIS LLj
- 42. (Ilih}IFICAIIDliSihllls
~ SEISMIC Q ENV Lj
- 43. IM LI
- 44. FBEB Ss.nin ~Q 9IIIIIII~~
N ]
PATE Ct)EC ED Y
IF}I'IIII-ZB IML.;
( ~>> fit OAlf l Mfz~li'f.
En> 88 8555 EOM ~ n
."4.1 RtJl'l'l.Y SYS'l'l.'ll C) IECK ONE:
U Alhl e new EPN end Information to MEI.
MEL INPUT DATASIIEET El 2.35 nct tAOE NnB6-0l56,-M-Olr(
strrgnstngs tngvtntrst.y Arrnovrn QCt tAGgr lgJ Modilyexhting EPN Information on MEL l l I)E!~ te exhtlng El'N from MEL faqllfrlrnsnt no longer exists In Ihe plant) 0 OlaEA oilt CIEISAhl flolds (aqsdPmenlno fongar sabty related, but still exfrtsl Note:
Pfaca ~ "tf"Symbol or e Qfn Ftsld No. 30:
Saa El 2 35 Instructions for Field No. 30 l I Iag Change Only; Enter former EPN here (2 t
(
J:
Enter new Ef'N ln Plaid ff I (jENEflALINS fAUCTIONS 1.
Place a It synlhol ln eny field that Is not applicable 2.
11ace a 'ymlxllln any llcld for which data lr to be deleted 6.
Show numeric sero as f, end alphabetic "ohtt as O. to avoid confusing the conlputar.
3.
leave field blantr lfdata ls to he entered, but 4 currandy unavallabla 8.
Left hand lustily all cntiler.
4.
Colnplata the numbers<<f flafds below per El 2.35 I Epn L22-~BZIIFLL3S~F'
-~l~tl I
- 4. I)ESC
- 2. NSSS
- 3. SYS~~~J B.MFOLL~~J B.MCCEL LXF'3 Z OFI l 7 2 E
- 7. SIN l~~~
- 8. LOC. OEYAIL
- s. AEr DwrizoNE
~/
- 10. IlLDG LLj 1'I. ELEVIEONE
>>.CYIlb~ ~~~
- 13. Pwil/OUT
'l0. CON1/EXP L
- 17. CONN/ASM
- 20. E I.EM
- 21. SEIS LLj
- 11. CLEAII LI 23 FPI~
I ~L3 ~1 ~
2<<.EOUIFOESCII. LIJ 26.FSFOF
- 20. TP2
- 27. CEPN
- 28. FSI I~2
- 29. PS2 2
- 14. VEND I 15.OC~
1$ IEEE LLJ IS CODELlJ
'I'2".
2'.",."3:..;.."AE lt'."ELF'..~';j.'-"Ept;:i'.":,MK~O'3",'ptLBr.<<'5 LNg'BL Q C S flh IE D tjN
<<R4WS:."<<> Sp'p JtQi<<4eg i i "l3.2'-'.N...::'.c.'>>:M'<'<<8'3 pe<<A";Et<<oiÃ'r'5~>>~<<2"..l: gp
- 30. TI.AriIc,s,o.of 1 0) LIJ
- 31. cotftfthct 311.11OIIIIS~SB. SAFE'FY FUIICIIOULL3~1 SEISMIC (ttfhl.: 30. IIYttltOlOADS Lt
.40. TEST LI Lj
- 42. QUAuplcAIION stATus - sElsMIC Lj ENv 0
- 43. tM 34.USE Lj Lj
<<B.OIO~~
- 32. LEvELLj
- 33. EQUIP clAss LI
- 37. AcrtfnAcYLLj
- 41. ANALYSfS LLj
<<<<.FnEa ~
- 35. nM~I tttktAttEts tty I IF'
O ~ Mq ~ SBSSS BMO ~ ~
4'Ul'It 7 SYSl'I':M CIIECK ONE:
0 A<Id ~ new Epft end Information (o MEL MEL INpur DArASltEET Et 2.3s ncp pAOE No.
86-0lS6,-0/l-OFj st)) Enssnss r nfvlausL Y /Lrrnovao tll'tPAOS<
Bt.Mod<fy e.>>(fnN an> I<storms)lo<<n MEL s
I I l)sist ~ ~ Missfng EPtt ttom MEL tsq>>fp<ztsn( no fonpsr ~ 3<fr(s In the pfen(t
~
~
I I Blank oul C IEISflMffslds tsqufpmsn( no longer sate(ywsfelsd, bul still ~ 1<fr(s} No(~: Place a "0 "symbol or a 0 tn Field No. 30:
See Kl235 fns(tuc(fons fnt Field tto. 30 U
Tsy Ct<anps Oofy: Kn(et former FPN hate Q2 Enlst new EPtl ln Field 8 I OENEIIAI.INST IIUCTIONS 1.
rtaca a 8 symbol In any llsfd (hal ts nol applicable 2.
Place a 'yn<bot In any'field lor whlctr data ls to ba dele(ed 6.
Show numeric seto as/ and efphsbe(lc -ohsr as 0, lo avofd confusing (he compu( ~r.
3.
I.eave Iles<I bte<A It dots Is lo be en( ~ rest, bul fs curren)ly unavailable 8.
Lel( hend lIts(fly all enlr far.
4.
Co<up!ate sha numtzersd flsfds below pet Et 2.35
'-'i "-. "-'.L-"-:"-'.(<(Sx<'~~(': I(<'z T "QIP(z~ke'Y!":X'jf<<"'Z-"<".'8O'."T@ci)"".'xi(IEQKIIA'K
. NPI A A"<~MR'5,"".'Vr).".z'M."rwe."~x',rssfs:%.a? fit':: Lz<%'Ak$ A+P-"0<<'tzrr".) zx':"-:B"
'.<rnLZI-IR~QLh~s~l' I~CA
- 4. OESC
- 1. NSSS J s. s YsL~~
E.MFB~~
B.MDDEL LN I I<3 2 0 6 I I '2 E I sIII.~
- e. LOC. OETAIL
~~J B.IIBFIIWO/ZOIIE~~~~/
IB,BLOC L<j 1<.ELEV/ZOIIE~L/~
l1. CVI
- 13. Pyff(IOVT ls. coo</Exr
- 17. CONNIASM L
~
.- ~ ~
~
~
- 23. FPI Lt~
~J 24.anvfpoasctt. LLj zs.pstor Lj zs. Trz~<a
- 21. CEPN zs, rsI Lz<~
J ze.pszz II.YECD~~ Is.ncLI
)
- 18. IEEE LLj
- 19. CODELLj
. '..'k.~"r::4'(". ~ w~'~":"P2(ÃzQ:;s".(o:i FI.'RIT."'F/x Arne'FE'sT;."-Ii"%'"i'YN'S";,B."B.KOC1 Slt E
0 I 3'(MDSTT)(iVZ)'u2LÃz'A';.:".z'N':.."3-'"Res"%xATÃ4M'i~kMLR:;s'hxÃs.n&z~.,l"!'.
- 30. F l.Ao [c,s. (I.ott a) LlJ 3t. Cot( TIIACr 3S. IIIIUIIS~336 SAFETY FIIOCTIOIII SEISMI(.'()Ijht.: 39. IIYI)lloLOAt)S Lj
- 40. TEST
}
) Q Lj
- 42. (1UAl.lP ICATION STATUS - SEISMlc
[J ENV Lj
- 43. TM
- 31. ACCUIIACY 41.ANALYSIS~
s<.rnso ~
IB. OID 32, LEVKf LI
- 33. EQUIP CLASS U 34.(ISE Ll U
zs. IIMLI~&
I'IIIIAIIIIIIIY DATE
~nt~)L z.o,>vga OA '(E
) vl< fe
~ >
I\\e IIIV ~ I ~ I II ~ IIIIE B O'L Sl)l'I'IY SYS'l'I!M t:IIEr,KoNE:
4l j Ashl ~ new EPN and Information lo htEL htEL INPUT DATASIIEET El 2.35 tn:p pAcE No.
86-OISIN-Oh-Olg strrrnsanrs pnavioust.Y Arrnovso nCr rhagl
~ ~I
~ I I.j Mrsdify exhtlng EPNInforrnatlon on MEL
+1 I l I)Cl~ r ~ exlrtlug EpN from MFL Isqulprrtsnt no longer exists ln the plant) t )
Ills>>14 out CIE/Sfthf flsfrlr lsrprlpmsnt no longer safety related, but still exhtr) Nota: Place a "0 "symbol or a Oln Field Ho. 30:
See EI 2.36 Instructions for Tlshl tJo. 30 4 'I I I Tag Change Only: Enter former Epf)here t
Enter nsw EPNln Field I/ I OENEIIALINSfAIJCTIONS 1.
Place a //symbol In any flaid that ls not appllcebla 2.
Place a 'ymbol In any field for which data ls to deleted 3.
Leave field btaih lfdata ls to be entered. but b currently unavailable h.
Compla'I ~ I'he numbered llslrb below psr El 2.3$
6.
Show nunr ~rlc rsro ss J(, and ~lphsbetlc "oh" as 0, to avotd confustng the conrputsr.
B.
Left hand lustily all entrlsr.
BV:o"4 "2-~". '<"QP'at'I'."'V',".Ia;"/4"~'V'XSP'j=t.."<<V <.:-";Bo(4VP~MZ".')f2(tH GE Efl I; ME f/PU
. OA A Bli'( 7"'""">I':"-B"4<;('v"'.o~'3o."..".o~ln<'"A"'c@L'tie%".BPW 5":.Vso'4 2;:I":::~.:...'S'-'-:
I EFN LSI-IKjaah-S~I' 1 7~
~. I)ESC S.LIFGLI ~1 E.Moos(. LNIF 8 3 2 0 T.s/N LI I B. IIEF l)wolTBBE
/
(2. cv( LEEI le. coNT/Exl
- 2. NSSS I IZE
- e. Loc. DETAIL
. Ie. BLOB LLj Il. ELEV/ZONE~~/~
- 13. PVJII/OtjTI
- 17. CONN/ASM
le. IEEF LLj
- 10. conE LIJ
- 23. II'ILI I
J
- 20. rr2
- 20. PS1L2-M~~I~I'l.IE(s LLj
- 22. CLEAN Ll 24.EGUB 6Esco. LIJ Ie.rsror LJ
- 21. CEI'N
- 28. PS2 2 3(l. fLA(lIC.S.F.O(l Ol LIJ 3I. CBIITIIACT 1 LJ ~l
- 32. LEVELLI
- 33. EGUIF CLASS U
- 42. (Il(AI.(rrcAI(oilI'IA'Ills-IElssllc [J ENv Lj
- 43. TM Lj I~. FIIEB he. OID I +XI'
~su
)>3k l AfIIIL28',)986.'
8/( ~(( 8<<
SI2o B". Fo:3 te45$ i2(<<."..oS<('2'-'(43A..'BR".I IAM4"
Prepa~
by
- i. L.MILc'5 INSTAL'TION AND T ST R-CUIR-"1ENiS 0"P Page B, -O)56 -OA
- DI i APRIL Z9iI 9'IER 2-86
-CI!5E - 0: Modi~ication Ti.le SOL=MQID NtLVc F=P'LAC=M=)J IIISTALLATIONR""OUIRPfcNTS o
The o11owing design associa.ed with this PcF"R TO FPM's LISTE:D speci. ications must be adhere'd to during ield work (IZEr G~'.MIFICAYIOMS 2805 2I-SAMD-ZZQ)
BE. LOW.
o Inspections are reconmended to check the following impor ant parameters Insoec.ion Parameter To Se Checked Fr.KFdRM Ul&UAL INSF'="CTIDM 70 VER)FY QO QBVIbUS DE'rrrTX AMID CORE""~i IA!GihLLATIQIU CONSTR U i ION i=S i S The following cons.ruc-.ion test should be per.ormed
.o con. irm cons.ruc.ion c~pl etion.
~RFDRM V/!FiIQG CPIQTIIUVI i Y C)f-CKS'AMD QP:-KRT)ONRL FUNCT(OhfAL i UT Ar iCR KcPLAC""M=57.
F'PM'5 IQ.2.'ll>
1IJZiALLAiIDN/MbDlrIChtlblo C)r' INSIDE.
~~ PROD~ TUZIIGb.
)O.Z,'{.1Z.
IM)TR.'TUSI!U&5 RTT)Mb USA&: 1~ii KUCilDM.
)Q.~&.)P ARM)h)FitlC))J h13D SPLICING IMSiQl&d7Ji
)Q.Z.)D rRSY=QKR ibRQQE AND i t.h'SIQQ %)6.
D=-EGN BA"GJP AND P=Y::-8 AP. B{D+X D
P Ho.
QZ-86 Ol56 O
I his e genel" corm'ns we oesigr be" mo enC r<<vi~ oco ~en w=ion a,""=.
i e:e~ "'orina oevelopnen of ".his DG'.
The an:er'z C.:ris Aooendi" e.
D~aian Ve. i,ia,=ion Re "r" Cellule ion aver She
=s Spe=iei Design ¹vi~ Re"o-.m es.ollow:
peges Of9 GZ.w purges peg es peges pages
2 we
~ ~
'Rev.
3
~
~
DCP Page BE -Q!56-5/
-'cDl5 DKIBH VVcIFICATIOH p rempare d b
~ L ~ lQ 5 APR IL 2 3v l 9Fa Yew.ier Dame B.W Ua>>
E Rem:
lni,al Doc; Dam ak'/eri
- fie, Name Veri,ier Bare Inia al Ini.ial
- Da-~
- Dan PliR 2-R( -0(50- 0; Modification Ti
- le SodFAoro j/4'vr g lac'acdooS>~
Design veri ication was demeaned u be necessa~
and
~allows was pe&armead as AL'AMATIVECALCULATION: yes no o
Calcula ion Ho.
(cir le ane)
, Revision 2.
RIMG:
yes no
~.
o Spew, icz.ian, o.
Ti'e (cir"la one)
Revision Da-Chai~n DAIL RYIBf o
Formal Design Review per
="I 2.7:
yes (RepoW Ho.
- Dam P~aarer
Ha& One Ho.
I ~
Dues-.lan nero.me allowna inaum car ~ly.
sele~
and incarpora-d inm ~e desi gn'!
Yes No Ho.
Aaolicable 0
Basic,unc-ions o
each s w=~, sys ~ and c~onen:.
P ~armance reoui~enw:
(caaac'~,
raw ng, sys-m au~u:...
)
Desi gn cond',~ ons:
(". ssu~,
. zw~..luid chemis vol ~ae....
)
Qo 2te Rev ~
3 DCP Page S. -5!9, -Qk -@g Hc.
Ques icn Preoarer:
Yes Ho HaH:
One Yer ', ied B)t By ADDiicabi e (In~.,al )
0 External loads:
(seismic, wind, Werml, dypamic..'..)
m~ vironaenwi condi ions anN ci-patef during storage, consmc-Non and opera%on:
(pr ssu~,
era~re, humi di.y, corro-siveness, sin elevation, wind dirac ion, radia<on, dura".ion 0f exposure
~ )
RBcfQi~~ 'mposed on Ale design bv fun~onal and physical i~rfaces wiN s~ucwres p sys
~ and componen~.
Retrial reauiremenw:
(compa-ibility, eie~cai insuiaaon prope~es, pro~ve coa ing, corrosion resis ~ce.
Sui ~il-i for applicz<on and environ-
@en~see o ~ ~ )
Hechani cai reoui~n-:
(vibraaon, s~ss,
- shock,
~mon for s....)
S~J~ml ~uirmenw:
(eou'lP-men-founda
.Ons, pipe SJppo~
~ ~ )
Hvdraulic renirmenw:
- HPSH, allowable pressur
- droos, allowable
,luid veiociues....)
Chemis~ reou'..
menw:
(provi-sions for sampling, limiu:ions on wa ~r chew% s~~
~ ~ ~ ~ )
(She - 2 cf 7)
Rev.
3 OCP Pane 86 0-!9 -OA -GZl
~~oa~
Na+
Cll No.
Once%on Na ves No Awolicaol e Verified Py
('ni iala)
=ie
. ic21 requiremenM:
(sour e
of power, fuse lis=, voltage rac~y requiremenm, ele~cal insula.ion, motor requi~nw, grounding, maintainabili.y, separaNonfor safe shuAom....)
o Layou-and arrangement reauir men w:
(access cl earances, ins umen. loca ion, &ermtl expansion, seismic 2/1....)
i 0
Opem ional requirenenM under vawous condiaons:
(plan
- sump, normal plan operation, plant shuAcwn, plan emergency "
opera ion, special or inf~ent
- opemNon, and sys~
abnormal or emergency opemtion....)
o Indus~al -safety/, ire protection requirenen w:
(Appendi" R consi d-era ions, fire e~nguisnmen:,
inert gas ha-aHa, use of aarius-able materials, ven<lzaon or ven. ng paN changes,.ir pro-wc.ion systems....
)
o Ins~~vu<on and con-ml
~ui~nw:
(including fns~-
merrz, con~is'and ala~ reouired for ooe. aw on/test ng,/maintenance, se-.wins,
- ranges, ape o,. insw-men-, inswlled spares, loca
.'on of con~1 fndicaNon....)
(Y filo
-I AiAchment 2
O I a I
D-"P Paae Bf, 0(54 -DA -c5
~ Z Preparer:
Ham One Oues:ion Failure ei.e~w and prevention
,or s~cwr s, systems and componen=":
( ailsafe design, redundancy, diversi>....)
o Nafnwfnabfli.y requfremenw under normal and o
-normal condiions.
Not Yes Ho Aool icab i e Veri,ied pv
(.ni".ial) o Personnel r quiremenN and lfmf&-
.ions:
(physical, A'R', human 0
Transpo~f1 fty
. qufremenw:
(size, shipping wef gh, I.C.C.
regulations; o%er handl ing/
storage/shipping considera-
<<I ons
~
~ )
Cl eanliness requirenen w.
Suf <<ab'ilf.yo, pa~
and equip-ment for Pe application:
.(environmen~l qual fffcation, sef smf c qual fFicz< on, sf e, ref gh-,
operaang range, availabil f~.... )
Are assump-icns necessary u perform
%e design adequa-~ly des-.ibed and reasonable?
'das an appropria design method used?
ls ~e oiwu reasonable "impar d'm inpu <<s?
Are We m~~ qual
.y cz~cory and qual im assurance requf~nm speci-,ied?
4A
4 kg
- -~ 2.5 Rev.
3 Pa.
Oues lan D."P Page Sa
-O.SS -o~ -O":-
Preoarer:
Ha&
One Verified Ho By Yes No Aoolicabl e (Ini='al )
'W ~
Are We applimble codes, standards and regulatory reauiremen~
(including issue and addenda) properlv icienti,ied and are Weir requiremenw for oesion met?
Have cons~~ cn and opem:i ng expe~-
ence be n factored inn %e design?
Have inspe~on and maintenance requir-ments be n saw sfied?
10.
Are ac=essibili y and oWer design provisions adejuam for perRrnance of maintenance,
'SK and calibra ion?
Have Ne
(}C insp~on-aA ria be n fncorpora~d in Ne DCP m alla> veri-ficaNon Na. design requiremenM have be n sa-'sfacwry ac"~1ished?
Have adeoua
~
@equi reme~ bem appropria~lv speci,ied in De D"??
A cnmelt " 2 (She
"" cf 7)
~
~
q
4
0"P Paae~<
C-e5F -OA -CZ!
P~aarer:
Mark One No.
Ouestion Hat Yes No A ali"able Veri,ied By
(:ni=ial}
11.
Goes this DCP a
ect i wm designed/
cans~~
~d o-, i-~ cave, d by ~K rules?
any "Yes",
~viM Items covered by Subsection HB, NC, ND, or NK?
I-ms cavered by Subsec-ion NF?
I~ covered by Se~on YI:I?
If the AStK CDS fs affected, has a
DCP plate be e included fn the package fdenN~ng Ne change?
Has a P~,essfonal
~wginee.
Ce~, ica an Form be n completed and included fn Ne DCP ~ar all changes being mace M %e KS?
Has Ne ce..f ied s ess r po~
be n uada d
or AWK I:I-1 changes?
Have design calcula.,ans ar othe>>.
AW"= I.I hark be n updated?
I re ove~ressur pro w~d on repar.
a, ec~ by afs &ange?
Does the DCP include recufrnf changes ta te ASHES I:I-l over-pr ssure prate~~~i on repor <<?
Has an evaluation be
- r. ma¹ as M Ne imaac o
We mange on overpressure on MME-:II-2 and AW I:I-3 systems?
o-, the '~ 11 questiorrs ar mam%
We DCP cast be rau=d.ar ~~i.=
pe.
40
.lee (She
=
". a.
7}
I A~
V
~ 0, i
~
~
f04 Zt v
- Rev.
3 DCP Page~(
Qtyg Qg f75 Preaarer:
Hark One Ho.
Oues-ion 12.
Does Ne design impose r quiremenm on inurfacing svs~,
s.. vc wres, or cartponenw wiiich r~: be accaaadz~d by a desi cn change or a reanalysis?
l so, We required aesign change or r analysis mus be incluoed in tis DCP.
12.
Mas a physical inspec"ion included in We plan-?
14.
Does His DCP a-.fe~
any af We four Safe Shutdown analysis
( ire prow~on, ele~cal separation, pipe break/fossil e or con~1 sys~ faile
)?
I, so, an updawd anarysis mus be performed.
(Se EI 2.40) 15.
Have normal and oR'-normal opemaonal requiremnw (including personnel limiaeons
} beon considered?
l6.
Does %is DCP resul-in a change M
<<'te design 'inwn far nades or me Nods of one. aNon?
17.
Has same or all af We DCP be n dis-cussed H W knurl edgeable pl an-s=,
personnel?
18.
Did Ne original design incluae r parable oefecw as cefined in 10 W Zl?
No-Yes Ho Paalicable Ve. i,.ted By
(:n3= a'I)
(Shee.
7 of 7)
4 r
s
~ ~ 4
""t I
ff~
t MNP-2 G"-N"-RA:ION ""NGIN:-"=RING Priori.y
'7" TD Documen. Control ingineering TRANSYIt tAL NO.
H7~ 7 DAT-"
M /- FL.
DOCUHEH B=ING TRAHSMIi i =D 0- o/5 d-c,d Resbo nds To P
(.- 05@
Reference Doc.:
Advance
.'n o For".
I 7
V t
CUMiiVi'KC:-IPT ACKNOML""DG"-"L~NT I
Signature Da ~e HHP-2 Document Cont. ol is reques ed to make
.he foliowing dis:ri'"u ion o he issved Design Change Package (DCP)
L. Barndt Q. Densley C.
Hexum (1)
(2)
(1) 1020 {""Or) 9sl 5 R+i
- a. ~eil (2)
KA Viliovghby (1)
Design/Dra
-itng {1) gssu 9Kb (V-"/C) 4
~
7t