ML17304B012
| ML17304B012 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 01/31/1989 |
| From: | ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | |
| Shared Package | |
| ML17304B009 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 NUDOCS 8902140045 | |
| Download: ML17304B012 (239) | |
Text
FINAL REPORT ON PRESSURE-TEMPERATURE LIMITS FOR THE PALO VERDE NUCLEAR GENERATING STATIONS 85'02140045 S90131 PDR ADOCK 05000528 P PDC 1669K/0059K
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TABLE OF CONTENTS SECTION TITLE PAGE
1.0 INTRODUCTION
2.0 ADJUSTED REFERENCE TEMPERATURES 3.0 REACTOR PRESSURE VESSEL BELTLINE PRESSURE-TEMPERATURE LIMITS General Method 9 Cooldown Limit Analysis 12 Heatup Limit Analysis 14 4.0 COMPOSITE PRESSURE - TEMPERATURE LIMITATIONS General Limitations 15 LTOP Enable Temperatures 17
5.0 CONCLUSION
S 19
6.0 REFERENCES
20 Page 2 of 21
II LIST OF TABLES NO TITLE Palo Verde Unit 1 Materials
- 2. Palo Verde Unit 2 Materials Palo Verde Unit 3 Materials Controlling Materials and Their Adjusted Reference Temperatures Comparison of Current and Previous ART Predictions for Controlling Materials RCS Pressure/Temperature Curves O'F C/D at 1/4t Location (8 EFPY)
RCS Pressure/Temperature Curves 100 F C/D at 1/4t Location (32 EFPY)
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LIST OF FIGURES NO. TITLE ANPP Cooldown P-T Limits (10 Years)
ANPP Heatup P-T Limits (10 Years)
- 5. ANPP Inservice Hydro Limits (10 and 40 Years)
Reactor Coolant System Pressure/Temperature Limitations for 0 to 10 Years of Full Power Operation Reactor Coolant System Pressure/Temperature Limitations for 10 to 40 Years of Full Power Operation Page 4 of 21
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1.0 INTRODUCTION
The following sections describe the basis for development of reactor vessel beltline pressure-temperature limitations and composite Reactor Coolant System (RCS) pressure-temperature limitations for the Palo Verde Nuclear Generating Station. These limits are calculated to meet the regulations of 10 CFR Part 50 Appendix A,<~> Design Criterion 14 and Design Criterion 31. These design criteria require that the reactor coolant pressure boundary be designed, fabricated, erected, and tested in order to have an extremely low probability of abnormal leakage, of rapid failure, and of gross rupture. The criteria also require that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, and testing the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized.
The pressure-temperature limits are developed using the requirements of 10 CFR 50 Appendix G<~>. This appendix describes the requirements for developing the pressure-temperature limits and provides the general basis for these limitations. The margins of safety against fracture provided by the pressure-temperature limits using the requirements of 10 CFR Part 50 Appendix G are equivalent to those recommended in the ASME Boiler and Pressure Vessel Code Section III, Appendix G, "Protection Against Nonductile Failure."<~> The general guidance provided in those procedures has been utilized to develop the Palo Verde pressure-temperature limits with the requisite margins of safety for the heatup and cooldown conditions.
The Reactor Pressure Vessel (RPV) beltline pressure-temperature limits are based upon the irradiation damage prediction methods of Regulatory Guide 1.99 Revision 02<4> . This methodology has been used to calculate the limiting material adjusted reference temperature (ART) values for Palo Verde Units 1, 2, and 3.
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This report provides composite pressure-temperature limits for the reactor vessel in accordance with 10 CFR 50 Appendix G for two representative points in the RPV lifetime corresponding to 10 and 40 calendar years of full power operation. The events analyzed are the isothermal, 10'F/hr, and 100'F/hr cooldown conditions and the 10'F/hr, 20'F/hr, 40'F/hr, and 100'F/hr heatup conditions.
Based upon the pressure-temperature limit analyses provided within this report, no life limiting vessel integrity issues are anticipated to exist during the 40 calendar year design life of the reactor pressure vessel.
However, heatup and cooldown rate or administrative restrictions may need'o be examined to ensure compatibility between the pressure-temperature limits and the Low Temperature Overpressure Protection (LTOP) analyses.
2.0 ADJUSTED REFERENCE TEMPERATURES In order to develop pressure-temperature limits over the design life of the reactor vessel, ART values for the controlling beltline material need to be determined. The ART values of reactor vessel beltline materials for Palo Verde Units 1, 2, and 3 have been calculated at the 1/4t and 3/4t locations after 10 and 40 calendar years operation. By comparing ART data for each material, the controlling material for each Palo Verde unit, as well as the controlling materials for all three Palo Verde units, have been determined.
The ART values have been calculated using the procedures in Regulatory Position 1.1 of Regulatory Guide 1.99 Revision 02. The calculative procedure for the ART values for each material in the beltline is given by the following expression:
ART Ini.tial RTgQ7 + RT>>> + Margin Page 6 of 21
1
~K
Initial RT110T is the reference temperature for the unirradiated material. 5 RT~0T is the mean value of the adjustment in the reference temperature caused by irradiation and is given by the following expression:
d RTHDT (CF) f(0.28 - 0.10 log f)
CF is the chemistry factor for the beltline materials, which is a function of residual element content, i.e., weight percent copper and nickel. Regulatory Guide 1.99 Revision 02 provides values for the chemistry factors for welds and for base metal plates and forgings. The term f is the neutron fluence at any depth in the vessel. The neutron fluence at any depth is given by the following expression:
(e-0 '4x)
The term fgUpf is the calculated value of the neutron fluence (10 n/
cm2, E> 1MeV) at the inner wetted surface of the vessel at the location of the postulated defect (1/4t or 3/4t), and x is the depth into the vessel wall from the inner wetted surface in inches.
Margin is the quantity that is added to obtain a conservative upper bound value of ART. The margin term is given by the following expression:
Margin - 2 2
Q'
'2 g'
't
'I 8
represent the 'tandard deviation for initial
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and 0>
RT<0 T and the standard deviation of the mean value for reference temperature shift.
The following information provides the basis for the calculated ART values for Palo Verde Units 1, 2, and 3.
- 1. Plant capacity factor is assumed to be 80%.
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- 2. Peak-end-of-life fluence was assumed to be 3.29 x 10"~
n/cm~ (E>lMeV) at 32 Effective Full Power Years (EFPY) for both the lower and intermediate shells. Peak fluence at 8 EFPY would therefore be 0.82 x 10 n/cm based on a linear fluence accumulation.
Shell course minimum reference thickness were assumed as 9.0625" for the lower shell and 11.190" for the intermediate shell.
- 4. Calculations were based on the procedures in Regulatory Position 1.1 of NRC Regulatory Guide 1.99, Revision 02 (May 1988).
ART values for all beltline materials at the 1/4t and 3/4t locations after 8 and 32 EFPY were calculated using Regulatory Guide 1.99 Revision 02 and the results of the calculation are listed in Tables 1, 2, and 3 for Palo Verde Units 1, 2, and 3. The controlling materials are shown in Table 4. Note that 8 and 32 EFPY correspond to 10 and 40 calendar years of full power operation given an 808 plant capacity factor.
Review of the calculated ART values identifies the controlling material for all three Palo Verde Units to be the Palo Verde Unit 1 intermediate shell plates M-6701-2 and M-6701-3. In all three Palo Verde Units, the welds always showed lower reference temperatures than the base metal, i.e., lower initial RTQQ7 and lower ART after irradiation. Therefore, only the base metal and not the weldments is predicted to be controlling during design life. The limiting ART values based upon the Palo Verde Unit 1 intermediate shell plates are 102'F and 90'F for the 1/4t and 3/4t locations for 10 years of operation, and 116'F and 103'F for the 1/4t and 3/4t locations for 40 years of operation.
Note that two different sets of chemical content data were available for the reactor vessel beltline welds; one set being the weld metal certification tests, and the other being vessel weld seam sample analyses. The former set tended to be more limiting (i.e., produced a Page 8 of 21
L slightly higher chemistry factor) and, therefore, was used in calculations of ART. Even with the more conservative weld chemistry factors, the plates remained as the controlling vessel beltline materials in each of the three Palo Verde units.
Table 5 provides a comparison of the ART values for Palo Verde Unit 1 using Regulatory Guide 1.99 Revision 02 and the previously calculated values using the C-E design curve (Figure 5.3-5 of the Palo Verde Final Safety Analysis Report) and radial flux attenuation factors without adjustment for flux spectrum variation through the wall thickness.
Two significant features are evident from Table 5. For the 1/4t location, the Regulatory Guide 1.99, Revision 02 predictions are more restrictive (higher) at 8 EFPY and less restrictive at 32 EFPY compared to the previously calculated values. For the 3/4t location, the Regulatory Guide yielded more restrictive values of ART at both 8 EFPY and 32 EFPY. These results are typical of the impact of the Regulatory Guide on reference temperature shift predictions.
3.0 REACTOR PRESSURE VESSEL BELTLINE PRESSURE - TEMPERATURE LIMITS General Method The analytical procedure for developing reactor vessel pressure-temperature limits utilizes the methods of Linear Elastic Fracture Mechanics (LEFM) found in the ASME Boiler and Pressure Vessel Code Section III, Appendix G, in accordance with the requirements of 10 CFR Part 50 Appendix G. For these analyses, the Mode I (opening mode) stress intensity factors are used for the solution basis.
The general method utilizes Linear Elastic Fracture Mechanics procedures. Linear Elastic Fracture Mechanics relates the size of a hypothetical flaw with the allowable loading which precludes crack initiation. This relation is based upon a mathematical stress analysis of the reactor vessel beltline and upon experimental measurements of general beltline'material fracture toughness properties as described in Appendix G to Section III of the ASME Code.
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The reactor vessel beltline region is analyzed assuming a semi-elliptical surface flaw oriented in the axial direction, with a depth of one quarter of the reactor vessel beltline thickness and with an aspect ratio of one to six. This postulated flaw is analyzed at both the inside diameter location (referred to as the 1/4t location) and the outside diameter location (referred to as the 3/4t location) to assure the most limiting condition is achieved. The above flaw geometry and orientation is the maximum postulated defect size (reference flaw) described in Appendix G I
to Section III of the ASME Code.
At each of the postulated flaw locations, the Mode I stress intensity factor, K~, produced by each of the specified loadings is calculated and the summation of the K> values is compared to a, reference stress intensity, K><, which is the critical value of K~ for the material and temperature involved. The, result of this method is a relation of pressure versus temperature 'or 'reactor vessel operating limits which preclude brittle fracture. K>< is obtained from a reference fracture toughness curve for reactor, pressure vessel low alloy steel as'efined, in Appendix G to Section III of the ASME Code. This governing curve is defined by the following expression:
KrR 78 + 1 223 e[.0145(T-RTNDT + 160))
- where, KlR reference stress intensity factor, Ksi~n temperature at the postulated crack tip, 'F RTuo v adjusted reference nil ductility temperature at the postulated crack tip, 'F.
For the postulated heatup and cooldown events, the pressure stress intensity factor K~>, is calculated for the assumed reactor vessel beltline pressure (pressure is assumed as the independent variable in the analyses). Also, for the heatup or cooldown events the maximum temperature gradient across the reactor vessel wall is calculated and the Page 10 of 21
corresponding thermal stress intensity factor, K>>, is determined.
Through the use of superposition, the thermal stress intensity is added to the pressure stress intensity to determine the minimum temperature for the assumed pressure loading.
In accordance with the ASME Code Section III Appendix G requirements, the general equations for determining the minimum temperature for any assumed reactor vessel beltline pressure during Service Level A and B operation are:
2K)g + Kgy < Kig 1, 5 K~ N + K~ y < K~ g (Inservice Hydrostatic Test)
- where, K<< Allowable pressure stress intensity factor, Ksi~n K~ y Thermal stress intens ity factor, Ks i~in K~g Reference stress, intensity, Ksi~in Uncertainties for instrumentation error, elevation, and flow induced differential pressure corrections are accounted for in the final limits, as are temperature instrument loop error's. Consequently, the pressure-temperature limits are provided on coordinates of indicated pressurizer pressure and indicated cold leg temperature.
The pressure correction factors and the temperature correction factors utilized in this analysis are as follows:
Pressure Correction Factor Pressure Ran e 170 psia P > 750 psia 120 psia P < 750 psia Temperature Correction Factor - 8'F Page 11 of 21
The pressure correction factors are based upon flow differential pressures associated with three Reactor Coolant Pump operation.
Cooldown L mit Anal sis During cooldown, membrane and thermal bending stresses act together in tension at the reactor vessel inside wall. This results in the pressure stress intensity factor, K>H, and the thermal stress intensity factor, K>>, acting in unison to create a high stress intensity. At the reactor vessel outside wall the tensile pressure stress and the compressive thermal stress act in opposition resulting in a lower total stress than at the inside wall location. Also neutron embrittlement, the shift in RT~p>, and the reduction in fracture toughness are less severe at the outside wall compared to the inside wall location. Consequently, the inside flaw location is more limiting and is analyzed for the cooldown event.
In general, the thermal stress intensity factors are found using the temperature difference through the wall. They are then added to the assumed pressure stress intensity to find the allowable K~~ value and 1
consequently the minimum temperature.
The cooldown pressure-temperature curves are thus generated by calculating the minimum temperature at the reference flaw at the 1/4t location based upon, fl 2 K)g + K)g < KI R
- where, K~~ - Allowable pressure stress intensity, Ksi~in K~ g Reference stress intensity, Ksi ~in K>> - Thermal stress intensity maximum Ksi~in Page 12 of 21
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The pressure stress intensity factor K>> is determined by:
KrN 4 ~m
- where, Membrane stress influence coefficient 0~ - Hoop stress, Ksi The thermal stress intensity factor, K>> is determined by:
Kr v - Me DTu
- where, M< - Thermal stress influence coefficient DT~ Reactor Vessel wall differential temperature, 'F To develop a composite pressure-temperature limit for the cooldown event, the isothermal pressure-temperature limit must be calculated. The isothermal pressure-temperature limit is then compared to the pressure-temperature limit associated with a cooling rate. Then the more restrictive allowable pressure-temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline.
Tables 6 through ll provide the results of the isothermal, 10'F/hr, and 100'F/hr cooldown pressure-temperature limits. These tables provide the indicated RCS temperature versus the indicated pressurizer pressure for the various cooldown rates. The allowable pressure is in units of psia while temperature is in units of 'F. Figures 1 and 3 provide a graphical presentation of the cooldown pressure-temperature limit results found in Tables 6 through ll. It is permissible to linearly interpolate between i
the cooldown pressure temperature limits.
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Heatu Limit Anal sis During a heatup transient, the thermal bending stress is compressive at the reactor vessel inside wall and is tensile at the reactor vessel outside wall. Internal pressure creates a tensile stress at the inside wall as well as the outside wall locations. Consequently, the outside wall, location has the larger total stress when compared to the inside wall. However, neutron embrittlement, shift in material RTNpy and ~
reduction in fracture toughness are greater at the inside location than the outside. Therefore, both the inside and outside flaw locations must be analyzed to assure that the most limiting condition is achieved.
As described in the cooldown case, the reference stress intensity factor is calculated based upon the addition of the pressure stress intensity factor and the thermal stress intensity factor. For heatup, the thermal stress intensity is calculated for both the 1/4t and 3/4t locations based upon the temperature profile through the wall. The allowable temperature is then determined by superposition of the thermal stress intensity factor with the pressure stress intensity at the flaw locations. For calculation purposes, the allowable temperature is calculated using the same equation format as the cooldown case.
To develop composite pressure-temperature limits for the heatup transient, the isothermal, 1/4t heatup, and 3/4t heatup pressure temperature limits are compared for a given thermal rate. Then the mostt restrictive pressure-temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event.
Tables 12 through 21 provide the results for the 10'F/hr, 20'F/hr, 40'F/hr, and 100'F/hr heatup pressure-temperature limits. These Tables provide the indicated RCS temperature versus the indicated pressurizer pressure for the various heatup cases. The allowable pressure is"in units of psia while temperature is in units of 'F. Figures 2 and 4.
provide a graphical presentation of the heatup pressure-temperature limit results found in Tables 12 through 21. It is permissible to linearly interpolate between the'eatup pressure-temperature limits.
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The inservice hydrostatic limits are provided in Tables 22 and 23 and in Figure 5.
Reactor Vessel Data Reference (5)
Design Presure 2500 psia Design Temperature 650'F Operating Pressure 2250 psia Beltline Thickness 11.19 in Mean Beltine Radius 97.07 Material - SA 533 Grade B Class I 'Reference (6)
II Thermal Diffusivity- . 399 ft~ /hr Young's Modulus 29.2 x 10~ psi Coefficient of Thermal 7 x 10 ~ in/in/'F Expansion Yield Strength 44.5 x 10~ psi Poisson's Ratio 0.3 10 Years 40 Years 1/4t 102'F 116'F 3/4t 90'F 103'F 4.0 COMPOSITE PRESSURE - TEMPERATURE LIMITATIONS General Limitations The previous sections focused upon the RPV beltline pressure-temperature limitations which are significantly affected by fast neutron fluence changes and the irradiation damage predictions of Regulatory Guide 1.99 Revision 02. However, these beltline pressure-temperature limitations are not the only pressure-temperature limits which can restrict operation Page 15 of 21
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of the RCS in order to provide protection against non-ductile failure.
Consequently, other locations of the RPV have been evaluated, specifically the flange and nozzle regions, to ensure the composite pressure-temperature limits are appropriate for fracture prevention.
The composite RCS pressure-temperature limits provided in Figures 6 and 7 are developed based upon the RPV beltline and flange regions. The RPV primary coolant nozzles were found to be non-limiting. Examination of the composite RCS pressure-temperature limits generally illustrates that the flange region is controlling for the 100'F/hr heatup condition, while the flange and beltline regions are controlling for the inservice hydrostatic test, the 10 F/hr heatup to 100'F/hr cooldown limit and the 40'F/hr heatup limit. The isothermal, 10'F/hr cooldown, and the 20'F/hr heatup limits at the lowest temperatures are controlled by the RPV beltline region.
A core critical limit is also included in the figure as referenced by 10 CFR 50 Appendix G. When the core is critical (other than for the purpose low-power physics tests), the temperature of the RPV must be at least 1'f 40'F greater than the minimum permissible temperatures corresponding to the limiting heatup and cooldown curve, and greater than the minimum permissible temperature for the inservice hydiostatic test pressure. The core critical limits provided are based upon the 100'F/hr heatup and the
,I minimum temperatures for the inservice hydrostatic test.
The minimum boltup temperature provided is based upon the recommendation of Appendix G to Section,III of the ASME Code. It is recommended that when" the flange and adjacent shell region are stressed by the full intended bolt preload and by pressure not exceeding 20% of the preoperational system hydrostatic test pressure, the minimum temperature in the stressed region should be at least the initial Ropy, The most limiting Ropy in the flange region for all three Palo Verde units is 60'F. Consequently, when instrumentation error is taken into account the minimum boltup temperature is 68'F.
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The lowest service temperature is provided only as a reference point and does not impact operation since the RPV flange and beltline limits are more restrictive. The lowest service temperature is the minimum allowable temperature at pressures above 20% of the preoperational system hydrostatic test pressure. This temperature is defined by Article NB 2332 of Section III to the ASME Code as equal to the maximum (most limiting) Ropy 'for the balance of RCS components plus 100'F. The most limiting Ropy is associated with the fabrication material of the Reactor Coolant Pumps and/ is equal to 40'F. Consequently, when instrumentation error is taken into account the lowest service temperature is 148'F.
The composite RCS pressure-temperature limits have been corrected by the pressure and temperature correction factors identified in the previous section. These correction factors are used to correct'he results to indicated pressurizer pressure and indicated RCS temperature.
LTOP Enable Tem eratures I
The LTOP enable temperature defines'he RCS temperature at which LTOP is required in order to protect the composite pressure-temperature limits from being exceeded during a hypothetical transient. Above the LTOP enable temperature, the pressurizer safety valve with a setpoint of 2500 psia protects the composite pressure-temperature limits from violation.
lt During the cooldown event, the LTOP enable temperature determines the RCS temperature at which the LTOP is initiated, while during the heatup event, the LTOP enable temperature determines the RCS temperature at which LTOP is no longer required. For the Palo Verde Units 1, 2, and 3, the LTOP enable temperatures correspond to the cold leg temperatures at which the Shutdown Cooling System (SCS) suction line relief valves must be aligned to the RCS to provide overpressure protection.
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The LTOP enable temperature is also dependent upon the cooldown or heatup rate associated with the pressure-temperature limit which is protected.
For consistency with previous analyses the LTOP enable temperatures presented within are based upon a 100'F/hr cooldown pressure-temperature limit, and upon a 40'F/hr heatup pressure-temperature limit.
The LTOP enable temperatures associated with the composite pressure-temperature limits of this report are as follows:
10 Calendar Years of 40 Calendar Years of Full Power Operation Full Power Operation 267'F for Cooldown 281'F for Cooldown 302'F for Heatup 315'F for Heatup The present Technical Specification LTOP enable temperatures for 10 calendar years of full power operation are as follows:
255'F for Cooldown 295 F for Heatup Consequently, there is a 12'F and 7'F change in the LTOP enable temperatures for cooldown and heatup, respectively, for the 10 year time period.
While the LTOP enable temperatures have changed, other LTOP considerations need to be evaluated to ensure protection of the composite pressure-temperature limits during hypothetical transients. These other considerations include heatup and cooldown rate limitations over specific RCS temperature intervals, primary to secondary system differential temperature limitation starts, and High Pressure Safety Injection (HPSI) pump power lockout requirements.
It is therefore required that a LTOP analysis be performed to determine the LTOP parameters associated with the pressure-temperature limits provided in this report.
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5.0 CONCLUSION
S RCS pressure-temperature limits have been developed for Palo Verde Units 1, 2, and 3 for 10 and 40 calendar years of full power operation. The reactor vessel beltline pressure-temperature limits have utilized the irradiation damage prediction methods of Regulatory Guide 1.99 Revision
- 02. The use of this new Regulatory Guide has caused the predicted irradiation damage in terms of ART to increase at the 3/4t wall location by 50'F and 55 F for 10 and 40 calendar years of full power operation, respectively, relative to the original values. The use of this Regulatory Guide has also caused the ART values at the 1/4t wall location to increase by 27'F for 10 calendar years of full power operation, and to decrease by 21'F for 40 calendar years of full power operation.
The impact of the Regulatory Guide 1.99 Revision 02 shift prediction methods, coupled with the fracture mechanics analysis, has caused the reactor vessel beltline limits to become more restrictive than those of the previous analysis. The increased ART values at the 3/4t location are significant since the magnitude of the change is large, and since the 3/4t location typically controls the heatup pressure-temperature limits.
In addition, since the 1/4t location controls the cooldown pressure-temperature limits, more restrictive limits will exist earlier in life and slightly less restrictive limits will exist later in life when compared to previous analyses.
The LTOP enable temperatures have changed in response to the shift in the pressure-temperature limitations. The LTOP enable temperatures have increased to 267'F for cooldown and 302'F for heatup for the 10 calendar year time period. These temperatures represent a 12 F and 7'F change in the LTOP enable temperatures for cooldown and heatup. For 40 calendar years of operation, the LTOP enable temperatures have been identified as 281'F and 315'F for cooldown and heatup. The heatup LTOP enable temperature, which defines the minimum 'temperature at which shutdown Page 19 of 21
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cooling can be isolated from the RCS, is quite large given it is approaching the shutdown cooling design temperature of 350'F. This may impact operations in the future, especially if an extended service life is pursued.
The increased shift in the pressure-temperature limits can have a significant impact on other LTOP parameters such as the heatup and cooldown rate limits over specified temperature intervals, and the administrative controls used to limit the effect of the mass and/or energy addition transients analyzed for LTOP.
Based upon the pressure-temperatur'e limit analyses provided within this report, no life limiting vessel integrity issues are anticipated to exist during the 40 calendar year design life of the reactor pressure vessel.
However, heatup and cooldown-rate or administrative restrictions need to be examined to ensure compatibility between the pressure-temperature limits and the LTOP analyses.
6.0 REFERENCES
(1) Code of Federal Regulations, 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," January 1988.
(2) Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," January 1988.
(3) ASME Boiler and Pressure Vessel Code Section III Appendix G "Protection Against Non-Ductile Failure," 1986 Edition.
(4) USNRC Regulatory Guide 1.99 Revision 02, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
(5) Instruction Manual for Reactor Vessel-Arizona Nuclear Power Project Palo Verde Units 1, 2, and 3, C-E Book No. 78173/79173/65173, dated August 1978.
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(6) ASME Boiler and Pressure Vessel Code Section III Appendix I, "Design Stress Intensity Values, Allowable Stresses, Material Properties, and Design Fatigue Curves," 1986 Edition.
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Table 1 PALO VERDE UNIT 1 HATERI ALS ART 'F Location ID No. CF* Cu I Ni %
~14 T ~34 T ~14 T ~3'4 T Lower Shell Plate H-4311-1 26 .04 .65 -10 35 28 55 H-4311-2 20 .03 .62 -40 10 H-4311-3 20 .03 .64 -20 30 Intermediate Shell H-6701-1 44 .07 .66 30 83 114 Plate H-6701-2 37 .06 .61 40 102 90 116 103 H-6701-3 37 .06 .61 40 102 90 116 103 Intermediate Shell 101-124 35 .07 .03 -50 10 37 Long Seam Welds**
Lower Shell 101-142 28 .04 .04 -80 -32 -39 -19 Long Seam Welds**
Intermediate/Lower 101-171 34 .05 .07 -70 -20 15 Girth Seam Weld**
- CF = Chemistry Factor determined by Regulatory Guide 1.99 Rev. 2.
- Two sets of chemistry data were available for the welds. The more limiting set was used for conservatism.
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Table 2 PALO VERDE UNIT 2 HATERIALS ART 'F Location IO No. CF* Cu 'I. ~i X RTRDT
~14 T ~34 T ~14 T ~34 T Lower Shell Plate F-773-1 20 .03 .67 10 45 39 60 54 Plate F-773-2 26 .04 .64 38 65 F-773-3 31 .05 .66 -60 -15 10 Intermediate F-765-4 20 .03 .67 -20 29 22 Shell Plate F-765-5 20 .03 .65 10 44 38 52 F-765-6 26 .04 .67 10 54 46 74 64 Intermediate Shell 101-124 34 .06 .04 -60 -3 -14 23 10 Long Seam Welds**
Lower Shell Long 101-142 .09 .04 -80 -15 15 Seam Welds**
Intermediate/Lower 101-171 27 .03 .07 -30 16 36 28 Girth Seam Weld**
- CF = Chemistry Factor determined by Regulatory Guide 1.99 Rev. 2.
- Two sets of chemistry data were available for the welds. The more limiting set was used for conservatism.
0 Table 3 PALO VERDE UNIT 3 MATERIALS ART 'F Location ID No. CF* Cu I Ni % RTR>T
'F 8 EFPY 32 EF Y
~14 T ~34 T ~14 T ~34 T Lower Shell Plate F-6411-1 26 .04 .64 -40 25 17 Plate F-6411-2 26 .04 .65 38 65 57 F-6411-3 26 .04: .66 -60 -22 5 -3 Intermediate F-6407-4 .04 .62 -30 14 34 24 Shell Plate F-6407-5 31 .05 .61 -20 33 23 49 38 F-6407-6 26 .04 .61 -20 24 16 44 34 Intermediate Shell 101-124 26 .03 .06 -50 -6 14 4 Long Seam Welds**
Lower Shell Long 101-142 31 '.04 .07 -50 26 17 Seam Welds**
Intermediate/Lower 101-171 .05 .07 -70 -20 15 Girth Seam Weld**
- CF = Chemistry Factor determined by Regulatory Guide 1.99 Rev. 2.
- Two sets. of chemistry data were available for the welds. The more limiting set was used for conservatism.
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Table 4 CONTROLLING HATERIALS AND THEIR ADJUSTED REFERENCE TEHPERATURES ART 'F Location Location ID No. RT 'F 8 EFPY 32 EFPY
~14 T ~34 T ~14 T ~34 T Palo Verde 1 Intermediate Shell H-6701-2,3 40 102 90 116 103 Palo Verde 2 Intermediate Shell F-765-6 10 46 74 Palo Verde 3 Lower Shell F-6411-2 45 38 65
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Table S COHPARISON OF CURRENT AND PREVIOUS ART PREDICTIONS FOR CONTROLLING HATERIALS ART 'F Approach of Plant Haterial ID Calculation* 8 EFPY 32 EF Y
~14 T ~34 T ~14 T ~34 T Palo Verde 1 H-6701-3 RG 1.99 Rev. 2 102 90 116 103 CE Design Curve 75 40 137
- Note that there was a) so a difference in assumed plate thickness, but the effect was only on the order of 2'F ART.
1 Table 6 RCS PRESSURE/TEHPERATURE CURVES 0 F C/D AT I/4 T LOCATION (8 EFPY)
ALFA RTrdt I/4t RATE Hs Pzct Ss Sn/Sy Kis Kir T-RTnd'; Pind Tind 0.399 0 3 '51 3295 28583 0.642 92.93 185.9 175.19 3!2'000
?".5.2 THICK)IESS RTndt 3/4t dT(sa!I) 3 '51 3170 27499 0.618 89,40 1?a.a 172.07 282. I
!i.!90 H/A 0.00 3. 214 2670 23161 Q.520 74 '5 !48.9 !56 '6 2500 ?67.0 dT(flu.)1/4t dTct 3.!98 2420 20993 0.472 67.!3 134.3 148.!5 2250 2.58. 2 44500 102 0.00 3. 198 2170 18824 0.423 60.20 120,4 138. 62 2000 248, 6 Nt I/4t dT(flu,)3/4t QT 3. 198 2070 17957 0.404 57.42 II4.8 134,41 1900 244. 4 0,387 H/A 0. 44 3,182 1970 l?089 0.384 , 54.38 108.8 129.47 1800 230, 5 Nt 3/4t dT(inst'rr) Kit 3. 182 1870 16222 0.365 51.62 103.2 124.66 1?OO 234. 7 HIA 8.0 Q,OO 3. 182 17?Q 15354 0.345 48.86 9' 119.49 1600 229 5
~
HEAH RADIUS dP(inst err)
~ 3, Ig? 1670 14487 0.326 46.10 92.2 !!3,91 !500 ?23.9 97.07 170.0 750'S!A 3.182 1570 I36!9 0,306 43.34 86. 7 !07.8? !400 217,8 120 ~ 0 750 PSIA 3.165 1420 !2318 0 '77 38,99 78. 0 97.01 1250 20' 3.165 1170 10149 0.228 32.!3 64. 3 75.48 1000 !85,5 3.165 1070 9282 0,209 29,38 Saw 8 64.55 900 !74,5 3.149 970 8414 0. 189 26.49 53.0 5'? BOQ Ibi.a 820 ?113 0. 160 22, 40 44.8 24.96 700 !35,0 3.149 7?Q 6246 0.14Q !9.67 39. 3 0.04 600 I:0.0 3 '49 670 5812 0.131 I 8 30
~ 3b.6 -lb.a9 550 93.!
3.149 645 5595 0.1?6 17,62 <<5 -')7 'n cQz 8') 8 3.149 620 5378 0.121 16,93 33 9
~ -39.37 Soo 70.6 3,!49 608 5274 0.!!9 lb.61 33. 2 -46,06 488 67,9
- 3. 149 S70 4945 O,ill 15.57 31,1 -7> n4 <<50 37 ~ I 3.149 545 47?8 0. 106 14.89 29.8 -98.88 425 I:, I
~ 149 5?5 4SII O.lol !4 '0 28,4 -!40.9z 400 30 ~ 9 3,149 370 32!0 0.072 10,1! ".0,2 ERP 250 cion
(
fI
'1 J'
ti
Table 7 RCS PRESSURE/TE)}PERATURE CURVES 10 F C/D AT 1/4 T LOCAT}OH (8 EFPY)
A}.FA RTrdt 1/4t RATE Nn Pact Sn SI/Sy Kin Kir T-RTndt Pind Tind 0 ~ 399 0 10 3. 251 3295 28583 0.642 92 93
~ 190. 1 !77.00 312S 282 ?~
TH}CKHESS RTndt 3/4t dT(wa) 3 ~ 251 3170 27499 0,618 89.40 183 ~ 0 }73.96 3QQO 279 2
- 11. 190 N/A 10. 90 3 ~ ?14 2670 23161 0.520 74.4S 153 ~ 1 159.31 2500 264.5 Sy dT($ }0.) 1/4t dTct 3. 198 2420 20993 0.472 6i ~ 13 138.S }50.81 2250 256,0 44500 }02 -4.79 3.198 " 2170 18824 0.4?3 60,20 124.5 141.57 2000 246.9
)}t 1/4t dT(f}u.)3/4t XdT 3.198 2070 17957 0.404 57.42 119.1 137.64 }900 242. 8 0.387 M/A 0. 44 ~
3.182 1970 17089 0.384 54,38 113.0 132.93 1800 2.8. }
)}t 3/4t dT(inst err)
~ Kit a.!82 }870 }6222 0.365 51,62 107.5 128,37 }700 233 6~
HIA 8' 4.22 3.182 1770 15354 0.345 48 86 F !01.9 123.48 1600 228. 7 NEAN RADIUS dP(inst.err) 3.182 15io 14487 0.326 45.10 96.4 118.22 1500 223.4 97 07 F 170.0 750 PS!A 3,182 !570 13619 0 '06 43.34 90,9 !12.S2 !40O 2!7 ' 120.0 750 PS!A 3 165
~ !420 123!8 0.277 38.99 82 ' 102.47 !2SO 207.7 3, 165 }170 }0!49 0 '28 32 AD }3 68.5 82.84 !000 !BB.Q 3.165 }070 '?282 0 '09 29.38 63.0 73.}0 900 !78 '
3 '49 970 8414 O.!89 ?6.49 Si. 2 51.12 800 !65.3 3.149 820 71!3 0.150 22.40 49,0 3'?. 48 700 !44.7 3.!49 720 6246 0.!40 }9 67
~ 43 6 20.0. 600 !?5.2 3.149 570 5812 0 '31 }8 30
~ 40,8 i.7i SSO la3eo 3.}19 64S 5595 O. 1'26 17. 62 39.5 O. 7? 5')c5 }05 9 ~
3.!49 620 5378 0.121 16,93 38.! - ,15 500 98.! 3, })9 608 527-'. 0. 119 ! 5. b 1 -1! ~ 27 488 >0 ~ 0~ 3.149 570 4945 O. 1}1 ! 5. 57 35 4
~ -26 '1 450 i9.0 3, }49 c4c 4728 0.!06 14.89 34. 0 -38.17 425 bi. 0 3.149 520 45!1 0 ~ 10! 14,20 3?.6 -52,65 400 52. 6 3.}49 370 3210 0.072 10.!l 24 ' EP.R 25Q ERR
'0 4
Table 8 RCS PRESSURE/TEHPEPATURE CURVES 100 F C/D AT 1/4 T LOCATI})H <8 EFPY) ALFA RTndt !/4t RATE )}a Pact SA SA/Sy Kln Kir T-RTndt ?ind Tind 0,399 0 }00 3. 251 3295 28583 0,642 92 ~ 93 228 ~ 1 191.43 3125 25 TH ICKNESS RTndt 3/4t dT<wai}) 3,?51 3170 27499 0.618 8?. 40 221 ' '88 '?7 3000 251 ~ 0 11.190 H/A 108 ~ 97 3.214 2670 23161 0.520 74. 45 177.44 2500 239.5 191..'76,5 Sy dT! flu.) }/4t djct 3.!98 2420 20993 0 '72 67.13 171.00 2?50 2.3,! 44500 102 -47.95 3. 198 2170 }8&24 0.423 60. 20 162.5 164 '9 2000 2?6.3 Ht 1/4t dT(flu.) 3/4t XdT 3.!9& ?070 !7957 0.404 57,42 157.1 !6!.42 1'?00 223,5 0.387 H/A 0.44 3 'S2 1970 17089 0,384 54.38 }51.0 158.}2 1800 220.2 Ht 3/4t dT}inst. err) Kit 3.}S2 }870 16222 0.365 51,62 145.4 154.98 1700 217.0 H/A 8.0 42,2} 3.}82 1770 15354 0.345 48 AD &6 139.9 151 '0 }600 2}3. 8 HEAtH RADIUS dPiinst.err) 3.}82 1670 !4487 0.326 46.!0 134.4 148.?5 !500 210.3 97.07 170 ~ 0 750 ?SIA 3.!S2 1570 13519 0,305 128 ' !44.ol 1400 206.7
!20.0 C = 750 PS!A 3.155 1420 12318 0.277 38.99 }20e2 138,48 }250 200.5
- 3. 165 1170 10149 0.228 3?, 13 '05.5 127.5! 1000 189.5 0
3, 165 107Q 9?82 0.209 29. 3S !0!.0 122.59 900 !84.6
- 3. 149 '?70 8414 89 26. 49 95. 2 !!7.00 800 }79.1
- .149 820 ?}!3 0.!AO 22,40 87,0 }OS ~ 20 ?00 !?0 '
3,!49 7?0 5246 140 19.57 81.5 10!I65 600 153 ~ 7 3.149 620 5378 0.)2} }6,9 lb. } 94.40 500 }56 ~
~
'. 4o 608 5274 0. !19 15. 61 75. 4 93.47 488 ~ oo ~ w 3.!49 570 4945 0.}il }5 57 73 3~ 90.47 450 ~ ~ 2t I 3.149 520 45!1 O.!Ol 14.20 70. 6 86.30 400 148. 4 3.!49 470 4077 0,092 12.84 57.9 8!,Bb 350 }43.9
- 3. 14'? 370 3?1Q 0 '72 10.!I 62.4 ?2,03 250 134. I 270 2342 0.053 7. 37 57,0 50,55 !50 3 ~ 149 !20 !041 0.023 3. 28 48. 8 3S.70 0 }00.8
1 Table 9 RCS PRESSURE/TENPERATURE CURVES 0 F C/9 AT I/4 T LOCATION (32 EFPY) ALFA RTndt I/4t RATE Nn Pact Sn Sn/Sy Kin Kir T-RTndt Pind T:nd 0.399 0 0 3,251 3295 28583 0.642 92.93 185.9 175.19 3125 299. 2 THICKNESS RTndt 3/4t dT(walt) 3. 251 3170 27499 0.618 8'?. 40 178 ' 17?.07 3000 29A. 1 11.190 H/A 0.00 3.214 2670 23161 0.520 74. 45 148.9 156.96 2500 281.0 Sy dT(flu.) 1/4t djct 3.198 2420 20993 0.472 67 ~ 13 134.3 148.15 2250 ?1?e 2 44500 lib 0,00 3. 198 2170 18824 0.423 60 20
~ 120.4 138 '2 "
2000 "bz 6 Nt 1/4t dT(f!u,)3/4t XdT 3 ~ 198 2070 }7957 0.404 57 ~ 42 114.8 134.41 1900 258. 4
- 0. 387 H/A 0. 44 3. 182 1970 17089 0.384 54. 38 108.8 129.47 1800 253.5 Nt 3/4t dT(inst err)
~ Kit 3 ~ 182 1870 16222 0.365 S!,62 103.2 124.66 1700 248 ~ 7 H/A 8.0 0. 00 3. 182 }770 15354 0.34S 48 '6 97.7 119.49 1600 ?43.5 NFAH RADIOS dP(inst.err) 3.!8? 1670 14487 0 '26 46.10 92 ' 113,91 1500 237 '
170.0 750 PS!A 3.182 1570 13619 0.306 43.34 86,7 107,8? 1400 2aliu
!20.0 750 PSIA 3.!6S l420 !2318 0.277 38.99 78.0 97,0! 1250 ?21.0 3.!65 }170 }0}47 0.228 32 '3 64 ' 75.48 1000 iAA 5 3.165 !070 9?82 0.209 29.38 58. 8 64.55 900 !88.5 3 '49 970 8414 0.189 ?6.49 53.0 50.82 Boo 174.8 3.149 820 7! 13 P. tbp 2?,40 44,8 24 '6 700 !49.0 3.}49 720 6246 0. !40 !9.67 39.3 0.04 &00 I?a, 0 3 '49 62Q 5378 0,121 }6,93 33,9 -39.37 500 84. 6 3,149 608 5274 0.119 16.61 33.? -46.06 488 77 A 3.149 570 <<A4S O,lll !5.57 31,1 -72.94 450 5!.1
- 3. 149 545 4728 0. 106 14.89 29,8 -98.88 425 ?5.!
3,}49 52Q 4St 1 O. 101 !4.20 28,4 -140.94 400 lb G 3 '49 495 4294 P. 096 }3. 52 27,0 -267.44 37S -!43.4 3,149 42Q 3643 0.082 !},47 22 ' ERR 300 RR 3.149 395 3427 0 '77 10.79 21.6 ERR 275 ERR
I ( Ci
0 Table 10 RCS PRESSURE/TFHPERATURE CURt/ES 10 F C/0 AT 1/4 T LOCATION <32 EFPY) ALFA RTndt 1/4t RATE Hn Pact Sn Sn/Sy Kin Ki r T-RTndt Find Tf nd 0.399 0 10 3.251 3295 28583 0.&42 92.93 190. 1 li7.00 3125 296,? THICKNESS RTndt 3/4t difwall) 3 '51 3170 27499 0,5}8 89.40 183.0 !73.96 3000 293.2 11,190 N/A }0.90 3.2}4 2610 231&1 0 '20 74 '5 153.} }59.31 2500 27":.5 Sy dT(I}u.)f/4t dTct 3 '98 2420 20993 0.472 67.13 138 ~ 5 150.8! 2?cp ?70.0 44500 }16 -4.7'9 3. 198 2170 18824 0.423 60.20 124.6 141.57 2000 250.9 Ht 1/4t dT (I:".', ) 3/4t Xdj 3.19S 2070 17957 O.4O4 57,42 11'?.1 !37.64 }900 ?55.8 0.3Si N/A 0 '4 3,182 }970 !7089 0.384 54.3S 113. 0 !32 '3 1800 252.1 Ht 3/4t dT(inst.err) Kit 3 '82 1870 16222 0.355 5! ~ 62 )07.5 128.37 li00 247.6 N/A 8.0 4.22 3 '82 17io !5354 0.345 48.86 101.9 123 ~ 43 1&00 242.7 HEAN RADIUS dP!inst.err) 3,18? 1670 )4487 0,326 46.10 96.4 118,22 1500 ?dl.0 97 07 F !70.0 i50 PS!A 3.182 1570 13619 0.306 90. 9 3
~
)<> !000 ?0?,0 3.165 1070 9 8? 0 '09 29.38 53 0
~ 73I 0 900 !9:.3 3.14'? 970 84!4 0 'S9 ?6. 49 57.2 6),12 600 160.3 3.149 820 7!13 0.!60 22. 40 49,0 39,46 iop }cn 3.}49 720 5246 0.140 19. 67 43.6 20.03 600 3.!49 620 537S 0,!?1 lb. 93 38.1 -7.15 cPO !1?.:
3.!49 508 52i4 0.!!9 1& ~ 61 37.4 -!f,?7 466 }Pi.9 3.149 Sip 4945 0 ~ ili f5.57 35.4 -26.?) 450 93IO 3.!49 545 4728 0.106 }4. 89 34.0 -38.17 6}.O 3.149 520 4511 0.101 14.20 32.6 -52,55 400 6&a& 3.}49 495 4294 0 '96 13 52
~ 31.3 -71 00 F 0 I 'a 48 3.149 420 3643 0.082 1 l. 4i 17 -240.49 .DO }2) 3.!49 395 3421 0. 077 !0.79 25. 8 ERR ERR
I JI If I r, I 9
Table ll RCS PRESSURE/TEHPERATURE CURVES 100 F C/9 AT 1/4 T LOCATION (32 EFPY) ALFA RTndt 1/4t RATE )fn ?act Srr Sa/Sy Kin Kir T-RTnd'. Pind Tind 0 '99 0 100 3.251 95 285S3 0.642 92.93 22S.} 31?5 267. 5 THICKNESS RTndt 3/4t dT(rrall) 3.251 3170 27499 0.618 89.40 221,0 188.97 3QQO 255. 0 I 1, 190 N/A }08,97 3.214 2670 23161 0.520 74 '5 191.1 177 ~ 44 2500 253> 5 Sv dT(fin.)}/4t dTct 3. 198 2420 20993 0.472 67.}3 176.5 }71.00 2250 247 ~ ! 44500 116 -47.95 3 ~ 198 2170 }8824 0.423 60.20 162.6 164 ~ 29 2000 240,3 Nt 1/4t dT(..'Iu.)3/4t XdT 3.198 20ip 17957 0.404 57.42 157 ~ 1 !61.42 1900 237.5 0 '8i N/A 0,44 3,182 1970 170S9 0.384 54.38 !51.0 158.12 }BQQ 234.2 l(t 3/4t dT(inst'rr) Kit 3.182 !Bip 16222 0.365 51.62 }45.4 t(EAN RAOIUS N/A dP(inst.err) 8,0 42 '! 3.!82 1770 r 354 0.345 48.86 139.9 154 ~ 98
!51.70 1700 1600 23}.0 227.8
- 3. 182 1670 14487 0.326 46.10 '34.4 148.25 1500 224.3 97,07 170 0 ) 750 PSIA 3.! 8? 1570 }36!9 0 '06 43 '4 128 ' 144.61 !400 220.7
}'rp.p < = 750 PS!r" 3.! 65 }420 !23}E 0.277 38.99 !20.2 138.48 !250 214. 5 3.165 1170 }0}49 0.228 32 '3 106 ~ 5 127.51 }000 203.6 3.165 1070 9282 0.209 29.38 101 ' !22.59 900 198,6
- 3. !49 97G 8414 0.189 26.49 95. 2 1:7,00 BCO }03 3.149 820 i113 0.160 22.40 Si.p 108.20 700 rv>>I) i20 6246 0,140 }9.6": 81.'. 101.65 600 17i.i 3 '49 620 5378 0. 121 lb.93 75.1 94.40 5CO 170,5 3.r49 508 5274 0.!ro lb.6r 75. 4 93.47 488 }o9.5 3.149 570 4945 0 11'5 57 73 3
~ 90.47 450 }56 5 3 '49 520 45!1 0.}Q} 14.20 70. o 86.30 400 t62.4 3.!49 470 4077 0.09? t2.84 67. 9 S}.86 350 3.149 370 3210 0 '72 10.1} o2.4 72 '3 ngp 148.!
3e 149 270 2342 0.053 7,37 57. 0 60.55 150 135 '
- 3. 149 120 1041 0.023 3.2S 48 ~ 8 38,70 0 1}4.8
f 'I QY I I i V
0 Table 12 RCS PRESSURE/TEHPERATUPE CURVES 10 F H/U AT 1/4 T LOCATION (8 EFPY) ALFA RTndt 1/4t RATE Nn Pact Sa Se/Sy Kie Kir T-RTndt Pied T}no 0 '99 G 10 3.251 3295 28583 0.642 92.93 181.6 173.34 3125 ?88. 1 THICKNESS Rindt 3/4t dT(wa}l) 3.251 3170 27499 0.618 8'? ~ 40 174.5 )70.}3 3000 2C4e9 11.190 N/A 10.90 3,214 2670 23161 0.520 74. 45 !44,7 154 '4 2500 ?69.3 Sy dT(flu.)}/4t dTct 3.198 2420 20993 0.472 67.13 130 ~ 0 !45.39 2250 260.2 44500 102 4 '9 3,198 217Q 18824 0.423 60.20 1}6.2 135.44 2000" 250.2 Nt 1/4t dT(f!u.)3/4t XdT 3.}98 2070 17957 0.404 57.42 110,6 131.02 1900 245.8 0,387 N/4 0.44 3.182 }970 17089 0.384 54.38 104 ' )?5.82 1800 240 6 F Mt 3/4t di(inst err)
~ Kit 3,!82 !870 16222 0.355 51.62 99.0 120.75 !700 235,5 N/A 8 -4. 22 3.182 1770 15354 0.345 48.86 93 ' 115.26 1600 230. 1 REAM RADIUS dP(inst.err) 3.182 1670 }4487 Q.326 46. 10 88.0 109 '0 1500 224 '
97 07 F !70.0 750 PSIA 3.182 1570 13619 0.306 43.34 82 ' 102 78
~ 1400 217.6 120. 0 750 PSIA 3 '65 1420 12318 0.277 38.'?9 73 ' 9'7 !250 i05.9 3.165 1170 10149 0.228 32.13 60. 0 67.24 1000 182 ~ 0 3 ~ 155 1070 9282 0.209 2?,38 54.5 54,7? '?QO 169.6 3 '49 970 8414 0.189 26 49~ 48.8 38,70 800 )u0 ~ 5 3 ~ 149 820 71!3 0.150 22 ~ 40 40.6 5,54 700 21 ~ ~)
3 '49 770 6680 0.)50 21. 03 37.8 -8.68 550 !06 F 1 72G 6246 Q.!40 '9 ~ 67 35. } -?8. 24 600 e c~ u
~
~
3.! 49 570 58!2 0.13! )8.30 32.4 -55.54 550 59 1 ~ 620 53/8 Q.}21 !6.93 29.6 -!0!.80 500 !3,0 3~ !49 608 5i74 O. !19 }6.61 29.0 -!19.7! 488 3.149 570 4945 O.!l} 15.57 25.9 -3}2.":7 4SO ~
~
s i ~ II 3.!49 520 45!! 0.10} 14.20 24.2 ERR 400 ERR 3,149 470 4077 0.092 12 ~ 84 21.5 .."F.R 350 tnR 3.149 420 3643 0.082 11.47 18.7 ERR ~00 ERR
I Table 1 3 RCS PRESSURE/TEMPERATURE CURVE 10 F }}/U AT 3/4 T LOCATION (8 EFPY) ALFA RTndt I/4t RATE Ha Pact Sa Sa/Sy Kin Kir T-RTndt Pind Tind 0.399 N/A 10 3 ~ 251 3295 28583 0.642 92.93 188 ' 176.47 31?5 284. 6 THICKNE."S RTndt 3/4t dT(aa}I) 3.251 3170 27499 0.618 89 '0 }S},8 173,41 3000 281.5
}1.190 0 10.90 3 '14 2670 23161 0.520 74 '5 151,9 158 '3 ?500 266.8 Sy dT(fILI.) I/4t, dTct 3.198 2420 20993 0.472 67.13 137.2 150.04 2250 258.2 44500 N/A 10.13 3.198 2170 18824 0.423 60.20 123.4 140,78 2000 248.9 Mt }/4t dT(f!u.)3/4t Xdj 3.}98 2070 17957 0.404 57.42 !17,8 136.70 !900 244.8 H/A 90 0.93 3.182 1970 17089 0.384 54.38 1}1.7 131.93 }BOO 240 '
Ht 3/4t dT(inst.err) Kit 3,182 1870 16?22 0,365 5!.62 }06.2 127.30 1700 235.4 0.273 8.0 2,98 3.182 1770 15354 0.345 48,86 100.7 !22.33 shoo 230.5 HEAH RADIUS dP(inst.err) 3.182 1670 14487 0.326 46,10 95.2 1}6,98 }500 225.! 97.07 170.0 750 PSIA 3.182 1570 !36}9 0.306 43,34 89.7 111.17 1400 219.3
}20.0 750 PSIA 3,}65 1420 12318 0.277 38.99 81.0 100.9} }?ro 209.O 3 '65 1170 to}49 0.228 32.13 &7.2 80.75 }000 188,9
- 3. 165 1070 9282 0.209 29.38 61,7 70 '9 900 '78 8 3 t49
~ 970 8414 0.189 26.49 55.0 58 '4 800 !66 4 ~
3,149 820 7113 0.160 2?.40 47.8 35.51 7OO !43.6 3.149 720 6246 0.';40 19.&7 42 ' , 14 ~ 72 6('0 122.9 3 '49 670 58!2 0.!3! 18.30 1.38 %co 10 5 3 '49 645 5595 0,126 }7.&2 3S.? -b. 40 525 'o'i.7 3.}49 620 5378 0. !21 1& ~ 93 3&. 8 -!5,!8 500 90I2 3.149 ooS :?74 0.}!9 lb.b1 36.2 -19,8? 4SS 38. 3
- 3. 149 570 4945 0.!!I 15.57 34,1 -37.01 450 l or a 3.}49 5 0 51! 0.101 !4.20 3' -n9.}4 400 3910 a ~ )49 '20 1908 0,043 6.01 15 ~ 0 ERR 100 .:RR 3.}49 170 0.033 4.&4 !2.3 ERR 50 RR
t Table 14 RCS PRESSURE/TENPERATURE CURVES 20 F H/U AT 3/4 T LOCATION (8 EFPY) ALFA PTndt 1/4t RATE Na Pact Sa Sa/Sy Kia Kir T-RTndt Plnd Tind 0.399 N/A 20 3.251 3295 28583 0.642 92.93 191.8 177.73 3125 275 0 ~ THiCKNESS RTndt 3/4t dT(aa! I ) 3.251 3170 27499 0.618 89.40 184.8 174,72 3000 293.0 11 ~ 190 0 21.79 3.214 2670 2316! 0 520 74 45 154.9 160.25 2500 278.5 Sy dT(flu.)1/4t dTct 3.198 2420 20973 0 '72 67.13 140.2 !51 87F
) c'0 270. l 44500 N/A 20,27 3.198 2170 18824 0.423 60.20 126 ~ 3 142.88 2000 261.1 Nt I/4t dT(flu.)3/4t XdT 3.198 2070 17957 0.404 57.42 120,8 138,92 1900 257,2 N/A 90 0.93 3.182 1970 17089 0.384 54 '8 114.7 134. 31 1800 252.6 Ht 3/4t dT(inst. err) Kit F 182 1870 16222 0.355 51.62 109.2 129.84 1700 248.1 0.273 8.0 5.96 3.1S2 1770 15354 0.345 4S.B6 103.7 125,05 1600 243 '
)IEAN RADIUS dP(inst.err) 3.182 1670 14487 0.325 46.10 98 ' 119.92 1500 238.2 97.07 170 ' 750 PSIA 3 '82 1570 13619 0.306 43.34 92.6 114 ~ 36 1400 232.6 120,0 < = 750PSIA 3.165 1420 12318 0.277 39.99 83.9 104 '0 1250 222 ' 3 '65 1170 10149 0.228 32.13 70.2 85 65
~ lOCO 203.9
- 3. '!65 1070 9282 0.209 29.38 64.7 76 '3 900 194.6 3 '49 970 8414 0.189 26.49 58 9 64.94 800 183.?
- 3. !49 820 7113 0.160 22.40 5Q,7 44.66 700 !6":.9, 3 '49 7?0 5246 0 '40 19.67 45,3 26.82 &00 $
iv 4% ~ ~I 3.ld9 3.149 570 58I? Q.!31 !8.30 42 F 6 !5 '! 550
~
34 m! 5?0 537S 121 Ig 93 39.8 2 ~ 70 500 I 1.0 3> 149 08 5274 0,!!9 16,61 39.2 -0.86 4:8 !!7.4 3.!49 570 . 4945 O.ill 15.57 37.1 -!3 '1 4'.0 !04.9 5?0 45!1 0 IOI ld 20
~ 34.4 34e73 40~) 84 5
~
470 4077 0. )92 I?.84 31 F 6 -65.54 35) 52. 7 3~ !49 420 3543 0.092 11.47 98 0 -122 55
~ 300 -4.4 370 3?I<) 0.072 10.11 ERR 250 ERR
Table 15 RCS PRESSURE/TEHPERATURE CURVES 40 F H/U AT 3/4 T LOCATION (8 EFPY) ALFA R:ndt 1/4t RATE }(n Pact Sm Sn/Sy Kin Kir T-PTrdt Pind Tind 0.399 N/A 40 3.251 3295 28583 0 '42 92.93 }97.8 180.17 3125 318.7 THICKNESS RTndt 3/4t dT(wa}l) 3 ~ 251 3}iO 27499 0.618 89. 40 }90.7 177 ~ 27 3000 315.8 11.190 0 43 ~ 59 3.214 2670 23161 0.520 74. 45 160.S lou.aS 2500 301.9 Sy dT (f I u. ) 1/4t dTct 3,}98 2420 20993 0.472 67,13 146.2 155,40 2250 293.9 44500 }(/A 40.54 3.198 2}70 18824 0,423 60.20 132.3 146 89 F 2000 285 ' Ht 1/4t dT(flu.)3/4t XdT 3 ~ 198 2070 17957 0.404 57,42 126 ~ 8 143,!6 1900 ?8' N/A 90 0.93 3.l82 1970 17089 0,384 54,38 120 ' 138.83 !BOQ ?77.4 Ht 3/4t dT(inst.err) Kit 3.182 1870 16222 0.365 51.62 115.1 134.65 !700 273.2 Q.273 8~ 0 11.91 3.182 1770 }5354 0.345 48.86 109,6 130.20 !600 268.7 BEAN RADIUS dP(inst, err) 3 AD !82 !670 14487 0.326 46,10 104.! 125.44 1500 264.0 97.07 170,0 750 PS!A 3.!82 1570 13619 0 '06 43.34 98.6 120.34 !400 ~58.9
}?0.0 i50 PS!A 3.!65 1420 12318 0.277 38.99 89.9 11!.44 1?50 250.0 3.165 1!70 10149 0.228 32.13 76.2 94,52 }000 233,1 3,!65 1070 9282 0 '09 29,38 70.'7 86.39 90Q ?24.9
- 3. 149 970 8414 0. 189 26.49 64 ~ 9 76.66 800 215 '
3.149 820 7113 0. 160 2?.40 56.7 59.97 i00 198.5
- 3. 149 720 6246 0 '40 19.67 51.2 46.07 600 :84. 6 3.}49 620 5378 0,}2! lb.93 45.8 28 '4 500 167.?
e 3. 149 3.!49 3.!49 3,}49 3.}49 608 520 470 420 395 5274 4511 4077 3643 3427 0,!19 0.!0! 0.09? 0.08? 0.077
}6.61
- 14. 20 12 ~ 8>>
1!.47
- 10. 79 45.1 40 bleb 34.9 33.5 25 ??
5 '6
-}Q.?8
-30.3S
-43 F 16 488 400 350 300 2i5
}64.8
!4 .S 1".8,3
}03.?
95.4 3 }49 370 3?10 ') 072
~ 10 ~ !1 32. 1 -58,85 250 79.7 3 '49 355 3080 0 '69 g.iQ 31.3 >OI 30 33 68r~
Il 11
Table 16 RCS PRESSURE/TEMPERATURE CURVE=" 100 F )I/O AT 3/4 T LOCATION (8 i":FPY) ALFA PTrdt I/4t RATE )I Pact Sn Sn/Sy Kin Kir T-RTndt Pind Tind O. 399 N/A 100 3 ~ 251 3295 28583 0 '42 92 93
~ 21S.& 187.03 3125 a86>>4 THiCKNESS PTndt 3/4t dT(xaii) 3 ~ 251 3170 27499 0.618 89.40 208.6 184 ~ 41 3000 Ddt I
!1.190 0 108.97 3.214 '670 23161 0.520 74.45 178.7 172.01 2500 37! 4 ~
Sy dT(f!8 ) I/4t
~ dTct 3.198 2420 20993 0.472 67.13 164.0 165.02 2250 364.4 44500 N/A 101.34 3 ~ 198 2170 18824 0.423 60,20 150.2 157.63 2000 357.0 Nt I/4t dT(flu.)3/4t XdT 3.198 2070 17957 0,404 57.42 144.6 154 '0 1900 353 '
N/A 90 D. 93 3,}82 i970 17089 0.384 S4.38 138 ' 150 '4 1800 350.2 Ht 3/4t dT(inst>>err) Kit 3.182 1870 16222 0.365 51 62 133 0 147.3S 1700 346.7 0.273 8.0 29.78 3 '82 1770 15354 0.345 48 86 F 127.5 }43.&7 1600 343.0 )IEAN RADiUS dP(inst.err) 3 '82 i&70 i4487 0.326 46,10 122.0 13'?. 78 }500 339.1 97,07 i70.0 75Q PSIA 3.182 1570 }3619 0.306 43 '4 116 ~ 5 135.66 1400 335.0 120.0 75Q PSIA 3.165 1420 123i8 0.277 38,99 107.8 128.&3 1250 3ZB.Q 3.165 1170 10149 0.228 32.13 94.0 1}5 '2 1000 3i5.2 3.!65 }070 9ZBZ 0.209 29,38 88,5 109.94 900 309.3 3.149 970 8414 0.189 26.49 82,8 !03 17 F 800 302>>5 3 '49 820 7113 0.160 22.40 74,6 92 '6 700 294 3.149 72D 6246 0.140 !9.67 69.i 83.89 &00 "83 ~ 3 '49 620 5378 0.12! 16.93 &3.6 74.36 500 ?73. 7
- 3. 149 608 5274 0.!!'? 16.61 63.0 73.12 IBB
~l '>>~g 3.149 5?0 451! O. 10! 14.20 58.2 63.30 262. 6 3 '49 420 3643 0.082 11,47 S2.7 50.12 3>>)0 ?49. 5 3,149 370 3210 0. 072 IO.'.I 50.0 42.45 250 24}.8
;.149 320 277& 0. 062 8. 74 47.3 33.8I 200 I.kv ~ I
>>>> PJ 3, i49 270 ? 42 0. 053 7.37 44.5 23.?4 150
' 0 3 '49 220 1908 0. 043 6,0! 41.8 12.42 !00 211. 3
I II 0
Table 17 RCS PRESSURE/TEMPERATURE CURVES 10 F H/U AT I/O T LOCATION (32 EFPY) ALFA RTndt I/4t RATE Mc Pact SA Sc/Sy Kin Kir T-RTndt Pind Tind 0,399 0 10 3. 251 3295 28583 0.642 92.93 181.6 173.34 3125 302.1 THICKNESS RTndt 3/4t dT(~all) 3. 251 3170 27499 0.6!8 89.40 174.5 170.13 3000 298.9 11.190 N/A 10.90 3 ~ 214 2670 23161 0.520 74 '5 !44,7 154.54 2500 283.3 Sy dT(flu.)l/4t dict , 3,1"".8 24?0 20993 0.472 67,13 130.0 145.39 2250 274.2 44500 4,/9 3~ io8 2170 ISB?4 0 '23 60.20 116.2 135.44 2000 264.2 Mt I/4i dT(f)u.)3/4t 7dT 3,198 2070 17957 0.404 57,42 110 6~ 131 '2 19OQ ?59.8 0.387 3 N/A 0. 44 3 'S2 1970 !i089 0.3S4 54.38 104 ' 125.82 ISOO 254.o Mt 33/4t dT(inst.err) Kit 3.182 1870 16222 0.355 51.62 99 0~ 120.75 1700 249.5 N/A 8 -4. 22 3 '82 1770 !5354 0,345 48.86 o3.5 I!5'6 1600 244.! MEAN RABIUS ";P(inst.err) ae 1co v>> 1670 14487 0.325 )b !0 e 88. 0 IO9.30 !500 238.1 97.07 170.0 750 PSIA 3.182 1570 136)9 0.305 43.34 8? ~ 5 !02,78 1400 23!.6 120.0 750 PSIA 3.165 1420 12318 0 '77 3S ~ 99 73 ~ 8 9! Oi
~ 1250 2!9.9 3 ~ 165 1170 10149 0.228 32.13 60. 0 57.24 1000 196.0 3 ~ 165 1070 9282 Q~ 209 ?9.38 54,5 54,79 900 !83.6 3.149 970 8414 0. 189 26. 49 48.8 38a70 800 167.5 3 '49 820 7113 0,!60 22.40 40.6 6.54 i00 135.3 3.149 800 6940 0.156 21,85 39.5 0.85 680 129.6 3.149 770 6680 O,ISO 21.03 37. 8 -8.68 6SO 120.1 3.149 720 6246 0 '40 19.67 35.1 -28 '4 o00 100.5 3.149 620 5378 0,121 16 ~ o3 ?9.6 -101.80 500 27.0 3.!49 608 5274 0.119 16,61 29.0 -119.71 488 9.1 3.149 570 494S 0.111 !5.57 ?6.9 -312.27 4SO -183.5 3.149 520 4511 0. 101 14. 20 24.2 ERR 400 ERR
- 3. 149 470 40ii 0.092 12,84 21.5 ERR 350 ERR 3.149 420 3643 0.082 11.47 IS.7 ERR 300 ERR
I Table 18 PCS PRESSURE/;EMPERATURE CURVES 10 F H/U AT 3/4 T LSCATIQN (32 EFPY) ling Tird ALFA RTndt }/4t RATE Ma Pact Sa Sn/Sy Kin Kir :-RTndt Pind 0.399 N/A 10 3.251 3295 285S3 0.642 92m 93 188.8 }76.47 3!25 297.6 THICKNESS RTndt 3/4t dT(nail) 3.251 3170 27499 0 '18 S9.40 181.8 173 '1 3000 294.5 190 0 10. 90 3 ~ 214 2570 23!61 0 '20 74.45 151,9 }S53 2500 279.8 Sy dT(fju.)j/4t dTct 3.198 242Q 20993 0 '72 67 ~ 13 137.2 150 '4 2250 271.2 44500 N/A }0.13 3.}98 2170 18824 0.423 60 20
~ 123.4 140.78 2000 261.9 Mt I/4t dT(fin.)3/4t XdT 3 '98 2070 17957 0 '04 57 '2 117,8 136.70 1900 257.8 N/A 103 0.93 3.182 1970 }70S9 0.384 54.38 1!1.7 131.93 1800 253.1 Mt 3/4t dT(inst,err) Kit 3.182 1870 16222 0 '55 51 '2 106.2 127.30 1700 248.4 0,273 8.0 2.98 3.182 1770 15354 0.345 48.86 100 ' 122 33 F 1600 243.5 MiAN RADIUS dP(inst,err) 3. 182 1670 }4487 0.326 46.10 95.2 116.98 1500 23S.I 97.07 170.0 750 PSIA 3,182 1570 13619 0.306 89.7 111.17 }400 232 '
120.0 750 PSIA 3.}65 1420 12318 0.277 38 '9 81.0 }00.91 1250 222 ' 3.165 1170 !0149 0.228 32.13 67 ' 80 '5 1000 201. 9 3.165 }070 9282 0.209 29 ~ 38 61.7 70.69 900 191.8 3 '49 970 8414 0.189 26.49 56.0 58.24 SQC 179.4 3.149 820 71!3 0,160 22.40 47 ' 35.51 700 156.6 3 '49 720 6246 0.14Q 19.57 42.3 14.72 600 135.9 3.!49 620 5378 0.12l 16,93 36.8 -15.18 500 106.0 3.!49 608 5274 0.119 16 61~ 36.2 -19.82 4CS }0}.3 3.149 580 5031 0,113 15 '4 34 ' -32 06 F 460 89.1 3.!49 560 4858 0.109 15.30 33 ~ 6 "42 '5 440 78.8 3,149 550 4771 0,107 !5.02 33 ' -48,13 430 73,0 3.149 540 4&84 0.!05 }4. 75 32.5 "54.45 42Q 65.7
- 3. 149 520 451} 0,101 14 ~ 20 3!.4 bc.}4 400 52.0 3.149 470 4077 0.092 }~,84 28.7 -131.19 350 -10.1
J l 5 P ' K I l y l 1 II I
Tabie 19 RCS PRFSSURE/TEHPERATURE CURVES 20 F H/U AT 3/4 T LDCATION (32 EFPY) ALFA RTndt 1/4t RATE Hn Pact Sn Sn/Sy Kin Kir T-RTndt Pind Tind 0.399 N/A 20 3 ~ 251 3295 28583 0.542 92.93 191.8 177.73 3!25 309.0 THICKNESS PTndt 3/4t dT(wal I) 3.251 3170 O.618 89.40 184.8 174.72 3000 306.0 11.190 0 21.79 3.214 2670 23161 0.520 74 '5 154, 9 160 ~ ?5 2500 291,5 Sy dT(f I u, ) 1/4t dTct 3.198 2420 20993 0.472 67 AD !3 140.2 151.87 2250 283.1 44500 N/A 20.27 3.198 2170 18824 0.423 60.20 !26.3 142.88 2000 274.1 Ht I/4t dT(flu.)3/4t XdT 3.198 2070 17957 0.404 57.42 120.8 138 ~ 92 !900 270.2 N/A 103 0.93 3.182 1970 17089 0.384 54,38 114 ' 134.31 1800 265. 5 Nt 3/4t dT(inst.err) Kit 3.182 1870 !6222 0.365 51.62 109.2 129.84 !700 261. I 0,273 8' 5.96 3.182 1770 153S4 0.345 48,86 103 ~ 7 125.05 1600 256. 3 NEAR RADIUS dP(inst.err) 3,182 !670 14487 0.326 46.10 98.2 I!9.92 !500 25! ~ 2
'?7.07 170.0 750 PSIA 3. 182 1570 13619 0.306 43 '4 92.6 114.36 140O 245. 6 120.0 75O PS!A 3 '6S 14io 12318 0.277 38.99 83.9 104.60 !250 ?35. 9 3.155 1170 10149 0 '28 32.13 70.2 85,65 !epo ? 1 b. 9 3 '65 !070 9282 0.209 ?9.38 64 ' lb.33 '?00 207. 6 3.149 970 8414 0. 189 25 '9 aB ~ 9 o4.94 B)0 !96.i 3 ~ 149 820 71!3 P,!6P ?2 4P 50.7 44.66 700 175. 9 3.149 720 6246 0.140 19.67 45.3 26.8i 600 158.!
3.149 620 5378 0.121 !6 '3 9.8 2,70 500 '.34.0 3 '49 608 5274 0.119 '6 61 39.2 -0.86 488 130.4 570 4945 0.1!I 15.57 37.1 -13.51 450 117.8
$. 149 545 4728 0.106 lb.89 >5,7 - 3.cp 4?5 :OB.O 3 ~ 149 520 4'1' !01 14. 20 34.4 "34.73 400 95.5 3.149 500 4337 0.097 13. 66 33,3 -45.46 380 85. 8 3.14'? 480 4164 0.094 13.11 32.2 -58.!8 360 73. I 3.149 470 4077 0,092 !".84 3!.6 -65.54 350 5.7
t 1
Table 20 RCS PRESSURE/TE)}PERATURE CURVES 40 F H/U AT 3/4 T LOCATION (32 EFPY). ALFA RTndt !/4t RATE )}n Pact Sa Ss/Sv I'1 r
~8 T-RTndt Pind Tind 0 '99 H/A 40 3.251 3295 285&3 0.642 92 ~ 93 }97.8 180.17 31?5 33}. 7 THICKNESS RTndt 3/4t dT(wal}) 3.251 317Q 27499 0,6!8 89. 40 1'?0,7 177.27 3000 328.8 1}.190 43.59 3 '}4 2670 23161 0.520 74.45 }60. 8 163.38 2500 3}4.9 Sy dT(flu.)f/4t dTct 3,}98 2420 20993 0 '72 67.13 146 ~ 2 155.40 2250 306.9 44500 )i/A 40.54 3.198 2170 18824 0 '23 60.20 132. 3 146.89 2000 298.4 Nt 1/4t dT(flu.)3/4t ZdT 3,}98 207Q 17957 0.404 57.42 126 ' 143.}b 1900 294.7 H/A 103 0.93 3.182 197Q 17089 0.384 54.3S 120 ' 138.83 }800 290.4 Ht 3/4t dT(inst'rr) Kit 3.182 !S70 16222 0.365 51.62 115. 1 134 ~ 65 1700 286.2 0.273 8,0 11,91 3.182 1770 15354 0.345 4&.&6 109.6 130.20 1600 281.7
)}EA)} RADlUS dP(inst.err) 3.182 1670 14487 0,326 46.10 104 !. ~ 125,44 !500 277.0 97,07 170.0 750 PS!A 3.182 1570 }36}9 0.306 98.6 120 '4 !400 271.9
}20.0 750 PS!A 3.165 1420 12318 0 '77 38.99 89 9 ~ 111,44 1250 263.0 3.165 !17Q 10149 0.228 a2I }a 76,2 94.52 }000 246 F 1 3,165 1070 9282 0.209 29.38 70.7 86.39 900 237.9 3.! 49 970 84}4 -
O.f&9 26 '9 64 ~ 9 76,66 800 ~L&e L 3.149 820 7!13 0.}60 22 '0 56. 7 59 97 700 211 ~ 5 3.149 7?0 6246 0.}40 19.67 51.".'5.8 46.07 600 fc7 3.}49 620 5378 0.121 16.93 28 F 64 500 180.2 3.149 608 5274 Q.i}9 lb. 61 45.} ?6.22 488 f77.8 3.}49 570 4945 0.}!1 }5. 57 43.0 }7,93 450 169 ~ 5 3 '49 520 451! 0.101 }4. 20 40.3 5.26 400 }56.8 3,149 470 4077 0 '92 12.84 37. 6 -10.28 350 !4!.3 3.149 420 3643 0.082 11.47 -oQ 300 121.2 3 '49 370 3210 0.072 !0.}! 32.1 3'58,&5 250 92.7 3,149 340 2949 0.066 9.29 30 ' -84.}l 220 67.4
l t kyle \ S I'0 tf
~ ~c lt i
E
Table 21 RCS PRESSURE/TEMPERATURE CURYES 100 F M/U AT 3/4 T LOCATION (32 EFPY) ALFA RTndt 1/4t RATE Hn Pact Sa SI/Sy Kin Kir I-RTndt Pin d Tind 0.399 H/A 100 3,251 3295 28583 0.642 92. 93 215. 6 187.03 3125 397 4
~
TMICKNE."S RTndt 3/4t dT(~all) 3.251 3170 27499 0.618 89 ~ 40 208.6 184.41 3000 396.7 1.'.190 0 108.97 3.214 2670 23161 0.520 74. 45 178.7 172.01 2500 384,4 Sy dT(;!u,)I/4t dTct 3.198 2420 ?0993 0.472 67.13 164 ' 165 Q? 2250 377.4 44500 Hi A 101.34 3.198 2170 18824 0.423 60. 20 150.2 157.68 2000 370.0 Ht I/4t dT(flu,)3/4t XdT 3 '98 2070 17957 0.404 57 42
~ 144 ~ 6 154,50 1900 366.8 H/A 103 0 '3 3.182 197Q 17089 0.384 54 ~ 38 138,5 150 '4 !800 363. 2 Ht 3/4t dT(inst.err) Kit 3.182 1870 16222 0.365 51.62 133.0 147.35 1700 359.7
- 0. 273 8.0 29.78 3.182 l?70 15354 0 '45 48 '6 127,5 143.67 !600 356.0 HEAH RADIUS dP(inst.err) 3.182 !670 14487 0,326 46.10 122 ' !39.78 1500 352.1 97.07 170.0 '750 PSIA 3.182 1570 13619 0.306 43.34 116.5 135.66 1400 348.0 120,0 750 PSIA 3.165 1420 12318 , 0,277 38 '9 107.8 128.63 !250 341.0 3.165 1170 !0149 0.228 3?.13 94.0 115.82 !QQQ 3?8.2 3.165 !070 9282 0.209 .29,38 88.5 109.94 900 322.3 3.149 970 8414 O !89 S 26.49 82.8 103.17 800 315.5 3.149 820 7!13 0.160 22.40 74 ~ 6 9? ?6 700 304.6 3.149 720 6?46 0.140 19 67
~ 59. I S3 ~ B9 600 296.2 3.149 620 5378 0.121 !6.93 63.6 74 '6 500 286,7 3.149 6QB 52?4 0.119 !5.61 63.0 73.12 488 ?85. 5 3.149 520 4511 O. 101 14,20 58,2 63.30 40Q 2?5.6
- "149 420 3643 Q,082 !1.47 '2,7 50.12 300 262.5 3,149 370 3210 0 07? IO,!j 50.0 42.45 250 254.8 3.!49 3?0 2776 0.062 8.74 47.3 33.81 20Q 246.2 3.!49 270 2342 0 '53 7.37 44,5 23.94 150 236.3 3.149 220 !908 0.043 6.0! 41. 8 ? 4? 100 ?24.8
S
'r t, It 4
(
'I 4
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Table 22 RCS PRESSURE/TFNPERATURE CURViS IHSFRVICE I>> HYGROTEST (8 EFPY) ALFA RTndt I/4t RATi rn Pact Sn Sn/Sy Kim Klf T-RTndt Pind Tind H/A 0 0 3 ~ 251 3295 28583 0.642 92.93 139.4 151 ~ 37 3125 261.4 THICKNESS RTndt 3/4t dT(wajj) 3.251 3170 27499 0.6!8 89.40 I34.1 148 05F 3000 258.1
!1. 190 H/A H/A 3.214 2670 23161 0.520 74 '5 111.7 131. 39 2500 241. 9 Sy dj(flu.)I/4t dTct 3.198 2420 20993 0 '72 67.13 100.7 122 '3 2250 232.3 44500 102 H/A 3.198 2170 18824 0 '23 50.20 90.3 111 ~ 87 2000 221.9 Nt I/4t dT(f!u,)3/4t kdT 3.198 2070 17957 0.404 57.42 86.1 107.20 1900 217.2 H/A H/A H/A 3. 182 1970 17089 0 '84 54.38 Bl.b !01 '8 1800 211.7 Nt 3/4t dT(inst.err) Kit 3.!82 1870 16222 0.355 51.62 77,4 96 '6 1700 206,3 H/A 8.0 H/A 3,182 1770 15354 0.345 48 '6 73.3 90.38 1600 200.4 NEAR RADIUS dP(inst.err) 3,182 1670 14487 0.326 46.10 59.1 83.94 1500 193.9 97.07 170.0 750 PSIA 3.182 1570 13619 0.306 43.34 55.0 76 85 F 1400 !86.9 120.0 750 PSIA 3,165 1420 12318 0.277 a8.99 58.5 63.95 1250 174.0 3.165 1170 10149 0.228 32 '3 48.2 36.87 1000 146 ~ 9 3.165 1070 9282 0.209 29.38 44.1 22.13 900 132.I
- 3. 149 970 8414 0 ~ 189 26.49 39.7 2.25 800 112.3 3.149 820 7113 0,160 22.40 33m 6 -42 08F 700 6..9 3.149 720 6246 0. 140 19.67 29.5 -105. 48 500 4. 5 F 149 620 5378 0.121 lb.93 25 ' ERR 500 iRR 3.149 o"0 4511 0 IOI
~ 14 ~ 20 21.3 ERR 400 ERR 3,149 420 364. 0.082 1!.47 17 ~ L ERR 300 ERR 3,149 370 3210 0.072 10 '1 ~ U~ 2 ERR 250 EF.R 3.149 320 2775 0.052 8.74 13. I iRR 200 ERR 3 '>>9 270 2342 0 '53 7.37 ERR 15C'RR 3.149 220 1903 0 0~3 5 9.0 iRR 100 ERR 3.149 170 1475 4.54 )'.033 7s0 LRR 50 iRR
I If I
. Table 2 3 RCS PRESSURE/TEHPERATURE CURVES IHSERVICE 0 HYDROTEST (32 EFPY)
ALFA RTndt I/4t RATE Hn Pact Sn Sn/Sv K]n Kir T-RTndt Pind Tind N/A 0 0 3 ~ 251 3295 28583 0,642 92.93 139.4 151.37 3125 275. 4 THICKNESS RTndt 3/4t dT(waII) 3.251 3170 27499 0.618 89.40 134.1 148,05 3000 272.1 11.190 H/A HIA 3 '14 2670 23161 DE 520 74.45 111.7 131 '9 2500 255.9 Sy dT(flu.) I/4t dTct 3,198 2420 20993 0,472 67 ~ 13 100.7 122 '3 2250 246 ' 44500 116 H/A 3.198 2170 18824 0.423 60.20 90.3 II l. 87 2000 235.9 Ht 1/4t dT(flu.)3/4t %dT 3.198 2070 17957 0 ~ 404 57.42 86 F 1 107.20 1900 231.2 H/A N/A H/A 3.182 1970 17089 0.384 54.38 81.6 101.68 1800 225.7 Ht 3/4t dT(inst.err) Kit 3.182 1870 16222 0.36S 51.62 77 4F 96,26 1700 220.3 H/A 8.0 N/A 3. 182 1770 15354 0.345 48.86 73.3 90.38 1600 214.4 HFAN RADIUS dP(inst. err) 3 AD !82 1670 14487 0 '26 46.10 69,1 83 '4 1500 207.9 97.07 170. 0 750 PSIA 3.182 1570 13619 0.306 43.34 65.0 76.85 1400 200.9 120 ~ 0 75D PSIA 3.165 1420 12318 0,277 38.99 58,5 63,95 1250 188.0 3.165 1170 10149 0.228 32.13 48 ' 36 87 F 1000 160.9 3,165 1070 9282 0,209 29.38 44.1 22.13 900 146.1 3 '49 970 8414 0 '89 20 '9 39.7 2.25 800 126a3 3,149 8zo 7113 0.160 22.40 33 ' -42 08 F 700 81 ' 3.149 720 0246 0 ~ 140 !9.67 29.5 -105.48 18.5 3.149 620 5378 0.12l i6.93 25.4 ERR SOD ERR 3.149 520 <SII 0,101 !4.20 21. 3 cnn 400 ERR 3,!49 420 3643 0.082 11 ~ 47 17.2 ERR 300 tRR 3.1<9 370 3210 0 '72 !0.11 15.2 ERR '5) thR 3,(49 320 2776 0.062 8.74 13 ERR 200 ERR 0 ~ 149 270 2342 0.053 7.'7 ERR 150 FPR 3.149 220 1908 0.043 6.01 9.0 ERR 100 ERR 3.149 170 1475 0 0 3 4,6e 7~0 ERR 50 ERR
S it t
F e 1 ANPP COOLD N P T LIMITS ('IO YEARS) 3500 3000 IJJ 2500 2000 M 1500 I 1000 O O K LEGEND 500 10'F/HR 100'F/HR 0 100 200 300 400 INDICATED TEMPERATURE Tc (')
I' F 2 ANPP HEA P P T LIMITS (10 YEARS) 3500 r I I 3000 l N r II I 0 r Ir
/
IJJ I I I 2500 / N r
/
N r I I
/
LIJ
/
0'L I I 2000 / CL / IJJ / N //
//
CL N / N 1500 r /C
/'EGEND IJJ
/
//
,/
CL'L
/
C) bJ I 1000 r r
/
/
ISO O 10'/HR O Z -----'0'F/HR 500 40'/HR-100'F/HR 0 100 200 300 400 INDICATED TEMPERATURE Tc (o F)
j(f S H
Figur ANPP COOLDOII P T LIMITS (40 YEARS) 3500 3000 2500 2000 (A V) 1500 bJ CL Cl LJ 1000 O LEGEND 500 10'F/HR 100'F/HR 0 100 200 300 400 INDICATED TEMPERATURE Tc (')
if f I
ANPP HEA P P T LIMITS (40 YEARS) 3500
/I I
3000 / I l I I I l 2500 /
/ I,
/ I
/
/
/
/ I 2000 !I ,/
bJ I I N j K //
/
V) I V) 1500 /
/
LLI CI / LEGEND CL CI IJJ I 1000 /
/
/
/
ISO r r / 10'/HR O C) z rrr 20'F/HR 500 --- 40'F/HR 100'F/HR 0 100 200 300 400 INDICATED TEMPERATURE Tc (')
I J j ( t
ANPP INSERV E HYDRO LIMITS 3500 (10 AND 40 YEARS) r 3000 r Q r LLI 2500 r M r U) LLI I K r r 0 2000 r r bJ r N r K r V) V) bJ 1500 r EL 0 C) LIJ I 1000 LEGEND r (3 O 10 YEARS Z 500 40 YEARS 0 100 200 300 400 INDICATED TEMPERATURE Tc (0 F)
't I
C E
+
ee /II l I II C II Va t Q i, 4 y I
Figure 6 FIGURE 3.4-2a REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITATIONS FOR 0 TO 10 YEARS OF FULL POWER OPERATION 3000 2800 2600 '. 2400 0 -Q0. 2200 . 0: = 0) 2000 . 0 Q.. LLJ K Q' D Z 1800 Ill IX LL gN 1600 Ql I. jDI- I IX "Q'. IR-
~
-LLI
- LLI N 1400-'LJ LOWEST
' K zK o~ LLJ
~OK IX Q O. 'ERVICE 0 Q wQ 0LLI 1200 TEMP.t - 148 oF
~
1 O I-O 1000 R 800 SOT HERMA 600
~ ff-'
;. 10oF/HR.
4oo COOLDOWN.
'2'0 F /Htt:.
-'+= HEATUP COR RITICAL~~
200 ;,'INIMUMBOLTUP ",
'"', TEMPERATURE 68oF
.i ~ tC 0
0 100 200 300 400 500 INDICATED RCS TEMP'C ( F)
p Q4' II 0<~' I II p I
Figure 7 FIGURE 3.4-2b REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITATIONS FOR 10 TO 40 YEARS OF FULL POWER OPERATION 3000 2800 2600 2400 iQ 2200 Q
-Q.
.Q'Q 2000 1800 K
N 1600 1400
~ LOWEST l I
~O 0
Pg l~, i 1200 f SERVICE . TEMP. 1484F o W x O L o Q 1000 O 800
+IS OT HERMAL 600 400 10OF/HR COOLDOWN 2P'F///R CORE CRITIC I ~
AL 200 HEATUP , MINIMUNBOLTUP: TEMPERATURE 68 F 0 100 ~ 200 300 400 500 INDICATED RCS TEMP. TC ( F)
t i
06/26/89 PCP Number Description I Design Change 50 '9 Report
- I 85-02-CD-017 This change replaced the alarming seal water floaters on This change did not introduce an unreviewed safety question.
85-03-CD-017 the condensate pumps with new non-alarming rotameters. This The replacement rotameter is similar in configuration to the change was implemented in Units 2 and 3 during this existing meter and is not a safety related instrunent. The reporting period and affected FSAR.Section 10.4. elimination of the alarm function also reduces unnecessary alarms in the control room. 86-01-CH-003 This design change revises Plant Change Package 86-01-CH-003 This revision did not introduce an unreviewed safety 86-02-CH-003 to add a time delay relay to cabinet E-ZAN-C02 to prevent question. The revision ensures that CHE-UV-231P will fail the momentary closing of contacts TS-231HH and LL and T231 open on a loss of power to the instrwent cabinet as noted (due to a relay race condition) from closing valve CH-231 in CESSAR Table 9.3-7 and FSAR Figure 9.3.13. Thus this PCP (seal injection isolation valve). This modification is is not a design change as it corrects the wiring to allow necessary to ensure valve CHE-W-231 operates as originally the valve to function as originally designed. intended on a loss of power condition within the loop electronics. 86-02-CH-030 This change added narrow range pressure instrunentation for This change did not introduce an unreviewed safety question. the reactor drain tank nitrogen blanket monitoring in the No safety related equipment is involved and the added control room. This affected FSAR Sections 9.3.4. instrunentation will enable more accurate reactor drain tank pressure monitoring. 85-02-CH-039 This change replaced indication scales and recalibrated This modification did not introduce an unreviewed safety process cards due to the installation of new Foxboro cards. question. The change only affected the scales and did not change the function of the instrunentation. 85-01-CH-112 This change replaced the nitrogen regulator to the reactor This modification did not introduce an unreviewed safety drain tank with an identical regulator of a lower operating question. The replacement of a pressure regulator with an range. The previous pressure regulator had a spring range of identical regulator does not change any previous analysis. 2-6 psig and was set at 3 psig. This allowed a continuous The new regulator will perform the same function as the flow of nitrogen and unnecessary gaseous radwaste to the previous but with reduced gaseous radwaste. radwaste surge tank. The new regulator has a lower range and will be set at .5 psig. 85-03-CH-170 This change was previously reported as item 4 in the 1987
Page No. 06/26/89 2 0 Design Change 50.59 Report PCP Number Description gunnery 50.59 Report. 84-01-CN-004 This change replaces the level switches for the chemical This modification did not introduce an unreviewed safety
~aste tank. The setpoints are too low to set on the present question. The chemical waste tank performs no safety range of the instrunents. The set point on the old switches function and affects no equipment which performs a safety falls within the first 10 to 15K of its range making it function. The replacement of the level switches does not inaccurate and preventing the system from operating as it change the system operation but enhances the system was designed. performance.
85-01-CP-005 This change was reported as item 142 in the 1987 50.59. Report. 85-02-CM-023 This change involved Pitot tube taps for use during testing This change did not introduce an unreviewed safety question. of the cooling towers. The taps were in place but required No important to safety equipment is involved in the change modification for installation of the pitot tubes. An and the change allows for testing of the cooling towers to isolation valve and blind flange were added to the taps to trend their performance and allow predictive maintenance. aid in isolation of the taps for cooling tower performance testing. Excavation was also necessary to uncover the taps. The change was implemented in Unit 2 during this reporting . period and affected FSAR Section 10.4. 85-01-DG-035 This change replaced a diesel generator manifold temperature This modification did not introduce an unreviewed safety control valve with a different model valve. question. The repalcement valve meets the same specifications as the original and performs an identical function. No system operation is affected. 85-01-DS-004 This change added a pipe support to the emergency shower This modification did not introduce an unreviewed safety stand at the spray pond metering pump house. question. No equipment important to safety or safety related was affected by this modification. The emergency shower is for personnel protection only. 85-01-DS-019 This change installed a backflow preventer in line AS-313 of This change did not introduce an unreviewed safety question. the domestic water system at the auxiliary boiler. The The change does not involve any equipment that is important change also installed vaja breakers on 1 inch lines AS-106, to safety.
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06/26/89 0 Design Change 50.59 Report PCP Number Description AS-107, AS108, and AS-522. The changes were iapiemented in Unit 1 during this reporting period and affected FSAR Section 9.2. 85-01-DN-020 This change reduced the pressure to the seals on the control This change did not introduce an unreviewed safety question. building oily waste sump purps. The previous supply pressure The reduced seal pressure does not constitute a change to was higher than the vendor recamnedation and so this change the facility as described in the FSAR. No system operation was made to comp;y with the vendors recomendation. or design criteria are affected by this change. No safety related or iaportant to safety equipment or systems are affected by this change. 85-01-ED-012 This change rewired the heater drain pwp low flow trip to This modification did not introduce an unreviewed safety provide a time delay prior to the trip signal being question. There are no safety related or iayortant to safety generated. This will prevent spurious trips caused by equipment affected by this change. The change will prevent momentary low flow conditions. spurious trips of the heater drain pumps and consequently reduce the probability of a loss of feedwater occurrence. 85-01-ES-004 This change removes the existing "Loss of Power" nuisance This aodification did not introduce an unreviewed safety 85-02-ES-004 alarm from the Safety Equipment Status System (SESS) for the question. The change to the alarm circuitry did not affect 85-03-ES-004 contaireent refueling purge valves CPA-UV-2A, 28; CPB-UV-3A, the operation of the equipment. The SESS is for operator
- 38. During normal operation the valve circuit breakers are convienence only and performs no safety function. The change racked out per procedure 40AC-OZZ06. Since these valves are will eliminate nuisance alarms from the control room and ESF actuated and monitored by SESS for loss of power a improve safety from a hwen factors standpoint.
continuous nuisance alarm in the control room. The change will enable the SESS to alarm only when the valve is not in a safe position and loss of power occurs. In addition the loss of power alarm will be deleted from SESS for the valves SIC-UV-653, SID-UV-654, SPB-HV-SOA, and SPB-HV-508. These valves are not automatically actuated ESF devices. During normal operation the circuit breakers for these valves are racked out per procedure 40AC-OZZ06. Ths causes a continuous nuisance alarm in the control room. 85-01-FH-008 This change replaced the cadmiun plated bolts on the spent This modification did not introduce an unreviewed safety
( 06/26/89 0 Design Change 50.59 Report PCP Number Description fuel handling machine tool storage bracket with stainless question. The stainless steel bolts are stronger than the steeL bolts. cadmiun plated bolts and have better corrosion resistance. 85-02-FH-041 This change installed bellville washers to remove the gap on This modification did not introduce an unreviewed safety the end of the load brake pinion shaft to assure maxisxIa question. The installation of the bellvi lie washers corrects clutch engagement and proper operation of the spent fuel an existing problem and does not affect the operational or handling machine and refueling machine gear boxes. Reduction failure modes of the spent fuel handling machine or of the gap on the end of the load brake pinion shaft will refueling machine. FSAR Section 9.1.4 was reviewed and correct a fuel hoist load brake overspeed problem. installation of the washers does not change the safety evaluation. 85-02-FH.043 This change added spacers to limit the travel of the This modification did not introduce an unreviewed safety refueling machine hoist spreader. This eliminated an question. All the existing requirements applicable to the interference between the spreader and core support barrel design and operation of the refueling machine remain intact which occured when extracting fuel from the outer edge of and unaltered by the change involving the spacers. The the core. spacers allow the refueling machine to function as designed. No changes were made to the operational or failure modes of the refueling machine. 84-01-FT-003 This change provided a separate power circuit to the This modification did not introduce an unreviewed safety mainfeedwater pump turbine turning gear. The original question. The main feedwater Fxmp turbine turning gear is circuit routing was not adequate to supply sufficient power for equipment protection only and serves no safety function to the solenoids. and affects no equipment important to safety or safety related. This change will allow it to perform its intended function. 85-01-FM-018 This change removed the low flow cutoff feature of the This modification did not introduce an unreviewed safety Foxboro square root extractor cards in the main feedwater question. The modification involves balance of plant flow system to provide automatic operation and flow equipment and affects no safety related or important to. indication at low flow. safety equipment. The change does not affect the operation of the main feedwater system other than to allow automatic operation at lower flow rates. 81'-01-GA-002 This change modified socket welds (6 locations) on the This change did not introduce an unreviewed safety question.
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06/26/89 0 Design Change 50.59 Report PCP Number Description Sunna ry 87-02-GA-002 nitrogen supply line to the Steam Generators and Reactor The modified socket welds meet the requirements of FSAR Drain Tank. The subject line GANL008 (S30 class) is Section 3.7.2.8. Heeting these requirements brings the plant connected to valve GAA-UV002 (GIB class) which is identified back into conformance with the FSAR and design bases. as a contairment isolation valve per Table 6.2.4-1 of FSAR. The subject valve is Seismic Category I and the line is Seismic Category III. Section 3.7.2.8 of FSAR discusses the techniques used to consider the interaction of Hon-Category I structures with Seismic Category I structures. The result of this analysis requires modification of the socket welds. 85-01-GH-006 This modification provided a non-linear scale for the main This modification did not introduce an unreviewed safety generator hydrogen purity meter in the control room. It question. The hydrogen purity meter is for equipnent also provided for a curve fit to be performed on the Plant protection of the main generator and does not perform an Coeputer point for generator hydrogen purity. The purpose ot important to safety or safety related fumtion. The change the change was to match the non-linear output of the General corrects the control room indication to provide a more Electric hydrogen purity monitor with the control room accurate indication of hydrogen purity. indications. 86-01-HC.008 This change installed backdraft dampers in the discharge This change did not introduce an unreviewed safety question. 86-02-HC-008 stacks of the control element drive mechanism (CEDH) The dampers are not safety related and are qualified to cooling fans 2MHCNA02A, 8, C, and D. The change was withstand a safe shutdown earthquake. An analysis has been implemented in Unit 2 during this reporting period and performed to ensure the structural integrity of the entire effected FSAR Section 9.4. CEDH cooling unit. 85-01-HC-013 This modification added flanges to the Control Element Drive This modification did not introduce an unreviewed safety Hotor (CEDH) cooling system collecting ring duct. the question. This change did not affect system operation. The purpose of the change was to facilitate duct removal and change was within the design criteria of the system and did re-installation during refueling. not affect its siesmic qualification. 85-01-HC-015 This change added five contaiment air temperature inputs This change did not introduce an unreviewed safety question. from thermocouples JHCNTE42A, B, C, D, and E to the plant The plant computer does not perform any safety functions and computer. Signal conditioning cards were also installed in isolation devices are included in all safety related inputs balance of plant analog instrunent cabinet (JZJNC02C) and to the computer. an input card was replaced in mltipoint recorder
>J 06/26/89 0 Design Change 50.59 Report PCP Nurrber Description Sunna ry JRHHUJR0005. These changes were implemented in Unit 1 during this reporting period and affected FSAR Secton 9.4 and drawing 13-E-ZJC-039, which is incorporated into the FSAR by reference. 86-01-HJ-014 This change installed an electrically operated exhaust fan This change did not introduce an unreviewed safety question. with automatic back draft shutter, wall mounted thermostat, The changes are not related to safety related equipment, but a fixed lower with bug screen and dust filter in the help reduce the temperature in the elevator machinery room. corridor building elevator machinery room. The change was implemented in Unit 1 during this reporting period and affected FSAR Section 9.4 and drawings 13-J-HJL-004, 13-J-HJE-051, 13-E-HJB-002, 13-E-ZJC-037, 13-E-ZJC-038, and 13-J-ECL-002 which are incorporated into the FSAR by reference. 87-01-HP-003 This change modified a pipe support on the inlet line to This modification did not introduce an unreviewed safety 87-02-HP-003 contairment Post-LOCA hydrogen monitor B. This modification question. The hydrogen monitoring system is discussed in is required to bring the pipe support in compliance with the FSAR Section 6.2.5.2.2.2. This change brings the piping for ASNE Code Section III requirements. this system in accordance with ASHE Section III and thus establishes and improves the operability of the system. 84-01-HP-007 This modification changed the clamping and support for the This modification did not introduce an unreviewed safety calibration and reagent gas bottle of the Post LOCA hydrogen question. No change in system operation was performed. The analyzers. The purpose of the modification was to ensure an changes do not affect the system as described in FSAR acceptable seismic mounting was provided for the bottles and 6.2.5.2.1 ~ The change ensures the hydrogen reagent bottles pressure regulators. will be adequately supported for a seismic event. 85-01-HP-019 This modification installed two valves on the Post LOCA This modification did not introduce an unrevicwed safety Hydrogen analyzer inlet lines to accomodate a sampling bomb. question. The addition of these valves'does not affect The purpose of this change was to provide an alternate system operation. The valve installation is )n accordance sample flowpath for the contaiwent atmosphere. with all plant design criteria and does not affect any equipment important to safety. The valves allow sampling of the contaiment atmosphere in the event the hydrogen analyzer is out of service.
06/26/89 Design Change 50.59 Report PCP Hmher Description gunnery 85-02-HS-004 This modification provided a supplemental air conditioning This modification did not introduce an unreviewed safety system to reduce the temperature of the microwave room question. The addition of this non-quality related air building) to 75 degrees fahrenheit. The purpose of conditioning unit does not affect any safety related (160'ontrol this change is to improve electrical equipment life in the equipment. The failure of the system would have no effect on room. safe shutdown of the plant. 84-01-LO-002 This change added circuitry to the Hain Turbine Emergency This change did not introduce an unreviewed safety question. 84-02-LO-002 Bearing Oil Punp starting circuit to: 1. Start the EBOP 'on The EBOP is for equipment protection only and its operation loss of power to the Turning Gear Oil Pump; 2. Added does not affect any safety related or important to safety redundant pressure a~itches to the EBOP starting circuit. equipment. This change isproves the reliability of the EBOP. The purpose of this change was to improve the reliability of the EBOP. 83-01-LR-003 This change added a stainless steel bypass line around the This change did not introduce an unreviewed safety question. 83-02-LR-003 Liquid Radwaste System Evaporator. The purpose of the line All the original requirements applicable to the design and is to allow the use of alternate means of processing liquid operation of the LRS remain unaltered by the addition of a radwaste when the evaporator is out of service. permanent line bypassing the LRS evaporator to an alternate method of radwaste processing. This change does not affect any equipment required for safe shutdown or to mitigate the consequences of an accident. 85-02-LR-004 This change replaced the meter face on chemical drain tank This change did not introduce an unreviewed safety question.'he level indicator to read in percent versus inches of water. FSAR does not cover this level of details Ho equipment function or operation is affected. 85-02-LR-022 This change rewires the foxoboro module for the chemical This change did not introduce an unreviewed safety question. drain pump high discharge pressure trip to trip the pump if Ths feature is not described in the FSAR and is only the pressure transmitter signal is lost. This provides a utilized for protection of non-quality related equipment fai l safe mode of operation. (chemical drain pump) ~ Ho safety related or important to safety components are affected by this change. 85-03-LR-032 This change added stiffeners to the top steam tray of the This change did not introduce an unreviewed safety question. Liquid Radwaste System (LRS) evaporator and changed the The stiffeners and change to the recycle line reduce the
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06/26/89 Design Change 50.59 Report PCP Number Description Swmary recycle line input from the cold side of the distillate stresses in the evaporator body thus improving its cooler to the hot side. The purpose of this change was to structural integrity. The evaporator function is not changed prevent the recurrence of fatigue cracking in the evaporator and all project design criteria are met. body. 85-AO-N-009 This change constructed a second evaporation pond east of This change did not introduce an unreviewed safety question. the existing pond, with all reLated piping and structures. The construction is a planned addition in the original The change was implemented in all three units during this design. The FSAR accident analysis are not affected. reporting period and affected FSAR Sections 2.4 and 9.3. 86-01-PC-007 This change added a check valve and bleed valve to the air This change did not introduce an unreviewed safety question. 86-02-PC-007 supply for the spent fuel pool and cask loading pit gate This change improves the reliability of the spent fuel pool 86.03-PC-007 seals. The purpose of this change was to prevent deflation seals. The spent fuel pool is a backup source of borated of the seals on a loss of air supply. water for the reactor coolant system (RCS) and this change reduces the possibility of a loss of this source of borated water due to leakage from the gate seals. Thus improving the margin of safety. 86-03-QA-007 This change provided a 240VAC/30A outlet in the hot This modification did not introduce an unreviewed safety chemistry lab. question. The receptacle is powered from non-class power and affects no safety related or important to safety equipment. 86-03-QC-004 This change installed additional area lighting in the This modification did not introduce an unreviewed safety protected area. The additional lighting.was required to meet question. The additional lighting is powered from non-class FSAR Section 9.5.3.1,2 and 13.6.2. This change maintained busses and has been evaluated to ensure it does not the PVHGS conmitment to provide .2 foot candles minimus for potentially affect any safety related or important to safety area lighting. equipment. 84-01-QF-017 This change installed the Emergency Notification System and This modification did not introduce an unreviewed safety Health Physics Network dial up phones as required per NRC question. The plant connunication system is not required for emergency preparedness guidelines. safe shutdown nor is it required to mitigate. the consequences of an accident. This modification did not change the function or purpose of the plant coaaunication sys'em
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PageHo. 9 06/26/89 Design Change 50.59 Report PCP Number Description Sunna ry 86-01-QF-021 This change installed soundproof booths for the existing This change did not introduce an unreviewed safety question. 86-02-QF-021 phones located in high noise level areas in contairaent. The installation of the booths was evaluated by Engineering for fire protection, potential for clogging contaiment recirculation sumps, and introduction of aluninun into contairment. All aluninun was removed from the booths. The amount of material postulated to be removed from one of the booths in the event of a high energy line break was determined to be neglegible and have no afffect on the contairment sutps operation. No fire protection or combustible materials loadjng problems were identified. The booths have no affect on safety related or important to safety equipnent. 84-03-QF-026 This design change installed an additional Security base This change did not introduce an unreviewed safety question. radio station radio to provide an backup security channels The addition'of a Security base radio station enhances security and does not affect plant safety analysis. 85-01-QF-043 This change deleted the <<REQ FAL<<comnand from the Plant This change did not introduce an unreviewed safety question. Hultiplex System software. This comnand changed the This is not a change to the facility as described in the controlling processor from the online processor to the FSAR.,The deletion of the comnand will result in better backup processor. However, if the online processor was still performance of the Plant Hultiplex System. No safety related operative when the comnand was given the two processors systems are affected by this change. would share the controlling function and the system would become unstable. There is an operator switch which performs. the same function without the resultant unstable operation. 85-01-QF-055 This change: 1. Installed radio base stations, battery This change did not introduce an unreviewed safety question. racks, and antenna to~er for Security primary and backup This change does not affect the facility design as described radio channels at the corridor building Unit 3 in the FSAR. This change does not affect any systems which
- 2. Installed new 7/8<< coax antenna cable for 160'levation.
are important to safety or safety related. operations, and Security primary and backup radio systems in the auxi lary and radwaste buildings. 3. Installed new antenna and 7/8<< coax antenna cable at the contaiwent
I 06/26/89 0 Design Change 50.59 Report Stannary PCP Humber Description building 74'levation. The purpose of the change was to; enhance Security and Operations radio commnications within the protected area. 85-01-RC-006 This change replaced the overload heaters on valves This change did not introduce an unreviewed safety question. J-RCE-HV-430, 43'1, 432, to the correct full load amp (FLA) These valves are not required for safe shutdown. The change rating of 0.7 amps and changed the valve nameplates to provides the correct overload heater values for the reflect the correct FLA settings. installed valve motor operators. 87-01-RC-008 This change modified pipe supports on lines RCE-L009 and This change did not introduce an unreviewed safety question. 87-02-RC-008 RCE-L144. This was required to bring the pipe supports in The pipe supports do not affect system operation or compliance with the ASHE Code Section Ii. function. The change brings the pipe supports in coapliance with the ASHE Code. 87-01-RC-011 This change added standpipes to the pressurizer level taps This change did not introduce an unreviewed safety question. 87-02-RC-011 on the bottom head of the pressurizer. The purpose of the The addition of standpipes does not affect the reactor change was to prevent sludge from clogging the sensing coolant system (RCS) pressure boundary. The standpipes will lines. improve the reliability of the pressurizer level indicators by preventing sludge from clogging the sensing lines rendering the indicators inoperable. 87-01-RC-041 This change modified the reactor coolant pump (RCP) shafts This change did not introduce an unreviewed safety question. to reduce the susceptibility to fatigue cracking. The The modified shafts perform the same function and their modifications consisted of removing the chrome from the operation is not affected. Reducing the shafts keyway areas of the shaft, extending the'shaft stop seal to susceptibility to fatigue cracking reduces the probability act as a thermal barrier to the keyway, and beveling all of a RCP sheared shaft. This improves the safety margin for. step changes in the shaft. this analysis in FSAR Section 15.3.4.. 85-01-RC-047 This item was previously reported as item 146 of the 1987 50.59 report. 86-01-RD-005 This change removed the sump level Lo-Lo nuisance alarms in This change did not introduce an unreviewed ssafety 86-02-RD-005 the control room for LSLL-200 (fuel building), LSLL-107 question. The Lo-Lo alarms are not important to safety and (ESF), LSLL-108 (ESF), and LSLL-013 (reactor cavity). The do not affect equipment important to safety. The sumps are
C 06/26/89 Design Change 50.59 Report Summary PCP Number Description power supply also changed to the contaiment st level transmitters was to Class 1E to provide level monitoring in the normally dry and the control room alarms were nuisance alarms. The powering of the contairment sunp level event of loss of offsite power. The change was implemented transmitters from a Class 1E source allows determination of in Unit 1 during this reporting period and affected FSAR RCS leakage in the event of loss of offsite power. Section 8.3.1.4.1.1-6 and drawings 13-E-ZJC-037, 13-E-PHA-001, and 13-E-NHA-019, which are incorporated into the FSAR by reference. 87-03-RH-003 This change added a ground cable to the plant computer typer This change did not introduce an unreviewed safety question. per vendor recomnendations. This addition of the ground cable meets the original specifications. The Plant Computer is designed for operator convienence and performs no safety function. 85-03-RZ-007 This change added plant vent, fuel building exhaust vent, This change did not introduce an unreviewed safety question. and condenser air removal vent flow signals to CRACS and The installation of flow sensors with interfaces to CRACS ERFDADS for Unit 3. This affected FSAR section '11.5 and and ERFDADS satisfies recpirements of ANSl H13.1-1969 and drawings 13-E-ZJC-014 and 13-E-ZJC-039, which are Regulatory Guides 1.21 and 1.97. There are no changes to the incorporated into the FSAR by reference. safety design basis. 85-01-RZ-022 This change modified the Post Accident Sampling System This change did not introduce an unreviewed safety question. (PASS) to provide a permanent PASS gas sampling system by The PASS is not a safety system used in mitigating the installing sample conditioning panel 01-J-SSH-A01C and consequences of an accident. The purpose of PASS is to 01-J-SSN-A01D and associated racks. This complies with the provide samples for Post Accident analysis. requirements of NUREG 0737, 11.8.3. 85-02-RZ-024 This change was previously reported as item 54 of the 1985 85-03-RZ-024 50.59 report. 85-02-SB-036 This change installed supports for multiconductor cables This 'change did not introduce an unreviewed safety question. inside the Plant Protection System Cabinets (1-J-SBA, SBB, The addition of the supports does not affect the function or SBC, SBD). This change prevents the cables from moving due operation of the system. The supports ensure the connectors to their own weight and loosening the connector back piece. remain tight thus improving system reliabi lity, 85-01-SB-052 This change replaced variable setpoint cards J642, J635, and This change did not introduce an unreviewed safety question.
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06/26/89 Design Change 50.59 Report PCP Rwher Description Burma ry 85-02-SB-052 J638 in cabintes J-SBA-C01, J-SBB-C01, J-SBC-C01, and The new cards perform the same fmction the same as the J-SBD-CO'I. The purose of the change was to reduce the signal original cards. The new cards will prevent spurious reactor noise present in the cards and prevent spurious reactor trips which may challenge safety systems as a transient trips. The new cards were supplied by Combustion Engineering initiator. This change does not involve a change to the as an enhancement to the present variable setpoint cards. facility as described in the FSAR and does not affect the basis of Technical Specification 3/4.3.1. 8T-01-SC-002 This change modified the hanger design for pipe support This change did not introduce an unreviewed safety question. 87-02-SC-002 hangers on the Steam Generator Sample lines. As a result of The pipe support modifications do not affect system DER 86-32 a reanalysis of the pipe supports was performed operation. The pipe support modifications are required to and the above pipe support changes determined to be meet the design criteria of FSAR Section 3.9.3, Regualtory required. Guide 1.84, and FSAR Table 5.2-2. This change corrects an identified deficiency and returns the plant to the requirements of the FSAR ~ 85-01-SC-028 This change removed the buck/boost transformers in the power This change did not introduce an unreviewed safety, question. supplies to the hotwell sample pumps, changed the po~er The hotwell sample pwps perform no safety function. The source to eliminate long cable runs which were causing change only involves the non-class power system in the excessive voltage drops, and eliminated the capabHity to turbine building and does not affect any safety related or start the Ixnps from the chemistry laboratory. The new power isportant to safety equiFment. The change isyroves the source provides the correct voltage for pump operation. The operation of the pumps by ensuring the proper. voltage is remote start capabHity is not required as the pumps are run supplied. continuously. 85-01-SC-045 This change was previously reported as item 149 of the 198T 85-02-SC-045 50.59 report. 86-01-SF-001 This change revised the steam bypass control system (SBCS) This change did not introduce an unreviewed safety question. quick-open, quick-open permissive and auto modulate sequence The Steam Bypass Control System is not required for safe to provide more even steam loading for condenser shells. shutdown and does not mitigate the consequences of any Previously, all the steam bypass valves in condenser shell accident described in the FSAR. The change improves system "A" modulated full open prior to any of the steam bypass performance and does not affect the system design bases. It valves in either of the other two condenser shells opening. also prevents unnecessary turbine trips. This change does This was causing uneven heating of the main turbine exhaust not affect any equipment important to safety or safety
j Page No. 06/26/89 Design Change 50.59 Report PCP Naker Description Sutmary hoods and high exhaust hood trips of the main turbine. This related. change evened the steam loading between condenser shells and eliminated this problem. 86-0'1-SF-006 This change modified the Feedwater Control System (FNCS) to This change did not introduce an unreviewed safety question. coordinate valve switchover. at 15K po~er and provided This change improves the performance of the FMCS. The FMCS dynamic coapensation of feed pmp turbine speed setpoint is not required to remain functional per CESSAR Chapter 15 demand. The purpose of the change was to alleviate the analysis. The change involves only internal wiring and does potential for plant trips on loop temperature cold mismatch not affect functions of the system as described in the FSAR and prevent erratic control of steam generator water level Section 7.7.1.1.4. due to feed ixmp turbine speed over response to control signal. 87-01-SG-004 This change was previously reported as item 154 of the 1987 87-02-SG-004 50.59 reports 86-0'1-SG-027 This change modified the main feedwater pipe whip restraints This change did not introduce an unreviewed safety question. to provide adequate clearance for thermal expansion growth. The modification of the pipe supports does not affect system operation. Eliminating the interference bet~can the main feedwater piping and the pipe whip restraint reduces the loading on the pipe and isproves its ability to withstand other dynamic loading. Thus the margin of safety of the system is inproved. 86-01-SG.039 This change removed temporary special instrunentation This change did not introduce an unreviewed safety question. installed in the Unit 1 S/G P2. The instrunentation was for Removal of the instrunentation will restore the steam data gathering on steam generator performance. The testing generator to its as designed condition. The instrunentation program was completed and the instrunentation was removed. . removed did not provide the operators with any information and did not perform any control or alarm function. 85-01-SG 044 This change replaced the existing accumlator low pressure This change did not introduce an unreviewed safety question. 85-02-SG-044 switches (JSGAPSL0260, JSGBPSL0261, JSGAPSL0262, The new switches are qualified to the same criteria as the JSGBPSL0263, JSGAPSL268, JSGBPSL0270) on the Hain Steam original switches and perform the identical fwx:tion. only isolation Valves (HSIVs) and the Feedwater isolation Valves the reset point is changed. The change in the alarm reset
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Page No. 06/26/89 Design Change 50 '9 Report PCP Humber Description Su@nary (FWIVs) with new Barksdale pressure switches. This switch point will prevent unnecessary control room alarms. The replacement was required to eliminate nuisance alarms in the alarm reset point does not affect the Technical Safety Equipment Inoperable Status (SEIS). Specification Bases. 85-01-SG-067 This change replaced the existing accumulator low pressure This change did not introduce an unreviewed safety question. 85-02-SG-067 switches (JSGAPSL0264, JSGBPSL0265, JSGAPSL0266, The new switches are qualified to the same criteria as the JSGBPSL0267, JSGAPSL0269, and JSGBPSL0271) on the Hain Steam original switches and perform teh identical function. Only Isolation Valves (HSIVs) and Hain Feedwater Isolation Valves the reset point is changed. The change in the alarm reset (FWIVs) with new Barksdale pressure s~itches. This switch point does not affect the Technical Specification Bases. replacement was required to eliminate nuisance alarms in the Safety Equipment Inoperable Status System (SEIS). 85-01-SG-068 This change replaced the existing accumlator low pressure This change did not introduce an unreviewed safety question switches (JSGAPSL0260, JSGBPSL0261, JSGAPSL0262, The new s~itches are qualified to the same criteria as the JSGBPSL0263, JSGAPSL0264, JSGBPSL0265, JSGAPSL0266, original switches and perform the identical function. Only JSGBPSL0267) on the Hain Steam Isolation Valves (HSIVs) and the reset point is changed. The change in the alarm reset the Feedwater Isolation Valves (FWIVs) with new Barksdal'e point does not affect the Technical Specification Bases. pressure switches. This switch replacement was required to eliminate nuisance alarms in the Safety Equiixnent Inoperable Status System (SEIS). 85-02-SG-095 This change installed wave springs under the piston rings in This change did not introduce an unreviewed safety question. Steam Bypass Control Valves (SBCVs) 1JSGRPV1001, 1002, 1003, the Steam Bypass System is not required for safe shutdown or 1004, 1006, 1007, 1008. A wave spring was already installed to mitigate the consequences of any accident described in in 1JSGNPV1005 and it was not included in this change. The, the FSAR. This change does not affect the system operation. purpose of the change was to iaprove the reliability of the The addition of the wave spring will inprove'valve piston ring seal by preventing foreign debris from entering relaibility and implements a recoaaendation of the valve the area above the piston ring and holding the ring away vendor Control Components Inc.. from its seating surface. 85-02-SI-004 This change relocated Ultrasonic flow transmitt'ers This change did not introduce an unreviewed safety question. 2J-SIN-FI-303 and 304 from the discharge to the suction line The change ensures compliance with FSAR Section 6.3 by of the High Pressure Safety Injection Pumps (HPSI). The achieving reliable Ultrasonic low flow indication. In change was implemented in Unit 2 during this reporting addition, redundant input is available for low flow
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Page No. 15 06/26/89 Design Change 50.59 Report PCP Number Description Su@nary period and affected FSAR Section 6.3 and drawings indi cat ion. 13-J-ZAF-001, 13-J-ZAF-002, 13-E-ZAC-001, and 13-E-ZAC-002/ which are incorporated into the FSAR by reference. 87-01-SI-015 This change was previously reported as item 157 of the 1987 50.59 report. 86-01.SI-039 This change was previously reported as item 150 of the 1987 50.59 report. 86-02-SK-012 Safeguards Information Safeguards Information 86-01-SK-0)3 Safeguards Information Safeguards Information 86-02-SK-013 85-01-SK-026 Safeguards Information Safeguards Information 85-02-SK-036 This change modified security hatches to incorporate hinges, This change did not introduce an unreviewed safety question. handles, and test slot. This change does not impact any systems outside of security. 85-01-S0.001 This change rerouted the Generator Seal Oil Tank vent line This change did not introduce an unreviewed safety question. from the south wall of the turbine building to the east wall The change does not affect any equipment important safety or of the turbine building. The change was necessary to move safety related. The change does not affect system operation. the vent oulet away from any potential ignition sources Moving the vent line away from potential ignition sources since the vent line may contain hydrogen. meets the General Electric requirements for locating the vent and reduces the probability of a fire. 85-02-SP-030 This change replaces the JAECO essential spray pond chemical This change did not introduce an unreviewed safety question. injection pumps with Hilton Roy pumps. The new pcs will FSAR Section 9.2.1 and 9.2.5.2 describes the, Essential Spray perform the same function as the previous pcs and meet the Pond System. This change does not affect that description. same design specifications. The new pcs perform the same fmction and meet the same design requirements. 85-01-Sa-015 This change was previously reported as item 151 of the 1987
i' Page No. 06/26/89 Design Change 50.59 Report PCP Neer Description Suanary 50.59 report. 85-02-SQ-023 This change installed structural tie down hardware to This change did not introduce an unreviewed safety question. 85-03-SQ-023 provide an anchor for the portable airborne radiation This change does not change the facility as described in the monitors in each unit for Seismic 1X considerations. FSAR. Addition of the tiedowns will prevent the portable monitors from damaging any important to safety or safety related equipment during or after a seismic event. The change does not alter the function or operation of any equipment or system. 85-01-SR-004 Ths change made modifications to the Unit 1 solid radwaste This change did not introduce an unreviewed safety question. system as follows: 1. Lower the ~sate feed tank discharge The solid radwaste system is not safety related and no header (SRN-004) below the tank water line. 2. Add a check systems or components protecting the public from accidents valve below the water line. 3. Add a block valve and flush within the solid radwaste system are modified. water line upstream of the new check valve. 4. Extend waste feed tank drain line (SRN-204) into the radwaste building floor drain. This affected FSAR Sections 11.4 and 9.3. 86-AO-SR-014 This change made all the site preparations necessary for This change did not introduce an unreviewed safety question. interface with dry active waste processing'and storage The dose limits specified in the Technical Specifications facility (DAMPS), which was reported as a separate change will not change. The safety evaluation performed for the (ref. 1987 report, item no. 42). The site preparations were DANPS facility encompasses the changes performed thru this implemented in all three Units during this reporting period design change package. and affected FSAR Sections 9.2 and 9.5, and drawings 13-E-ZW-008, which is incorporated into the FSAR by reference. 86-AO-SR-023 This change was reported as item 42 of the 1987 50.59 report. 86-03-SS-003 This change was previously reported as item 153 of the 1987 50.59 reports 85-01-SS.011 This change installed an AC noise filter in the This change did not introduce an unreviewed safety question.
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Page No. 06/26/89 0 Design Change 50.59 Report PCP Number Description Surmary hydrogen/oxygen analyzers. The noise filters are an internal modification to the analyzer circuitry which will enable the system to fmction as designed. Operation of the system as described in FSAR Section 9.3.2 is not affected. 85-01-SS-016 This change provides demineralized flush water capability to This change did not introduce an unreviewed safety question. 85-03-SS-016 the reactor coolant sampling line. The purpose of the line The flush line is installed per all applicable design is to allow flushing of the line to reduce radiation levels criteria. The sampling system is not safety related. The in the hot chemistry lab in the event a post accident sample flushing capability is required to meet Technical is taken from the RCS Hot Leg Loop 1. Specification 3.3.3.1 that requires inititation of a Preplanned Alternate Sampling Program to monitor Post Accident Sampling System parameters. 85-02-SV-002 This change moved the signal cable shield ground for the This change did not introduce an unreviewed safety question. Loose Parts and Vibration Honitoring System (LPVHS) from the This change does not constitute a change to the facility as instruIant ground bus to the chassis ground. This change described in the FSAR. The operation of the system is reterminates the shield ground wire in accordance with the . unaffected by this change. The retermination of the shield vendor requirements. ground brings the installation in accordance with the vendor requirements. 85-01-TC-009 This item was previously reported as item 159 of the 1987 50.59 report. 85-03-ZA.055 This change installed local storage racks for gas cylinders This change did not introduce an unreviewed safety question. at the Post LOCA Analyzer J-HPB-E02, Post LOCA Analyzer The change did not constitute a change to the facility as J-HPA-E02, and the Hydrogen and Oxygen Analyzer. The purpose described in the FSAR. The storage racks do not affect any of the racks was to facilitate testing of this equipment equipment important to safety or safety related and were with a calibration gas. designed to meet Seismic Category iX requirements. 84-01-ZF-002 This change added eyebolt attachment points to the fuel .This change did not introduce an unreviewed safety question. building 10 ton crane trolley in order to secure safety The addition of safety chains ensures a fuel pool gate will chains while lifting the fuel pool gates. The addition of not be dropped. This is not a change to the facility as safety chains was a NUREG 0612 cannittment to prevent described in the FSAR. possible dropping of the fuel pool gate.
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06/26/89 0 Design Change 50.59 Report PCP Number Description Sunna ry 85-01-ZR-007 This change replaced the existing hook block on the radwaste This change did not introduce an unreviewed safety question. 85-02-ZR-007 crane (NSRN-001) with a power-operated rotating hook block. This change did not constitute a change to the facility as 85-03-ZR-007 This allo~s the crane operator to remotely position the described in the FSAR ~ The 30 ton radwaste crane is not hook to pick up containers and minimize personnel exposure. safety related. Its failure Mill not impact any safety related equipment or systems.
l Page h . 1 06/26/89 Site Hodification 50.59'Report Site Hodification Number Description Safety Evaluation Smeary 1,2-SH-AF-002 This modification transferred the open indication switch for This modification did not introduce an unreviewed safety valve JAFA-NV-54 from rotor 2 switch 7 to rotor 4 switch 15 question. The change increases the reliability of the closed for enhanced position indication. This change was made in position indication on the valve and does not affect the response to INPO SOER 86-2. trip/throttle function of the valve. 2-SH-AR-002 This modification installed a shaft guard meeting the This modification did not introduce an unreviewed safety requirements of DSNA 1910.219.C.2 on post filter blower question. The addition of this shaft guard will not affect H-ARN-A02 (main condenser exhaust filter). the performance of the exhaust filtration unit H-ARN-FDT as described in FSAR Sections 10.4.2, 10.4.3, and 11.3. Neither the blower H-ARK-A02 nor filter H-ARN-F01 have any safety function. 1,2,3-SH-AS-001 This modification lowered the condensate cross-tie pressure This modification did not introduce an unreviewed safety safety valve setpoints (PSV25) to 100 psig. The system question. The Auxiliary Steam (AS) System is not described design pressure is 100 psig and the system normally operates in FSAR Section 9. The AS system is not quality related nor at 50 psig. safety related and does not affect any quality or safety related systems or components. This change does not affect the system operation. The change places the PSV setpoint in accordance with the system design pressure. 1,2-SH-CD-008 This modification reversed the polarity of computer input This modification did not introduce an unreviewed safety points CDJ11, 12, and 13 to provide correct readings. These question. Reversing the polarity of these points allows them points provide input for condensate punp kilowatts. to be correctly displayed by the Plant Computer. The change affects no important to safety equipment as described in FSAR Section 3.2. 1,2-SH-C0-010 This modification removes the annunciator and audible alarm This modification did not introduce an unreviewed safety for condenser reheat tray high levels There is no operator question. The purpose of the alarm is to alert the operators action specified for this alarm. The computer alarm point to a clogging of the strainer in the bottom of the condenser will remain to alert operators to a potential problem but false floor so action may be taken to clean the strainers at the annunciator and audible alarm are eliminated. the next refueling outage. Thus an audible alarm and annunciator are a nuisance to operators and serve no useful purpose.
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>Control Systems Not Required for Safety."
1,2,3-SH-RJ-003 This modification added trend recorder K-RJN-UJR-0019. The This modification did not introduce an unreviewed safety purpose of the trend recorder is to give operators a better question. The recorder serves no safety related function. indication of Assymetric Shape Index (ASI) osci llations. The recorder inputs are from the Plant Computer which is
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06/26/89 Site Modification 50.59 Report Site Hodification Nunber Description Safety Evaluation Sunnary also not safety related or iaportant to safety. The recorder is located in the control room and its mounting has been Seismicaly analyzed. 1-SH-RJ-005 This modification revised the software for the Core This modification did not introduce an mreviewed safety Operating Limit Supervisory Program to be consistent with question. The new software ensures core operating limits are the assumptions made for the Unit 1 Cycle 2 reload analysis. computed in accordance with the safety analysis performed for the Unit 1 Cycle 2 reload. 2,3-SH-RJ-013 This modification changed the Plant Honitoring System This modification did not introduce an unreviewed safety database to correct inaccurate information on the Metering question. The Plant Honitoring System is not iaportant to logs. The Metering logs record incoming and outgoing power safety. This change corrects inaccurate readings. and caqmte average megawatts. 1-SH-R J-014 This modification changed the Plant Honitoring System point This modification did not introduce an unreviewed safety configuration for the hourly and daily billing logs. This question. The chages were to the electrical billing points allowed accwulation of data for these points over the in the Plant Computer and do not affect any safety related required interval. or important to safety equipment. 3-SH-RJ-017 This modification deleted computer points NAYS4 and NAYS19 This modification did not introduce an unreviewed safety from the Unit 3 database. These computer points are only question. The plant computer is not safety related and the applicable to Units 1 and 2 and were inadvertently included computer points deleted were not applicable to the Unit. in the Unit 3 database listing. This change does not affect operation of the plant coaputer. 2,3-SH-RK-001 This modification deleted annunciator points AS14 and AS'15 This modification did not introduce an unreviewed safety for the Auxiliary Boiler from the computer data base in question. The alarms removed were not applicable to Units 2 Units 2 and 3. The Auxiliary Boiler is located by Unit 1 and 3. which maintains control of its operation. These alarms are not applicable to Units 2 and 3. 3-SH-RK-002 This modification installed a time delay in the Radwaste This modification did not introduce an unreviewed safety Trouble Alarm to the control room. The time delay allows the question. The alarm circuitry is non-class and is not Radwaste operators five minutes to acknowledge an alarm required for nuclear safety. before the control room recieves a trouble alarm.
Page No. 17 06/26/89 Site Hodification 50.59 Report Site Hodification Number Oescription Safety Evaluation Smeary 1,2,3-SH-RK-003 This modification changed the NSSS ESFAS anrxaciator logic This modification did not introduce an unreviewed safety to alarm only when an actual actuation has occurred. The question. The change is to a non-safety system which is for previous design would alarm when only a half leg actuation indication only and does not affect the actuation of any occurred. safety system. The nest logic will still anrxmciate an actuation but will eliminate false annuciation on a half leg trip. 1,3-SH-RK-004 This modification prevents two trouble alarms for the This modification did not introduce an unreviewed safety chlorine system from annunciating in the Hain Control Room. question. The alarms deleted are not required for personnel The equipment being monitored is under the jurisdiction of or plant safety and are not included in the FSAR. the Mater Reclamation Facility (MRF) operations personnel. No response to the alarm is required by Unit operations personnel. 'l,2(3-SH-RH-OOl This modification added a mimic of the Safety Injection This modification did not introduce an unreviewed safety System to panel RHNB02 to aid the operators in visualizing question. The mimic is an operator aid only and does not the system. affect operation of the Safety Injection System. The weight of the mimic on the panel is insignificant. 1-SH-RH-005 This modification replaced existing Control Room naitlpolnt This modification did not introduce an unreviewed safety recorders in Panel B07 with more reliable microprocessor question. The new recorders meet all applicable industry based rocorders. The modification was implemented in all standards, vendor qualification, and seismic qualifications. three units during this reporting period and affected FSAR Since the recorders are for monitoring only, their failure Section 'l0.2. would not prevent a PPS/ESFAS initiation. 2-SH-RH-007 This modification enhanced the control room namepiates and This modification did not introduce an unreviewed safety mimics. question. The change to the nameplates and mimics improves the operators ability to safely operate the plant by making it easier to identify components and visualize the system.
'I-SH-RH.010 This modification provided demarcation and labeling on panel This modification did not introduce an unreviewed safety 1JRHNB01 to group the Unit 2 and 3 bus controls. The pupose question. This modification provides demarcation and
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06/26/89 Site Hodification 50.59 Report Site Hodification Hwber Description Safety Evaluation SuImary of the change was to aid operators in discriminating labeling to group the Units 2 and 3 bus controls (on a Unit controls associated with the electrical busses of Units 2 1 control panel) and results in overall visual enhancement and 3. This change is part of the control room hunan factors that will improve the control room operators ability to program. prevent accidents. The physical change itself does not impact any safety or non-safety operations of the plant. 1,2,3-SH-SA-002 This modification installed wiring strain relief clamps in This modification did not introduce an unreviewed safety the Balance of Plant Engineered Safety Features Actuation question. The wiring strain relief clasps do not affect System (BOP ESFAS). The purpose of the modification was to system operation. The clamps do not affect the Seismic improve connector and wiring life. This change was qualification of the cabinet. recomnended by the vendor. 2-SH-SB-002 This modification replaced existing ESFAS s~itch blocks with This modification did not introduce an unreviewed safety gold sliding contact switch blocks. The purpose of the question. Gold sliding contacts will maintain the trip paths change is to prevent spurious ESFAS actuations. better than the heavy duty silver contacts previously installed. The actuation of the contacts will be the same as described in the FSAR to initiate a safety actuation. The switch contacts are normally closed and a fai lure of the contacts auld result in an action to the safe state. 1,2,3-SH-SB-003 This modification connected the reference buses for Foxboro This modification did not introduce an unreviewed safety Cabinets J-SBA-C02A, SBB, SBC, and SBD to the instrunent question. The grounding of the system reference busses of ground bus. The purpose of this modification was to the process protective cabinets does not have an adverse eliminate a signal attenuation problem affecting inputs to affect on the operation of the process protective cabinet or the CPCs and Plant Computer. the associated equipment. This has been demonstrated by testing. 1,2,3-SH-SB-005 This modification changed the Plant Protection System (PPS) This modification did not introduce an unreviewed safety power supplies to annunciate a bistable po~er supply that question. The changes made to the power supply do not affect has failed by overvoltage. A credible new power supply its operation except to annunciate a failure and prevent a failure mode (overvoltage) was identified which would not be po~er supply which has failed in the overvoltage condition annunciated under the previous power supply design. The new from affecting or potentially damaging other equipment. This design will detect an overvoltage failure and clamp the improves the system reliability by providing indication of a output voltage of the failed power supply to 2 volts which power supply failure in either the under or overvoltage
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Page No. 19 06/26/89 Site Hodification 50.59 Report Site Modification Number Description Safety Evaluation gunnery will annunciate the failed power supply. condition 1 g 2, 3-SH-SB-007 This modification increased the time delay for the Steam This modification did not introduce an unreviewed safety Generator Low Level pretrip alarm and trip of the reactor by question. The new time delay of 550 miliseconds is less than replacing R223 and R250 on the bistable comparator card. The the Technical Specification requirement of 1150 miliseconds. purpose of this modification was to prevent a reactor trip The function of the circuitry and operational design has not following a large load rejection. been changed and no new failure modes have been introduced by this change. The new resistors meet the same specifications as those originally installed. 1,2,3-SH-SB-008 This modification reconfigured the Control Element Asserbly This modification did not introduce an unreviewed safety (CEA) Reed Switch Position Transmitters (RSPT) power supply question. This change did not affect the design function or jmpers to increase the accuracy of CEA position indication. reliability of the RSPTs. The changes icprove the accuracy of the CEA position indication system and thus provide a more accurate signal to Core Protection Calculators (CPC) and Control Element Assembly Calculators (CEAC). This enhances their ability to perform their safety fmctions. 1-SM-SB-009 This modification implemented Revision 4 Software for the This modification did not introduce an unreviewed safety Core Protection Calculators in Unit 'I for Cycle 2 question. The Revision 4 software has been thoroughly operations. The changes are being made consistent with the tested. The revision does not change the generic algorithms Cycle 2 reload analysis. previously approved by the NRC. 1(2-SH-SC-006 This modification provided sample capability to the This modification did not introduce an unreviewed safety Chemistry Cold Lab of condensate demineralizer effluent question. This change was made to the condensate sampling upstream of the demineralirer bypass and chemical injection system and did not affect any safety related or important to points. safety equipnent. No changes were made to system operation other than the location of a sample point. 1,2,3-SH-SC-007 This modification installed a key lock override switch to This modification did not introduce an unreviewed safety allow rejecting condensate to the Condensate Storage Tank question. The components affected by this change do not have during of normal conditions. a safety function or safety design bases. 3-SH.SG-001 This modification was previously reported as item 128 of the
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06/26/89 Site Modification 50.59 Report Site Hodification Nunber Description Safety Evaluation Smeary 1986 50.59 report. 1,2,3-SM-SG-002 This modification performed the following items on the Hain This modification did not introduce an unreviewed safety Steam Safety Valves (MSSV): 1. Reduce insulation per vendor question. The function of the safety valves will not be recamendation, 2. Replace brass cotter pin with stainless affected in any way. The manual lifting device is never used steel per vendor recamendation, 3. Remove manual lifting and has been known to cause safeties to stick open. lever per vendor instructions and IEH 84-33. The pupose of Substitution of stainless steel for brass cotter pin the above modifications was to prevent insulation from alleviates material c'arpatability concerns. Insulation covering vent holes in the valve body, eliminate a potential removal is for personnel prot'ection. It has been blown off material Incompatibility, and prevent the manual operator several valves if packing leakage occurs upon Lifting. from holding the valve open after is has lifted. 2-SH-SG-003 This modification was previously reported as item ramber 163 of the 1987 50.59 report. 1,2,3-SH-SG-009 This modification replaced the needle valves in the This modification did not introduce an unreviewed safety pneunatic controls of the Steam Bypass Valves with question. The needle/check valves do not change the needle/check valves to improve valve position modulation. operation of the system. This modification was reccmnended by the valve vendor. There are no important to safety components in the Steam Bypass Control System (SBCS). 1-SH-SG-012 This modification changed the setpoint of 1J-SGH-PSV1127 This modification did not introduce an unreviewed safety from 700 psig to 730 psig. This change was required to question. The change in setpoint is allowed under ASHE code prevent normal variations in the Nitrogen supply system Section VIII. There is no change to the operation of the pressure from lifting the relief. system except to prevent unnecessary lifting of relief valve 1J-SGH-PV1127 and subsequent loss of nitrogen. 3-SM-SI-008 This modification adjusted the motor operator limit switch This modification did not introduce an unreviewed safety setting for valve 3JSIB-UV-615 to obtain the opening stem question. The adjustment of the limit switch ensures the travel as required by 73ST-3SI01 ~ valve will come to the proper safety related position as defined by Surveillance Procedure 73ST-3SI01. 2-SH-SI-009 This modification provided limit switch settings for valve This modification did not introduce an unreviewed safety operators of 2JSIA-UV-664 and 2JSIB-UV-665. It also rewired question. The change will not affect the operability of the
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Page No. 21 06/26/89 Site Hodification 50.59 Report. Site Modification Nunber Description Safety Evaluation Suanary the Safety Equipment Acutuation Status input from the valve valves. These valves are normally open and close on an RAS ~ operators to provide the correct valve position indication. SEAS indication previously did not provide a true indication of closure since it was taken from a mid-position contact. This change provides clear indication of valve status. The limit switch settings provided will iaprove the valve reliability. 2-SH-SI-022 This modification reworked pipe support SI-178-H-008 to This aodif ication did not introduce an unreviewed safety allow for thermal growth of the piping. question. This change restored the piping support to its as designed and as analyzed condition. No safety analysis assunptions are affected. A-SH-SK-001 Safeguards Information Safegaurds Information A-SH-SK-005 Safeguards Information Safeguards Information A-SM-SK-012 Safeguards Information Safeguards Information A-SH-SK-027 Safeguards Information Safeguards Information 1,2,3-SM-SP-002 The description and suanary of this change was included as item 51 of the 1987 50.59 report. 1-SH-SP-005 This modification extended the essential spray pond This modification did not introduce an unreviewed safety corrosion rack to a depth of 3" belo~ the Technical question. Corrosion racks are necessary as part of the spray Specification low limit for pond water depth. This change pond monitoring program of Technical Specification 6.8.4.f. was necessary to prevent alternate wetting and drying of the Thus accurate corrosion rate data is required. The only way corrosion coupons which produces incorrect corrosion rate to ensure accurate data is taken is to implement this data. modification. The modfication was evaluated for its potential impact on the safety analysis and no new failure mode was identified which could impact operation of the spray ponds',3-SH-SQ-003 This modification provides a time delay relay in the trip This modification did not introduce an unreviewed safety
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06/26/89 Site Hodification 50.59 Report Site Hodification Number Description Safety Evaluation Swmary circuit of radiation monitor RU0007A. The time delay question. The function of RU-0007A is not affected except prevents spurious trips of RU-0007A. When RU-0007A trips it that momentary spikes above the trip setpoint will not cause diverts the auxiliary steam condensate to the liquid a trip and diversion of the auxiliary steam condensate to radwaste system. Preventing spurious trips reduces the the liquid radwaste system. An actual increase in activity amount of liquid radwaste which has to be processed. in the auxiliary steam condensate will still actuate a trip and correctly divert condensate to the liquid radwaste system. Thus system operation is unaffected by this change and no safety analysis assumptions are affected. 3-SH-SQ-007 This modification was previously reported as item 53 in the 1987 50.59 report. 1-SH-SQ-008 This modification removed voltage from the check source of This modification did not introduce an unreviewed safety the fuel building radiation monitor. The purpose of this question. The check source will still be able to be was to eliminate undesirable ramping of radiation 'hange activated to perform the monthly check required by Technical rate on the fuel building radiation mointor causing the Specification 3.3.3.9. monitor to be declared inoperable. 1,3-SM-SQ.025 This modification added software to the Radiation Monitoring This modification did not introduce an unreviewed safety System mini-computer to allow it to perform hourly system question. The software change does not affect the function status print outs. of the RHS mini-computer except to give it the capability to provide hard copy storage of the hourly averages. 1,2-SH-SS-006 This modification added two ball valves and a check valve to This modification did not introduce an unreviewed safety provide additional isolation capability for the argon gas question. The function of the sampling system is unaffected. supply to the Nuclear Sampling System preparation panel. The The addition of these valves only provides a defense in purpose of this modification was to prevent sample fluid depth approach to prevent contamination of the argon system from entering the argon gas system. by back Leakage of RCS sample fluid into the system. The portion of the sample system affected by this modification performs no safety function. E 1-SH-SV-003 This modification changed the size of the reactor coolant This modification did not introduce an unreviewed safety pump vibration monitor proximity probe bracket. The purpose question. This change was only a minor modification in the of the change is to prevent the bracket from having a proximity probe bracket to change its resonant frequency
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Page No. 23 06/26/89 Site Hodification 50.59 Report Site Hodification Nunber Description Safety Evaluation Sunnary resonant frequency corresponding to the vane passing away from the vane passing frequency of the purp. The frequency of the Ixmp. Having a resonance in the proximity vibration monitoring system is not safety related and probe bracket near the vane passing frequency produces performs only monitoring of the vibration. No safety higher than actual pump vibration readings. analysis assumptions are affected by this modification. 3-SH-ZC-003 This modification provided for coping of a channel (stair This modification did not introduce an unreviewed safety stringer) to clear an interference with existing electrical question. Engineering reviewed the original design condu I t e considerations and determined that the coping of the stairway channel will not affect the structural integrity of the stair~ay. Thus no affect on its Seismic qualification or continued function. 1-SH-ZF-001 This modification provided the folowing changes to the Fuel This modification did not introduce an unreviewed safety handling crane: 1. Eliminate the "Geared Hoist Limit Bypass" question. The change is required to ensure that no hoist and change the keyed selsctor switch to a aGate Mode.a This loads in excess of 2000 Lbs are taken over the spent fuel new aode will allow for lifting the gates around the spent pool as required by Technical Specification 3/4.9.7. It adds fuel pool area as required during maintenance and refueling interlocks to prevent crane long and cross travel if the activities. This switch will bypass the 5000 lb. 'load hoist is operating as required in FSAR Section 9.1.4.2.2.17. interlock of the "Fuel Container" area and will be ~ All modes of crane operation with this new change are administratively controlled; 2. Add additional electrical previously stated and analyzed in the FSAR. interlocks to ensure that the crane long and cross travel will be prevented if the hoist is withdrawing or inserting a fuel assembly into its storage position; 3. Add electrical interlocks to the crane long and cross travel controls to ensure that anytime over 2000 lbs. is on the crane hoist, travel into the spent fuel area will be prohibited. In addition, if the hoist Load is over 5000 lbs. travel into the fuel container area will also be prevented, unless the crane is switched to the new "Gate Mode." 1-SH-ZR-002 This modification installed an 8" thick lead shield wall in This modification did not introduce an unreviewed safety the waste disposal area of the radwaste building at the question. The addition of the shield wall in the radwaste This shielding was required for the storage of100'levation. building does not affect the operability of any safety high level radioactive resin in this area. related, safe shutdown, or important to safety systems.
4 Page . 24 06/26/89 Site Modification 50.59 Report Site Modification Number Description Safety Evaluation gunnery 1-SM-ZT-001 This modification rerouted cable tray 1EZT1ENTDAC to gain This modification did not introduce an unreviewed safety enough slack to reterminate cable 1ETC01NC1CA. Cable question. The FSAR does not address routing of non class 1ETC01NC1CA was damaged and some cable slack was necessary cable trays. The routing of the cable trey does not affect to remove the damaged portion and reterminate the cable. any safe shutdown circuit nor is any safety related equipment in its vicinity. There is no change to operation of any equipment and all plant design criteria are maintained. 1-SM-ZY-002 This modification added a 2" ABS sleeved penetration through This modification did not introduce an unreviewed safety the east wall of the spray pond metering ixnp house to question. The spray pond punp house is non-quality related provide for a chemical supply line. and contains no safety related equipment. A-SM-ZY-005 This modification installed a 50'ntenna adjacent to the This modification did not introduce an unreviewed safety east end of the Adninistrative Annex. The purpose of the question. The antenna is located outside the protected area antenna is to provide a digital data camanications link and does not affect any safety related or important to between Palo Verde and Phoenix. safety structures, systems, or components.
06/26/89 Temporary Hodification 50 '9 Report Temporary Hodification Number Description Safety Evaluation Spermary 1-87-SH-033 This .temporary modification removed the trigger/switch card This modification did not introduce an unreviewed safety of the digital cassette accelerogragh. The purpose of this question. The siesmic system does not perform any safety modification was to determine the source of a spurious alarm function or initiate any automatic safety function. The of the system. siesmic system is an information system only, the failure of the system does not affect any other system. 1-88-CN-020 This temporary modification threaded the pipe nipple This modification did not introduce an unreviewed safety downstream of valve 1PCNNHV0923 and added a pipe cap. The question. The addition of the pipe cap which isolates the pipe cap was installed to prevent leakage of the valve. valve did not present any safety problems because the This valve is part of the system that vents gaseous radwaste pressure could be relieved by removing the cap and venting from the Reactor Drain Tank (RDT). The addition of the pipe or by using the feed and bleed method. The rupture disc cap resulted in the isolation of the valve. The alarm on the RDT provided over-pressure protection. The valve response procedure for high RDT pressure was revised to does not have a safety related function and its isolation remove the pipe cap prior to venting the system. will not adversely impact any safety related equipment. 1-88-NA-062 This temporary modification installed Cooper Roller bearings This modification did not introduce an unreviewed safety in Air Filteration Unit (AFU) 1H-NAN-J01B (Auxiliary question. Per UFSAR Table 3.2-1, these filter units are Building Normal AFU), in lieu of SKF spherical bearings not safety related, nor isyrotant to safety. There is no shown in vendor manual H7218-602. The SKF bearings had a safety design basis for the Auxiliary Building normal AFU poor maintenance history and the Cooper bearings are in UFSAR 9.4.2. In addition, this new style bearing should designed for heavier loads and higher speeds. extend fan and motor bearing life. 1-88-NA-069 This temporary modification replaced existing ginor flow This modification did not introduce an unreviewed safety switch (1J-HAN-FSL-0129) with a Hcgonnell F Hiller sail type question. The new flow switch will function the same as switch. The existing Alnor flow switch is no longer the previous switch. The system is non-quality related and manufactured and the HcDonnel L Hiller flow switch is an does not change the safety design bases of the UFSAR acceptable substitute. Chapter 9.4.2.1. 1-88-NC-019 This temporary modification installed a 60 inch blind flange This modification did not introduce an unreviewed safety on the discharge bell of the Control Element Drive Hechanism question. This modification allows for CEDH cooling with (CEDH) cooling fan 1MHCNA02A. The purpose of this one CEDH cooling fan failed. CEDM cooling is not a safety modification was to provide CEDH cooling with a single fan. related function. The blind flange was contructed to meet
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06/26/89 0 Temporary Modification 50.59 Report Teaporary Modification Nmher Description Safety Evaluation Sugary Seismic Category XI requirements and to preclude the failure of any safety related components. 1-88-HR-066 This teaporary modification installed a teaporary trailer, This modification did not introduce an unreviewed safety to decontaminate tools and equipment, in the Unit 1 fenced question. The trailer is located outside the walls of the in Radiation Controlled Area (RCA). The exhaust from this radwaste building and away from any safety related trailer was routed into the radwaste. building HVAC system to equipment which might be damaged by the trailer during a prevent a potential urn'onitored release of radioactivity. Design Basis Event. The trailer was evaluated as a potential missile and found to be of no significance. The trailer does not interface with any equipment important to safety. 1-88-MT-004 This tecporary modification disabled the annulation for This aodification did not introduce an unreviewed safety the Torsional Vibration Monitoring System (TVMS) while it is question. The TVMS is for monitoring of the Main Turbine out of service. and performs no safety fumtion. The system is not described in the FSAR or Technical Specifications. Removing anrxaciation while the system is out of service eliminates a source of spurious alarms in the control room. 1-88-HA-043 This temporary modification provided temporary power to This modification did not introduce an unreviewed safety several buildings during a planned outage of bus 1EHANS06. question. This temporary hook-up only affected non-class 1E power to non-safety related equipment and buildings. 1-88-HA-058 This temporary modifcation is the same as 1-88-)A-043. 1-88-HG-022 This temporary modification provided non-class power to load This modification did not introduce an unreviewed safety center 1ENGNL18 from load center 1EHGHL01 for purposes of question. The change only involved connecting two powering lighting panels 1EOAHD2 (100'Turbine Bldg.), D4 non-class busses. which are not isportant to safety or (176'urbine Bldg.), 008 (100'ontrol Bldg.), and D09 safety related. The change maintained lighting to various (120'adwaste Bldg.). This was necessary to maintain areas of the plant which improves operations ability to lighting and convenience outlets during an outage of bus perform tasks in these areas. 1ENANS02.
06/26/89 Temporary Modification 50.59 Report Temporary Modification Nunber Description Safety Evaluation Senary 1-88-HG-04C This terporary modification supplied power to panel This modification did not introduce an unreviewed safety AEHGNLCC, in particular the associated switchyard question. Ko safety related power systems or safety multiplexers, due to an outage of bus 1EHAHS06. The power related equipment were affected by the temporary po~er was provi'ded by the use of a station mobile diesel provided. generator. 1-88-NG-046 A temporary modification was required to supply power to bus This modification did not introduce an unreviewed safety AEHGNL43, in particular the startup yard multiplexers and question. No safety related power systems or safety respective multiplexer air conditioning, due to an outage of related equipment were affected by the supply of temporary 1ENANS05. power. 1-8&-KG-048 This temporary modification supplied power to normal This modification did not introduce an unreviewed safety lighting panels, which provide Lights and convenience question. This temporary power hookup used non class 1E outlets, due to an outage of bus 1ENANS01. 480 VAC power and fed only non safety related equipment. 1-88-NG-050 This temporary modification provided power to the This modification did not introduce an unreviewed safety contairment polar crane due to the outage of the 13.8 Kv bus question. This change only affects the non-class IE 480 1ENANS01. VAC power system as described in FSAR Section 8.3.1.1.1.. 1-88-NG-051 This temporary modification provided power from non-class This modification did not introduce an unreviewed safety Load center AEHGNL45 to non-class load center AENGNL48 due question. Ho safety related power systems per UFSAR to an outage 1EHANS05. This modification provided teaporary Section 8.3.1.1.1 or safety related equipment per UFSAR limited power to the Service Building prioritized loads. Section 3.2.1 were affected by this change. 1-88-NG-054 This terporary modification provided additional Load This modfication did not introduce an unreviewed safety capabi lityto Temporary Modification 1-88-NG-051. This question. Ho safety related power systems or safety modification was required due to the extended outage of related equipment was affected by this modification. 1ENAHS05 ~ 1-88-NG-056 This temporary modification is the same as modification 1-88-HG.044 This temporary modification supplied power to Loadcenter
'-88-NG-059 This modification did not introduce an unreviewed safety
I J Page No. 06/26/89 0 0 Temporary Modification 50 '9 Report Temporary Modification Number Description Safety Evaluation Summary AENGNL48, which provides power to the Administration and question. This teaporary change affected only non class Service Buildings, due to an outage on bus 1ENANS06. 1E 480 VAC power systems and impacted no safety related equipment. 1-88-NH-002 This temporary modification provided power to 1ENHNM0346 This modification did not introduce an unreviewed safety from 1EHHNH5028 due to an outage of bus 1EHAHS05. question. This temporary change affects only non-class 1E po~er as described in UFSAR Capter 8. 1-88-HH-023 A teaporary modification was made in Unit 1 during this This modification did not introduce an unrevieled safety reporting period to provide temporary power to feed the question. Fault current protection will be provided on contaiteent equipment hatch 1C-ZCN-M03 during the outage of both the equipment feed and the temporary cable feed. Load bus 1E-HAN-S02. This affected drawing 13-E-HHA-072, which is shed of SiAS/CIAS is provided on the motor control center . w incorporated into FSAR by reference. feeds. The change in po~er source is non-class to non-class and does not supply safety related equipment. 1-88-HH-024 This temporary modification provided non-class power to the This modification did not introduce an unreviewed safety t non-class liquid nitrogen pumps 1-M-GAN-P01A S B during an question. The modification changes the electric supply outage of bus 1ENANS02. source from an unavailable non-class motor control center (MCC) to an onsite construction power panel. The nitrogen pumps are not required for safe shutdown of the plant and no safety related equipment was affected. 1-88-HH-025 This temporary modification is the same as modification 1-.88-NH-023.
'l-88-NH-027 This teaporary modification provided power to the digital This modification did not introduce an unreviewed safety Radiation Monitoring System (RMS) during the rework of bus question. This change affects only non class power and 1EHANS02. maintained the function provided by the digital RHS.
1-88-NH-028 This temporary modification supplied power to the local This modification did not introduce an unreviewed safety plant rmltiplex battery charger 1EHANN01 due to the loss of question. The power supply to the battery charger is a power to bus 1ENANS02 ~ non class 1E 480 VAC supply which was temporarily replaced with an alternate non class 480 VAC source. No safety related systems or components were adversely impacted by
06/26/89 Temporary Modification 50.59 Report Temporary Modification Nwher Description Safety Evaluation Surmary this modification. 1-88-NH-031 This temporary modification provided po~er to the Slowdown This modification did not introduce an unreviewed safety High Total Dissolved Solids (TDS) Sump Pumps during the question. This modification only impacts a non class 1E rework of bus 1EHANS02. 480 VAC system and affects no safety related equipment. The sunup pcs were powered as before with a different Motor Control Center. 1-88-NH-033 This temporary modification installed power to the Reactor This modification did not introduce an unreviewed safety Makeup llater Pmp, 1MCHNP038, due to an outage on bus question. Only non class 1E power systems and non safety 1ENANS02. related equipment were affected by this change. 1-88-NH-034 This teaporary modification supplied power to the HVAC units This modification did not introduce an unreviewed safety for the Post Accident Monitoring Units (PAHU) due to an question. This power hookup affects only the non class.1E outage on bus 1ENAHS02. This modification provided 480 VAC power system as described in UFSAR 8.3.1.1.1 and assurance that the PAMU would remain operable, as required. assures the PAMU remains operational during hot weather conditions ~ 1 1-88-NH-036 A temporary modification was provided to supply power to This modification did not introduce an unreviewed safety M CPN M06 from E-NHH-D25 due to the loss of E-NNH-D26 as a question. This change only affects non 1E panels which result of the failure of the Auxiliary Transformer and the are part of the non 1E 480 VAC system as described in loss of 1ENANS02. UFSAR 8.3 ~ 1.1.1 ~ 1-88-NH-037 A temporary modification was made in Unit 1 during this This modification did not introduce an unreviewed safety reporting period to provide power from the S01 train for the question. There was no change to the operation of the SQ RU-7. This modification effected drawing 13-E-NHA-020, which system. The change involved temporarily powering RU-7 from is incorporated into the FSAR by reference. an alternate source to allow maintenance on the normal power source. The modification allowed continuous operation of RU-7 during the outage. 1-88-NH 041 This temporary modification involved supplying power to the This modification did not introduce an unreviewed safety Containment Power Access Purge Output Danper 1HCPNMOSA due question. This power hookup only affects non-class 1E to an outage on bus 1ENANS02 ~ power and is is provided to non safety related equipment.
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Page No. 6 06/26/89 Temporary Nodification 50.59 Report Temporary Nodification Nunber Description Safety Evaluation Smeary 1-88-HH-045 This temporary modification provided po~er to panel lENHHD03 This modification did not introduce an unreviewed safety from 1ENHHH5011 during an outage of 13.8 kV bus 1ENAHS01. question. This teaporary change affects onLy non-class 1E po~er as described in UFSAR 8.3.1.1.1. 1-88-NH-047 This terporary modification provided power to battery This modification did not introduce an unreviewed safety charger 1ENKNH19 and to aaltiplexer 1EHGNQ01 from a portable question. The power hookup involved only non-class diesel generator. This teaporary power hookup was required electrical systems as discussed in UFSAR Chapter 8. due to the outage of 1ENANS05. '1-88-NH-049 This temporary modification supplied power to the Radwaste This modification did not introduce an unreviewed safety Building and Normal Control Room Air Handling Unit Fan due question. This teaporary power hookup affected only non to an outage on bus 1EHANS01. class 1E 480VAC po~er system as discussed in FSAR Section 8.3.1.1.1 which fed non safety related equipment. 1-88-NH-052 This temporary modification supplied power to the Control This modification did not introduce an unreviewed safety Building Normal Air handling unit during the outage of the question. No safety related equipment or power systems 13.8 kV bus 1EHAHS01. were affected by this modification as described in FSAR Table 3.2.1 and Section 8.3.F 1.1. 1 1-88-HK-021 This temporary modification jtmpered out failed cell 17 of This modification did not introduce an unreviewed safety non-class battery 1ENKNF18. This left the battery in a 59 question. The capacity of the battery was originally versus 60 cell configuration. 109.68'. Removal of one cell will not reduce this to a level belo~ that required in the original specification. No safety related or important to safety equipment was affected by this modification. 1-88-RC-001 This temporary modification installed a jwIper on RTD This modification did not introduce an unreviewed safety 1JRCNTE0154 to permit continued operation of the RTD with an question. The RTD will still perform the same function open circuit on one leg of the compensation leads. This will with only a neglegible loss of accuracy due to the result in continued temperature indication with a neglegible shortened compensation lead. The RTD is for equipment effect on accuracy. protection and does not perform any control or safety function.
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Page No. 06/26/89 Temporary Hodification 50.59 Report Temporary Hodification Number Description Safety Evaluation Sunnary '1-88-RC-060 A teaporary modification was made in Unit 1 during this This modification did not introduce an unreviewed safety reporting period to remove RTD 01JRCBTE0125 and cabling from question. All of the fmctions of 01JRCATE0122CA/C1 will thermowell 01JRCBTM0125 and install dual element RTD be performed with the RTD in its new location. The removal 01JRCATE0122CA/Cl in its place. The original RTD for of TE-125 RTD results in the bistable that activates the 01JRCATE0122CA/C1 was broken in its.thermowell and could not 8-train portion of the Low Teeperature Over Pressure be completely removed rendering the RTD and thermowell (LTOP) annunciator being inoperable. A toggle switch is unusable. Consequently, a new RTD was obtained and installed being teaporarily installed in the annunciator cabinet to in the themowell previously occupied by 01JRCBTE0125. The perform this function and will be adminstratively new RTD will fuxtion as 01JRCATE0122CA/C1. controlled by procedures 410P-OZZ01 and 410P-1ZZ10 such This modification affected FSAR Section 5.1. that the B-train LTOP annunciator will be enabled whenever RCS temperature is less than 300 degrees Farenheit. The A-train annunciator is not affected. No separation or seismic criteria are violated. 1-88-SB-003 This teaporary modification installed a voltage disturbance This modification did not introduce an unreviewed safety analyzer to monitor 120VAC input power and 12VDC bistable question. The four channels of Plant Protection System logic power of PPS Channel C (SBC-C01) in Unit 1. The (PPS) are separated electrically and physically such that purpose of the modification was to troubleshoot spurious a fault in one channel will not prevent the PPS from Channel C trips. performing its safety function. This modification only affects one channel. Technical Specifications allow operation with one channel of PPS inoperable. 1-88-SG-063 This temporary modification installed a check valve and This modification did not introduce an unreviewed safety relief valve downstream of the quick open "RosT<<Solenoid question. The addition of the check valve and relief valve valve on steam bypass valves 1001 and 1004 to enhance the allow the steam bypass valves to meet their modulBte close valves modulation capability. time specification. The steam bypass valves are not important to safety or safety related. This change does not introduce any new failure modes to the steam bypass control system.
'I SG-068 This teaporary modification installed four pressure This modification did not introduce an unreviewed safety connections on Steam Bypass Control System (SBCS) valves question. The SBCS is not iayortant to safety and defined PV1001 and PV1005. The purpose of the connections was to as not required for plant safety by CESSAR Chapter 7.7.
06/26/89 Temporary Modification 50.59 Report Temporary Modification Number Description Safety Evaluation Swmary allow monitoring of the valve and valve operator during The pressure connections are for monitoring only and do testing. The system operation and control are not affected. not affect the operation of the valves or SBCS. 'I-88-SK-007 Safeguards Information 1-88-SK-029 Safeguards Information 1-88-SV-008 This temporary modification lowered the reactor coolant pap This modification did not introduce an unreviewed safety shaft vibration danger alarm setpoint to 'IOM. question. The lowering of the setpoint will provide an earlier warning of a problem with the RCP shaft. 1-88-SV-009 This temporary modification lowered the reactor coolant pump This modification did not introduce an unreviewed safety shaft vibration alert setpoint to 8M for sensors 01KSVHYSH21 question. The lowering of the setpoint allows more time to and 01KSVNYSH22. diagnose a problem with the RCP shafts and will alert the operators to a potential problem earlier. 1-88-SV-010 This temporary modification lowered the alert setpoint for This modification did not introduce an unreviewed safety reactor coolant pump sensors 01KSVHYSH27 and 01KSVNYSH28 to question. The lowering of the alert setpoints allows an 8M. earlier warning of a potential problem with the RCP shaft. 1-88-SV-011 This temporary modification lowered the danger setpoint of This modification did not introduce an unreviewed safety reactor coolant pump vibration sensors 01KSVHYSHH27 and question. The lowering of the setpoints will provide an 01KSVNYSHH28 to 10M. earlier warning of a problem with a RCP shaft and allow more time for operator action. 1-88-SV-012 This temporary modification lowered the alert setpoints of This modification did not introduce an unreviewed safety reactor coolant pump vibration sensors 01KSVNYSH25 and question. The lowering of the setpoint provides an earlier 01KSVNYSH26 to 8M. warning of a potential problem with a RCP shaft. 1-88-SV-013 This temporary modification lowered the danger setpoints of This modification did not introduce an unreviewed safety the RCP Shaft Vibration sensors 01KSVNYSHH25 and question. The lowering of the danger setpoints will alert 01KSVNYSHH26 to 10M. operators to a potential problem with the RCP shafts sooner than the previous setpoint. This system performs no control function and is used for diagnostics and
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06/26/89 Terporary Hodification 50.59 Report Temporary Nodification Number Oescription Safety Evaluation Smeary information only. 1-88-SV-014 This temporary modification lowered the alert setpoint of This modification did not introduce an unreviewed safety reactor coolant pump shaft vibration sensors 01KSVNYSH23 and question. The lowering of the setpoint provides an earlier 01KSVNYSH24 to BN. warning of a problem with a RCP shaft. 1 SV-015 This temporary modification lowered the danger setpoint of This modification did not introduce an unreviewed safety reactor coolant pump sensors 01KSVNYSHH23 and 01KSVNYSHH24 question. The lowering of the setpoint provides an earlier to 10H. warning of a problem with a RCP shaft. 1-88-SV-067 This temporary modhfication raised the setpoint of Loose This modification did not introduce an unreviewed safety Parts Vibration Honitor channel 5 to greater than 0.5 ft-lbs question. IJhile this modification is implemented Channel 5 to reduce the nunber of spurious alarms. of the LPVMS is considered inoperable and the action statement of Lco 3.3.3.7 will be implemented. This is an anlyzed condition with provisions for operation defined in the Technical Specifications. 2-87-HA-060 A temporary modification installed automatic grease feeders This modification did not introduce an unreviewed safety to the fan and motor bearings for Auxiliary Building Normal question. The AFUs are not safety related nor would the Air Filtration Units (AFU), 2N-HAN-J01A and J01B, in place lubrication frequency of the bearings adversely impact any of the ZIRC fittings. This modification was performed due safety related fmction. to the frequent failure of the bearings. Excessive lubrication frequencies of the bearings was suspected of contributing to the failures. 2-87-HR-057 This temporary modification replaced zirc grease fittings This modification did not introduce an unreviewed safety with automatic grease feeders on Air Filtration Units (AFU) question. The new automatic grease feeders should improve 2H-HRN-J01A Ec BE the reliability of the AFUs by providing continuous lubrication. Per FSAR Table 3.2-1, the system is non-safety, it is not mentioned in FSAR 6.4
~~Habitability," and per FSAR 9.4.3, it has no safety design basis; 2-87-SG.062 This temporary aodification installed a new design solenoid This modification did not introduce an unreviewed safety
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06/26/89 Temporary Modification 50.59 Report Temporary Hodification Nunber Description Safety Evaluation Swmary topworks cover and position Limit switches on valve question. The new equipment is built to the same design 2J-SGB-UV219. This was done to rectify a continuing problem specifications and has no impact on relibility/operation. with dual indication experienced with Valcor solenoid The valve will be tested per ASHE Section XI to verify valves. operability. 2-88-CH-011 This temporary modification lowered the RCP 2A seal This modification did not introduce an unreviewed safety injection low flow alarm setpoint to 2.5 gpm from 6.0 gpm. question. The revised setpoint meets the flow requirements The seal injection flow to RCP 2A is below the current of the RCP seals. setpoint but has been evaluated by Engineering as acceptable. Thus the setpoint has been reduced to eliminate the continuous alarm condition and restore the alarm function of alerting the operators to a reduction in flow below that required for operation. 2-88-HA-016 A temporary modification was made in Unit 2 during this This modification did not introduce an unreviewed safety reporting period to install a jumper in motor control center question. The essential cooling system is designed to (NCC) E-PHA-H3726. The purpose of the jumper was to actuate provide cooling to the turbine driven auxiliary feedpump the essential cooling system for the auxiliary feedpump room room when the feedpunp is running. Actuating the essential to permit maintenance on the non essential cooling system cooling system placed it in its safety related condition and maintain acceptable room temperatures. This affected of providing room cooling. In the event the feedpump were drawing '13-E-HAB-006, which is incorporated in the FSAR by actuated the essential coolers would already be in reference. operation performing its safety related faction. 2-88-HA-019 This temporary modification added automatic grease feeders This modification did not introduce an unreviewed safety to fan and motor bearings for Air Filtration Unit (AFU) question. Per Table 3.2-1, these filter units are not 2H-HAH-J01A in place of zirc fittings. The purpose of the safety related, nor important to safety. There is no change was to evaluate if providing a continuous grease safety design basis for the Auxiliary Building normal AFUs supply would result in extended fan and motor life. in FSAR 9.4.2. Providing a continuous grease feed should extend fan and motor life. 2-88-HJ-002 A temporary modification was performed to remove four pairs This modification did not introduce an unreviewed safety of Control Room Pressure Boundary Isolation Dampers for question. Control Room Special Pressurization Test was maintenance and testing. Sheet metal blanks were installed performed after the installation of the blanks to ensure in place of the dampers and subsequent testing was performed integrity of the Control Room envelope. The presence of
Page No. 06/26/89 Temporary Modification 50.59 Report Temporary Modification Number Description Safety Evaluation Swmary to verify the reestablishment of the Control Room pressure the blanks essentially placed the Control Room envelope in boundary. an 'Isolated configuration and did not adversely impact other plant systems. 2-88-LR-001 This temporary modification was installed permanently by Site Modification 2-SM-LR-005. See 1,2-SM-LR-005 for description of this modification. 2-SS-Na-O2O This temporary modification provided uniterruptible 120 VAC This modification did not introduce an unreviewed safety power to the Unit 2 Test Data Acquisition System (TDAS). The question. The TDAS system is a non-quality related system power is being supplied from non-class 1E 120 VAC used for data gathering during testing and plant distribution panel 2E-NQNr001. transients. The power supplied to the system is from a non-class 1E source. TDAS is not included in the FSAR and no credit is taken for its operation in the mitigation of an accident or its use for safe shutdown. 2-88-QH-005 This temorary modification lifted cable AEQH02NC1KQ to This modification did not introduce an unreviewed safety isolate failed rectifier AEQHNH16. None of the above question. The lifting of leads to the failed rectifier equipment is quality related. does not affect any equipment important to safety or safety related. 2-88-RC-021 This temporary modification disconnected shorted This modification did not introduce an unreviewed safety proportional heater 2-M-RCE-811'and rewired the other question. The heater which was disconnected was a heaters in the bank in an open delta configuration. non-class 1E heater. No RCS boundary or class equipment is being affected. 2-88-SB-018 This temporary modification installed a digital storage This modification did not introduce an unreviewed safety recorder on Channel C Reactor Trip Switchgear (RTSG) shunt question. The digital recorder is for monitoring only and and undervoltage coils to determine the cause of spurious did not impact the operation of the RTSG. actuations. 2-88-SC-013 This temporary modification installed a chemical fwe This modification did not introduce an unreviewed safety exhaust system on the hydrazine and aaaenia day tanks in the question. The installation of the chemical fune exhaust Secondary Chemistry Control System. system does not affect the operation of the Secondary
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Page No. 06/26/89 Temporary Modification 50.59 Report Temporary Modification Number Description Safety Evaluation Summary Chemistry Control System. This modification does not affect any quality related structures, components, or systems (PVNGS FSAR 3.2 Classification of Structures, Components, and Systems, Table 3.2-1). 2-88-SK-007 Safeguards information 2-88-SP-010 This temporary modification changed the "A" train essential This modification did not introduce an unreviewed safety spray pond differential flow alarm setpoint to 1200 gpm from question. Engineering reviewed the design calculations for 775 gpm. This change is necessitated by flow annubar FI-7 the differerntial flow alarm and determined that any which has been determined to be reading higher than actual setpoint up to 1400 gpm will meet the design basis of the flow. This higher indicated flow was causing excessive system. The '1200 gpm setpoint therefore meets the system differential flow alarms in the control room. No replacement design bases. The alarm is designed to warn of a major annubar is available so the alarm setpoint is temporarily line break and will perform that function. raised until a new annubar can be procured. 2-88-SQ-003 This temporary modification installed a disk eaalating This modification did not introduce an unreviewed safety system, consisting of a controller and a memory module, in question. The addition of the disk emlating system the Radiation Honitoring System (RHS) minicomputer. This enhanced the operation of the RHS minicomputer. disk was provided to enhance the RHS minicomputer. 2-88-SQ-023 This temporary modification installed insulation and heat This modification did not introduce an unreviewed safety tracing on the Condenser/ Gland Seal exhaust sample lines to question. The heat tracing and insulation can only improve the Radiation Honitor RU-141. The purpose of the operation of RU-141 and does not affect the operation of modification was to maintain the sample flow tepperature any other equipment. This modification was evaulated for above the dew point and prevent moisture accuIxjlation in the its effect on safe shutdown systems and determined to have saaple lines. no effect on safe shutdown systems coeponents, or safety functions. 2-88-SV-014 This temporary modification reduces the alert and danger This modification did not introduce an unreviewed safety alarm setpoints for the reactor coolant pwp shaft question. The lowering of the setpoint will provide an displacement to 9H from '12M. The purpose of the reduction ealier warning of the potential of a RCP shaft shear was to provide early warning of a RCP shaft shear event. event. System operation is not affected.
Page No. 13 06/26/89 Temporary Modification 50.59 Report Temporary Hodification Number Description Safety Evaluat>on Smmary 2-88-SV-015 This temporary modification lowered the alert and danger This modification did not introduce an unreviewed safety setpoints for the reactor coolant Fop shaft displacement question. The lowering of the setpoints will provide an alarms from 9H to 7H. earlier warning of the potential RCP shaft shear event. 2-88-TC-006 A temporary modification was required to to assure operation This modification did not introduce an unrevicwed safety of the Instrument Air (IA) compressors during the Turbine question. The use of a temporary cooling water supply did Cooling Mater (TC) System outage. The Domestic lister (DS) not alter the operation of the IA compressors. The System was connected to the TC by rubber hoses through IA interface between this portion of the TC System and other compressors IAN-C01A and IAN-C01B only.. systems was not impacted. I 3-88-N.018 This temporary modification installed a carbon fiber drive This modification did not introduce an unreviewed safety shaft in cooling to~er fan 3HCWNA02H. This shaft replaced question. The new drive shaft performs the same function the existing stainless steel shaft which had experienced as the old one and does not affect system operation. The problems with thermal expansion. The new carbon fiber shaft cooling towers are not required to mitigate the eliminates thermal expansion as a concern. consequences of an accident or for safe shutdown. 3-88-MJ-005 This temporary nodif ication replaced the original lo~er This modification did not introduce an unreviewed safety extension shaft on daaper 3H-NJA-M02 with a shorter shaft question. The operation of the Control Room essential while maintenance was being performed on the daaper. The ventilation system remained as described in UFSAR 6.4 ~ purpose of this modificatiori was to provide Control Room During the time that the modification was installed the envelope isolation of the outside air makeup duct during the associated air filtration unit, 3H-NJA-F04, was considered maintenance. inoperable. 3-88-MA-001 This temporary modification installed jumpers across the This modification djd not introduce an unreviewed safety Hain Transformer Bushing Potential Device (PBA) and lifted a question. The PBA is not quality related and the PBA-X coil wire to prevent coil damage. This modification installation of the jumper was to increase the system was performed because the PBA was causing the cooling fans reliability. and pumps to cycle on and off. 3-88-MA-010 A temporary modification was installed to reduce the Main This modification did not introduce an unreviewed safety Transformer Control Cabinet internal temperature. This question. This system as a non safety related system. modification utilized a window air conditioning unit and The addition of the cooling was intended to increase the provided ducting for closed loop cooling circulation through system reliability and reduce occurences of generation
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Page No. 06/26/89 Temporary Hodification 50.59 Report Temporary Hodi f ication Number Description Safety Evaluation Senary the control cabinet. down-time caused by nuisance trips of overload relays and breakers. 3-88-HT-002 This temporary modification disabled the annunciator for the This modificat'ion did not introduce an unreviewed safety Torsional Vibration Honitoring System (TVHS) until the question. The TVHS is for monitoring Torsional Vibrations system is operational. The purpose of the TH was to in the Hain Turbine shaft and performs no control or plant eliminate a spurious control alarm. safety fmction. The system is currently out of service and having the alarm continuously present in the control room s'erves no useful purpose. 3-88-QF-017 A temporary modification was required to troubleshoot the This modification did not introduce an unreviewed safety remote Hultiplexer cabinet 3ENGNQ03 to determine the cause question. The Circulating llater System, as described in of erroneous signals which caused the main feeder breaker UFSAR Section 10.4.5, performs no safety related fmction. and several cooling tower fans to trip. This modification The loss of condenser vacwn due to a loss of circulating lifted the output breakers of the discrete output cards on water has been previously analyzed in'ESSAR Section the remote Hultiplexer cabinet. '15.2.3. 3-M-SC-012 This temporary modification is the same as 2-88-SC-013 except it is installed in Unit 3. 3-88-SC-014 This temporary modification installed a threaded carbon This modification did not introduce an unreviewed safety steel spool in place of a damaged flow meter. This question. The removal of the flow meter and the addition installation was necessary due to the rupture of the of the pipe spool allowed for the regeneration of the Condensate Demineralizer Acid Dilution water flow meter and demineralizers for Unit 3 power ascension. The affected no replacement flow meter was available. portion of the Condensate Clean up System has no safety related function. 3-M-SK-006 Safeguards Information 3-88-SQ-009 This temporary modification added soft~are to enhance This modification did not introduce an unreviewed safety troubleshooting of the Radiation Honitoring System question. The addition of the software only aids in mini-computer. troubleshooting RHS camanication loops and does not affect the other operations of the computer. Thus the system functions are not ispaired.
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Page No. 15 06/26/89 Temporary Modification 50.59 Report Temporary Modification Nunber Oescription Safety Evaluation Sunnary 3-88-SQ-020 A temporary modification was pehformed to remove the This modification did not introduce an unreviewed safety insulation and install a fielding preheating blanket on question. The insulation removal and addition of the RU-141 process inlet line. This modification was performed preheating blanket enhanced the ability to obtain a to reduce the moisture in RU-141 and RU-142. representative samples during cool ambient conditions and increases the reliability of the Effluent Honitoring System. 3-88-SQ-021 This tesporary modification is identical to 2-88-SQ-023 but was implemented in Unit 3. . 3-88-SV-007 This teaporary modification lowered-the alert alarm setpoint This modification did not introduce an unreviewed safety of the RCP Shaft Vibration sensors 03KSVHYSH25 and question. The lower setpoint wilt provide earlier
. 03KSVHYSH26 to 7M. annunciation of excessive shaft displacement. This will provide more time to diagnose a problem and take appropriate corrective action.
3-88-SV-008 This temporary modification lowered the danger alarm This sodification did not introduce an unreviewed safety setpoint for RCP Shaft Vibration sensors 03KSVNYSHH25 and question. The lower setpoint will provide earlier 03KSVNYSHH26 to 9M. indication of a problem with a RCP shaft and provide more time for appropriate corrective action to be taken.
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06/27/89 Procedure 50.59 Report for 1988 Procedure Number Description Safety Evaluation gunnery and Revision 75RP-9ZZ78, Rev 6 The change to Procedure 75PR-9ZZ78, Rev 6, This procedure revision did not introduce an unreviewed
<<Decontaimination,<< incorporates a log of contamination safety question. These changes to a procedure described in monitor alarms at the Radiological Controlled Area (RCA) the UFSAR ensure that records are maintained of all exits and changes the contamination limits for removable personnel who alarm the radiation monitor at the exit to the contamination, setting them equal to those for fixed RCA and provide clear and more conservative requirements contamination. related to contamination limits for removable contamination in line with IE Circular No. 8'l-07.
76TP-9ZZ01, Rev 0 This new procedure, 76TP-9ZZ01, Rev. 0, >>Secondary Side This new procedure did not introduce an unreviewed safety Decontamination System," describes the procedural controls question. The temporarily installed equipment interfaces placed on teaporarily installed equipment used to reduce the with permanently installed plant equipment described in the amount of radioactivity in the plant secondary side prior to UFSAR and is utilized to reduce radioactivity levels in the discharge to the evaporation pond. effluent discharged to the evaporation pond to within Technical Specification 3/4.11.1 levels when the secondary side of the plant is radioactively contaminated. The temporarily installed equipnent will not affect plant operations and is not required for safe shutdown of the plant. 73TP-9HJ05, Rev 0 This new procedure, 73TP-9HJ05, Rev 0, "Control Room This new procedure did not introduce an unreviewed safety Pressure Test Utilizing AFU Manual isolation Dampers,<< was question. The new procedure was prepared to test the new prepared to test the ability of new air filter unit (AFU) AFU isolation dampers to verify that they function as isolation dampers to isolate the control room and maintain described in both UFSAR and as specified in the Technical pressurization requirements as described in UFSAR section Specifications. 6.4.2.4.A and further specified in Technical Specification 4.7.7.d.3. 42TP-2ZZ02, Rev 0 This new procedure, 421'P-22202, Rev 0, <<Unit 2 Antimony This new procedure did not introduce an unreviewed safety Cleanup,<< describes the procedural controls placed on the question. The procedure temporarily modifies plant Antimony cleanup program institued in PVHGS Unit 2 to reduce configuration described in the UFSAR as well as normal plant the Antimony buildup in the RCS. chemistry. This process has been widely practiced within the industry for reducing Antimony buildup.
li 06/27/89 Procedure 50.59 Report for 1988 Procedure Number Description Safety Evaluation Smeary and Revision 73TP-1RC01, Rev 0 This new procedure, 73TP-1RC01, Rev 0, aCEA Exercise,4 This new procedure did not introduce an unreviewed safety details the procedural controls associated with manual question. During performance (,in refueling mode) of the exercising of Control ELement Assembly (CEA) 56 during procedure, the borom concentration is such that the shutdown maintenance activities on the Control Element Drive margin is well within Technnical Specification limits. Hechanism (CEDH) 56 housing. 72IC-2RX03, Rev 0 This new procedure, 72IC-2RX03, Rev 0, "Core Reloading,<< This new procedure did not introduce an unreviewed safety provides the detailed sequence for the refueling activities question. Changes in the core fuel pattern from the initial for PVNGS Unit 2. fueling of Unit 2 were accounted for in the analyses performed for Cycle 2 and submitted to the NRC in the Reload Analysis Report. 75RP-9ZZ78, Rev 5 This procedure, 75RP-9ZZ78, Rev 5, "Decontamination" This procedure revision did not introduce an unreviewed incorporates additional guidance relating to survey and safety question. These changes to a procedure described in contamination control for hot particles. the UFSAR provide additional guidance related to hot particle controL to ensure personnel radiation exposures are maintained ALARA. 440P-2SC02, Rev 0 This new procedure, 440P-2SC02, Rev 0, "Resin Transfer This new procedure did not introduce an unreviewed safety Between Condensate Demineralizer and a Vendor Vessel," question. This procedure controls the installati'on of provides instructions for transferring resin between the temporary equipment which interfaces with permanently Condensate Demineralizer System (CDS) and a vendor transfer installed plant equipment described in the UFSAR and is vessel. utilized to transfer resin between the CDS and a vendor transfer vessel. The temporarily instalLed equipment will not affect plant operations and is not required for safe shutdown of the plant. ANF-1123, Rev 0 This new procedure, ANF-1123, Rev 0, "Procedure for This new procedure did not introduce an unreviewed safety Ultrasonic Inspection of Fuel Assemblies at Palo Verde Unit question. The test, which is not described in the UFSAR, is 2," provides instructions for performing an inspection to performed to identify those fuel assemblies that have failed identify Unit 2 fuel assemblies that had failed (experienced during Cycle 1 so that they can be properly dispositioned. breached cladding) during Cycle 1. This activity does not introduce any safety concerns beyond those previously analysed in the UFSAR.
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06/27/89 Procedure 50 '9 Report for 1988 Procedure Nunber Description Safety Evaluation Summary and Revision ANF-P71,048,Rev 1 This procedure, ANF-P71,048, Rev 1, alndexing X-Y Table," This procedure revision did not introduce an unreviewed provides instructions for utilizing a system providing safety question. The test, which is not described in the remote positioning of probes used to perform varous UFSAR, is performed in association with the inspection of measurements on irridated fuel assemblies. spent fuel assemblies to identify failures. The activity does not introduce any safety concerns beyond those previously analysed in the UFSAR. PVNGS 400-007, This procedure, PVNGS-400-007, Rev 1, "Procedure for Visual This procedure revision did not introduce an unreyiewed Rev 1 Inspection of Fuel Rods and Configuration and In Pool safety question. The test, which is not described in the Movement of Rod Storage Baskets,"provides information and UFSAR, is performed to visually (by camera or periscope) instruction to enable the inspection of fuel rods contained inspect spent fuel assmblies, contained in a Rod Storage in a Rod Storage Basket. Basket, to identify any unacceptable conditions. This activity does not introduce any safety concerns beyond those previously analyzed in the UFSAR. PVNGS-400-008, This procedure, PVNGS-004-007, Rev 1, aProcedure for the This procedure revision did not introduce an unreviewed Rev '1 Removal and Installation of Upper End Fittings on System 80 safety question. The test, which is not described in the Fuel Assemblies," provides information and instructions UFSAR, is performed to remove and reinstall the fuel concerning the operation of various tools for the purpose of assembly upper end fitting to access the fuel rods for removing and installing upper end fittings on system 80 Fuel inspection and relpacement, and/or to access the guide tubes Assemblies. for shiaeing. This activity does not introduce any safety concerns beyond those previously analyzed in the UFSAR. PVNGS-400-009( This procedure, PVNGS-400-009, Rev 0, aProcedure for Eddy This new procedure did not introduce an unreviewed safety Rev 0 Current Guide Tube Uear Inspection and Circunferential question. The test, which is not described in the UFSAR, is Indexing of CEA Guide Tubes in System 80 Fuel Assebmblies," performed to examine surface wear on the interior surface of describes techniques for testing non-sleeved CEA Guide Tubes . the CEA guide tubes. This activity does not introduce any for inside surface wear and circunferential indexing of wear safety concerns beyond those previously analyzed in the indications. UFSAR.
II Page No. 06/27/89 Procedure 50.59 Report for 1988 Procedure Number Description Safety Evaluation-Suanary and Revision ST 16275-002, This procedure, ST NO. 16273-002, Rev 0, aFuel Assembly This new procedure did not introduce an unreviewed safety Rev 0 Reconstitution," details the sequence of work necessary on a question. The test, which is not described in the UFSAR, is fuel assembly that requires a recage due to one or more. performed to sequence the work required to remove defective broken rods or other unrepairable defects. fuel rods from an assecbly and replacing them with substitutes. This activity does not introduce any safety concerns beyond those already analyzed in the UFSAR. STD-NSS-070, This procedure, STD-NSS-070, Rev 4, "Procedure for Removal, This procedure revision did not introduce an unreyiewed Rev 4 Transfering, and Insertion of Fuel Rods in 16x16 Fuel safety question. The test, which is not described in the Assemblies,>> provides the information and instructions UFSAR, is performed to remove and replace defective fuel necessary to remove, transfer, and reinsert fuel rods, ckswy rods in fuel assemblies. This activity does not introduce spacer rods, and stainless steel replacement rods in fuel any safety concerns beyond those previously analyzed in the assemblies. UFSAR ~ XH-HF-P71,099, This new procedure, XN-NF-P71,099, Rev 0, "ULTRATEST This new procedure did not introduce an unreviewed safety Rev 0 System-USIP-12," provides information and instructions for question. The test, which is not described in the UFSAR, is installation, adjustment, and operation of test equipment performed to ultrasonically examine the individual fuel rods for ultrasonic testing of fuel assemblies. of spent fuel assemblies for fai lures. This activity does not introduce any safety concerns beyond those previously analyzed in the UFSAR. XH-NF-P71,098, This new procedure, XH-HF-P71,098, Rev 0, aULRTATEST This new procedure did not introduce an unreviewed safety Rev 0 Computerized Data Acquistion System-COHpO 286,a provides question. The test, which is not described in the UFSAR, is information and instructions for installation, adjustment, performed to ultrasonically examine the individual fuel rods and operation of a computerized data acquisition system for of spent fuel assemblies for failures. This activity does use with the ULTRATEST failed fuel detection system that not introduce any safety concerns beyond those previously performs ultrasonic testing of fuel assemblies. analyzed in the UFSAR.
06/27/89 Procedure 50.59 Report for 1988 Procedure Number Description Safety Evaluation Swrnary and Revision STD-HSS-117, This new procedure, STD-HSS-117, Rev 2,"Procedure for Eddy This procedure revision did not introduce an unreviewed Rev 2 Current Defect Examination of 1Cx1C or 16x16 Single Fuel safety question. The test, which is not described in the Rods," describes the technique for eddy current defect UFSAR, is performed to identify, by eddy current methods, testing of individual fuel rods from either 1Cx1C or 16x16 failed fuel rods in spent fuel assemblies. This activity configuration fuel assemblies. does not introduce any safety concerns beyond those previously analyzed in the UFSAR.
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Page No. 06/27/89 Safety Analysis Report Change Notice 50.59 Report For 1988 SAR Change No. Description Safety Evaluation Surmary 3002 Reported as Plant Change PCP 85-13-ES-004-00 (Design Change DCP 1/2/3 OJ-ES-007) in this report. 3008 Reported as Item 160 in the 1987 Report. 3009 Reported as Site Hodification 1-SH-RC-004 in this Report. 3012 This change to Section 3.11 and Appendix 3E of the Updated This change did not introduce an unreviewed safety question. Final Safety Analysis Report (UFSAR) provides post-LOCA The inclusion of this information in the UFSAR reports dose integrated dose levels utilized in equipment qualification levels that were used in enviroreental qualification calculations for rooms containing Low Pressure Safety calculations previously reported to the NRC in the Injection recirculation fluids and for the 70 foot level of Environmental Qualification Report. It has no effect upon the west penetration room containing Post Accident Sampling any other information reported in the UFSAR. Systen (PASS) valves. 3022 This change to Section 1.8 of the UFSAR provides an This change did not introduce an unreviewed safety question. additional exclusion statement to the comnitment for The addition of the exemption statement provides further Regulatory Guide 1.143, Rev 0. This exclusion statement definition of the boundary of the radwaste system to include further defines the boundary of the radwaste system information in Revision 1 of Regulatory Guide 1.143. This addressed by Regulatory Guide 1 '43. change to Regulatory Guide 1.143 has been accepted by the NRC staff, and the ACRS had concurred, during the NRC review and approval of Revision 1. 3023 This change to Appendix 98.2 of the UFSAR revises This change did.not introduce an unreviewed safety question. requirements for fire door installation, to allow the The fire door installation deviations from the originally installation to deviate from the originally tested tested configuration have been found acceptable based upon configuration, that formed the basis for the original additional testing. These deviations are also in accordance testing. with the requirements of Generic Letter 86-10 to ensure that fire barriers continue to meet the level of protection originally required.
'A 3025 This change to Sections 1.8, 9.3, and 11.3 of the UFSAR This change did not introduce an unreviewed safety question.
deletes the requirement to sequentially analyze the Chemical The requirements to individually monitor the CVCS tanks is and Voiune Control System (CVCS) tanks and the waste gas satisfied by continuously monitoring the header to the Surge
Ij 06/27/89 0 Safety Analysis Report Change Hotice 50.59 Report For 1988 SAR Change Ho. Description Safety Evaluation Smeary decay tank for hydrogen and oxygen. The change also deletes Tank. Honitoring the hydrogen concentratrions in the waste requirements to analyze the waste gas holdup system for gas holdup system is no longer required as the hydrogen hydrogen. concentration will al~ays be asswed to be greater than four volune percents; therefore, the hydrogen monitors have been removed from operation. By always assuaing that the waste gas holdup system has a hydrogen concentration greater that 4X by volune, the margin of safety is increased as actions to reduce the oxygen concentration will be initiated upon reaching two volune percent oxygen without regard to the measured hydrogen concentration. This change was incorporated into the PVHGS Technical Specification as Amendaents 36, 23, and 12 for Units 1, 2, and 3 respectively. 3026 This change to section 9.2 of the UFSAR incorporated all the This change did not introduce an unreviewed safety question. seismic class 1 valves that are installed in 1 1/2 inch and During a safety system fmctionaL inspection, the HRC larger lines and are in the water flow path of the essential identified that not all in-line valve installed in 1 1/2 chilled water system. inch and larger lines of the water side of the essential chilled water system were included in the Listing in UFSAR Table 9.2-29. These valves were previously included in UFSAR Figure 9.2-10 and analyzed in accordance with applicable Seismic Design Criteria. This change incorporates those valves not previously included in Table 9.2-10 and introduces no other notifications to the UFSAR. 3028 This change to Table 9.3-3 of the UFSAR incorporated This change did not introduce an unreviewed safety question. dissolved hydrogen and total gas into the listing for The change incorporates information that was inadvertently reactor coolant letdown sample analysis. left out of the table during a previous revision. 3033 This change to section 9.3 of the UFSAR reflects more This change did not introduce an unreviewed safety question. accurately the testing requirements for determining The change provides consistency among documents which operability of the Post Accident Sampling System (PASS) to contain requirements and descriptions of the PASS. specify which components require testing.
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06/27/89 Safety Analysis Report Change Notice 50.59 Report For 1988 SAR Change Ho. Description Safety Evaluation Summary 3034 This change to Table 9.2-28 of the UFSAR is an update to This change did not introduce an unreviewed safety question. reflect the as-built configuration of the Essential Chilled The change reflects final design calculations and installed Mater System. conditions. 3036 This change to section 1.9 of the UFSAR deletes certain This change did not 'introduce an unreviewed safety question. inspection requirements for the Unit 2 steam generator The deletion of the inspection requirements for Unit 2.was feedwater sparger. based upon the satisfactory results of the feedwater water hanmer tests in Units 1 and 3 and is discussed in correspondence to the HRC (161-00963) dated April 22, 1988. 3038 This change to Table 1.8-1 provides a note to explain the This change did not introduce an unreviewed safety question. radiation monitor deviation to R.G. 1.97 accuracy The change reflects information provided to the HRC in requirements. correspondence (161-00667) dated November 20, 1987. 3040 This change to sections 1.9 and '15.1 of the UFSAR reports an This change did not introduce an unreviewed safety question. increase in maximIa feedwater flow beyond that previously The PVHGS main feedwater system has been determined to be incorporated in the UFSAR by reference to CESSAR. capable of supplying 125X of the nominal feedwater flow at 100X power (rather than the 110X previously reported in CESSAR) under the off-normal conditions analyzed in section 15.1.2. 3041 This change to sections 1.2 and 2.4 of the UFSAR This change did not introduce an unreviewed safety question. incorporates the additional (second) evaporation pond The site of the additional evaporation pond was previously recently coapleted. shown as a wfuture" evaporation pond on UFSAR figures. 3042 This change to Figure 5.4-1 of the UFSAR incorporates a ~ This change did not introduce an unreviewed safety question. change in the pressurizer High Pressure Alarm from 2350 to The change is conservative and resulted from the thermal 2285 psia. improvement program initiated by the Unit 1, Cycle 2 reload. 3043 The change to section 1.9 of the UFSAR takes an exception to This change did not introduce an unreviewed safety question. CESSAR regarding the presense at an SRO when fuel is being This change is incorporated to achieve consistency between handled. An exception to CESSAR is also taken with regard to the PVNGS Technical Specifications and the UFSAR. when inventory should be taken to ensure the tag board is correct.
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Page No. 06/27/89 Safety Analysis Report Change Notice 50.59 Report For 1988 SAR Change No. Description Safety Evaluation Sureary 3044 The change to Tables 8.3-1 and 8.3-3 of the UFSAR revises This change did not introduce an unreviewed safety question. class IE loading for two valve motor operators. This change reflects revised class IE loadings resulting from upgrading the motor operators on two safety injection valves. 3045 These changes to Table 1.8-1 of the UFSAR provide additional This change did not introduce an unreviewed safety question. information concerning the existing design of the plant with These changes,to Regulatory Guide 1.97 compliance reflect regard to Regulatory Guide 1.97. Exception is taken changes in response to NRC inspection findings docunented in concerning steam generator wide range level indication. Inspection Report 50-528/88-12 and provided additional information concerning the existing design of the plant. 3050 These changes to section 17.2 of the UFSAR provide This ch'ange did not introduce an unreviewed safety question. administrative changes imposed as a result of the past These changes reflect responsibility changes for quality reorganization. assurance activities resulting from a division reorganization. 3052 These changes to section 8.3 and Table 1.8-1 of the UFSAR This change did not introduce an unreviewed safety question. reflect a change to the power source of the contairment sump . The change reflects a realiganent of the power source for level transmiters. the containment sump level receivers from non-class to class IE. 3053 These changes to section 13.4 of the UFSAR reflect This change did not introduce an unreviewed safety question. organizational restructuring with respect to the Nuclear The change relects changes to the Nuclear Safety Department Safety and Licensing Department. resulting from a division reorganization. 3055 These changes to Chapter 'i3 of the UFSAR reflects This change did not introduce an unreviewed safety question. organizational restructuring with respect to PVNGS. The changes reflect realiganent of the PVNGS organizations resulting from a project reorganization. 3056 These changes to Chapters 6, 9, 10, and 11 of the UFSAR are This change did not introduce an unreviewed safety question. for consistancy with design docunents. These changes are for editorial and consistency purposes to achieve uniformity with design documents.
I ~ 06/27/89 Safety Analysis Report Change Notice 50.59 Report For 1988 'AR Change No. Description Safety Evaluation Swmary 3058 These changes to Table 1.8-1 of the UFSAR correct This change did not introduce an unreviewed safety question. typographical errors and clarify information. These changes correct typographical errors and add clarification to reflect the design docunents related to Regulatory Guide 1.97 carpiiance. 3059 These changes to section 17.2 of the UFSAR reflect the QA/QC This change did not introduce an unreviewed safety question. organizational restructuring. These changes reflect responsibility changes in the Quality Assurance/Quality Control organization and program. 3060 These changes correct typographical errors and change Table This change did not introduce an unreviewed safety question. 7.1-1 of the UFSAR to reflect the as-built configuration. The change corrects a typographical error and changes entries in the table to reflect the existing plant configuration. 3061 These changes to sections 6.2 and 6.3 of the UFSAR This change did not introduce an unreviewed safety question. incorporates reload anaysis report information for large These changes consist of editorial corrections as well as break LOCA, small break LOCA, and the post Cycle 1 information from the Unit 1 and 2 Reload Analysis Report evaluation for Units 1 and 2. that was previously submitted to the NRC. 3062 This change to chapter 7 of the UFSAR updates the current This change did not introduce an unreviewed safety question. Safety Analysis Report in support of the reloads for Units 1 Information, that was previously referenced to CESSAR, was and 2 incorporated *into the UFSAR in this change. Editorial corrections were incorporated as well as information'from the Unit 1 and 2 Reload Analysis Report that was only submitted to the NRC. 3063 The changes to sections 4.2.3 and 4.4 of the UFSAR are being This change did not introduce an unreviewed safety question. made in support of the reloads for Units 1 and 2. The change information, that was previously referenced to CESSAR, was incorporated into the UFSAR in this change. Editorial corrections were incorporated as well as information from the Unit 1 and 2 Reload Analysis Report that was previously submitted to the NRC.
l' 06/27/89 Safety Analysis Report Change Notice 50.59 Report For 1988 SAR Change No. Description Safety Evaluation Stannary 3064 This change to chapter 15 of the UFSAR incorporates tests This change did not introduce an unreviewed safety question. from Units 1 and 2 Reload Analysis Reports. The changes incorporate information from the Unit 1 and Unit 2 Reload Analysis Reports previously submitted to the NRC. 3065 This change to section 4.3 of the UFSAR incorporates CESSAR This change did not introduce an unreviewed safety question. 4.3 into the UFSAR for Units 1 and 2 and revises section 4.3 This change incorporates information previously included in to reflect Cycle 2 information for Unit 2. the CESSAR into the UFSAR and reflects Cycle 2 information for Unit 2 in the UFSAR ~ 3066 This change to Appendix 98 of the UFSAR deletes requirements This change did not introduce an unreviewed safety question. for a 3-hour rated fire barrier between the fuel, building No wall exists in the pipe tunnel and no design for a wall and the associated pipe tumed is provided. A wall is not required due to a lack of combustibles and ignition sources in the tunnel, as well as the distance from the nearest safe shutdown circuitry or equipnent. 3067 This change to section 5.2 of the UFSAR reflects application ~ This change did not introduce an unreviewed safety question. of Code Case N-247-1 to RCP Hex Nuts. This change docunents the application of a code case that was previously approved for use at PVNGS. 3068 This change to section 9.5, Appendix 9A, and Appendix 9$ of This change did not introduce an unreviewed safety question. the UFSAR incorporates editorial corrections and a revision The changes reflect editorial corrections and revised of combustible loading in the Fire Protection Evaluation combustible loading calculations. Report. 3069 This change to section 6.3 of the UFSAR revises the CESSAR This change did not introduce an unreviewed safety question. interface evaluation safety inspection task K-Factor range The ECCS small and large break LOCA was evaluated to expand provided in the ECSS small and large break LOCA analysis. the K-Factor range. The existing analysis remained valid and the ECCS performance criteria of 10 CFR 50.46 is satisflede
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Page . 7 06/27/89 Safety Analysis Report Change Notice 50.59 Report For 1988 SAR Change No. Description Safety Evaluation Senary 3071 This change to section 6.5 and Appendix 38 of the UFSAR This change did not introduce an unrevieMed safety question. includes the addition of two piping analysis computer These coaputer programs are used in the performance of programs not previously identified in the UFSAR. snubber reduction and eater hanner analyses and have been bench marked against other standard computer programs used in the industry. 3072 This change to section 12.3 of the UFSAR incorporates This change did not introduce an aereviesed safety question. additional shielding calculational computer programs used at This change incorporates additional coaster programs used PVNGS. at PVNGS to reflect current plant activities and does not affect plant systems or components.
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