ML20236B364
| ML20236B364 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/31/1988 |
| From: | DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20236B363 | List: |
| References | |
| NUDOCS 8903210053 | |
| Download: ML20236B364 (100) | |
Text
{{#Wiki_filter:-_ _ - _ _ _ . _ _ - _ _ _ _ - _ _ _ __ _ _____ _ _ _ _ . _ . l' Encl;::.ura to l NRC-89-0044 l Page 1 I I ' i t li l l 1 l 1 l
*i i
FERNI 2 SAFETY EVALUATION REPORT - 1988 i I i i i March 1989 l
/
8903210053 890317 - I PDR ADOCK 05000341 l-K PDC 9
Enclo:ura to i l NRC-89-0044 Page 2 SECTION 1 DESIGN CHANGES
Ecclosure to NRC-89-0044 Page 3 ENGINEERING DESIGN PACKAGE (EDP) DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 86-0113 UFSAR Text Change Figure Change Implementation Document No.: EDP 5175 System No.: D1100, T4100 Title of Change: Deletion of Radwaste and Turbine building vent exhaust radiation monitors Summary: This change deletes the Radwaste Building and Turbine Building Vent Exhaust Radiation Monitors from the Control Center Air Conditioning System (CCACS) Isolation Trip Logic. The work scope includes disconnection and sparing in-place of Auxiliary Relays, T41M214 and H217'from HVAC Panel H21-P296A, and T41M215 and H219 fron panel H21.P2968. Safety Evaluation Summary: This Engineering Design Package (EDP) was implemented in accordance with Licensing Amendment No. 7 to Facility Operating License No. NPF-43 for Fermi 2. In the amendment review it was determined that signals from the control room outside air radiation monitors, as well as the reactor protection signals, are sufficient to initiate the control room HVAC emergency ventilation mode, such that the dose guidelines are ulet with c.espect to all design basis airborne radioactivity release accidents, including the LOCL. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 86-0131 UFSAR Text Change Implementation Doc,u ment No.: EDP 1720 Systen No.: P4400 Title of Change: Isolation of Emergency Equipment Cooling Water makeup tanks.
< l 3
Enclorura to l NRC-89-0044
'Page 4 ]
Summary: The intent of this change was to prevent flooding of the Emergency ! i Equipment Cooling Water. (EECW) Make-up Tank during the Stand-by mode of the EECW systen in conjunction with normal operation of the RBCCW { j system. Motor operators were installed on the Make-up Tank outlet isolation valves to prevent by-pass leakage from the RBCCW system. This valva will open on an EECW system actuation signal. Safety Evaluation S = mary: The modification to the EECW system does not degrade or prevent the j EECW system from being operated locally or affect auto-initiation. In j addition, since a full capacity, redundant EECW loop provides j shut-down capability should failure of one of the new HOVs occur, no j unreviewed safety questions exist. l DCSIGN CHANGE
SUMMARY
Safety Evaluation No.: 86-0238 UFSAR Figure Change
' Implementation Document No.: EDP 1786 System No.: T5000 Title of Change: Primary Containment' Monitoring System radiation monitor change.
Summary: f j The Primary Containment Monitoring System (PCMS) radiation monitor ; particulate detection channel was converted to installed spare status j j in order to conform to the UFSAR. i Summary of Evaluation: The Primary Containment Monitoring System (PCNS) radiation moni. tor is not safety-related. With the particulate channel no longer energized, a fixed filter, replacing the moving filter is used to collect particulate to support proper operation of the remaining noble gas monitor. Converting the PCMS particulate channel to an installed spare was completed to agree with the UFSAR.
Enclosura'to-
.NRC-89-0044-Page 5 DESIGN CHANGE SUMARY -
Safety Evaluation No.: 86-0255 UFSAR Text ~ Change i Figure Change
~
Implementation Document No.: EDP 6671 i l System No.~: G33 ! Title.of Qiange: Replacement of the Reactor Water Cleanup system holdup line. , 1 Summary: To prevent flashing at the Reactor Water Cleanup.(RWCU) pump inlet, I the twenty four-inch (24") holdup line was reduced to a six inch (6") j suctior. pipe. - l Safety Evaluation S = mary: The.RWCU system was. unable to support plant operations during normal shutdown depressurization due to flashing at the. inlet of the RWCU pumps. The flashing resulted in pump trips. A. hydraulic analysis (Design Calculation No.:4457) of the RWCU pump suction showed that
. system response during normal shutdown depressurization could be improved by' replacing the twenty four inch (24") delay pipe with six inch (6") pipe and providing a true high point vent.
a DESIGN CHANGE SUMARY Safety Evaluation No.: 86 0250 UFSAR Figure Change i Implementation Document No.: EDP 6736 System No.: C1100 Title of Change: Relocated Control Rod Drive Minimum Flow Line. Summary: Relocated one inch (1") Control Rod Drive Minimum Flow line to tie , I into the Condensate Return Tank header downstream of valves E4100F153 and E4100F178. Sunscry of Evaluation: i This modification eliminated the possibility of overpressurizing the portion of the condensate return line piping upstream of E4100F158 l I
4 EncloIura.to. NRC-89-0044 Page 6 i and E4100F178 due to inadvertent operation of the Control Rod Drive (CRD) pump (s) with these valves closed. The relocated pipe was
~ analyzed.and_ supported to appropelate seismic criteria, the penetration was resealed to original criteria, and the CRD pump minimum flow capacity was not adversely affected.
DESIGN CHANGE SLRNIARY Safety Evaluation No.: 87 0045 , UFSAR Figure Change Implementation Document No.: EDP 6684 System No.: B3100 Title of Change: Reactor Recirculation Pump A sechanical seal piping. Samary: Change document-issued to add unions to Reactor Recirculation Pump A Mechanical Seal Piping to facilitate seal removal and correct system diagrams to reflect actual piping size, class, and material. Safety Evaluation Sammary: Neither the physical addition of unions to existing piping nor updating system drawings to more correctly reflect as-built conditions result in any unreviewed safety concerns. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0056 UFSAR Figure Change Implementation Document.No.: EDP 6642 System No.: P9500 i Title of Change: North and South Extraction steam instrumentation changes. Snmuary: Level instrumentation for the number 3 North and South Extraction steam drain pots was replaced or modified to prevent water level from flooding the extraction steam line.
Enclosur3 to NRC-89-0044 Page 7 1 Summary of Evaluation: The drawing change made was for clarity only. No physical changes in the plant were made that affected the UFSAR figure. The drawing j change was made to more accurately reflect the plant design as d licensed. DESIGN CHANGE SUletARY i Safety Evaluation No.: 87 0072 UFSAR Figure Change Implementation Document No.: EDP 5800 ' l System No.: B3100 Title of Change: Reactor Recirculation pump modifications. Summary: The Reactor Recirculation Pumps (RRP) permissive logic was modified f such that the permissive signal is actuated only for a period of 20 seconds after the control switch of the RRP is placed in the run ! position. This modification prevents automatic restart after an Anticipated Transient Without Scram (ATWS) trip event. Safety Evaluation Summary: The sequence of events for Reactor Recirculation Pump (RRP) shutdown or startup remain unaffected and therefore the safety function remains unaffected. Should an Anticipated Transient Without Scram (ATWS) e'/ent occur, the generator field breaker will open. This action de-energizes the Recirculation pump motor even though tb? drive motor ! continues to run. This modification prevented the generator field ! breaker from reclosing until the drive motor CMC switch has been cycled through the "0FF" position. l
=
Enclclurs to NRC-89-0044 i Page 8 DESIGN CHANGE SUlelARY
-Safety Evaluation No.: 87-0128 UFSAR Figure Change Implementation Document No.: EDP 7370 System'No.: N2100 Title of Change: Reactor Feed Pump Seal Water Return Pump modifications.
2= mary: The Reactor Feed Pump Seal Water Return Pump logic / control circuit was modified to reduce excessive cycling of the pump due to a small level range for on/off operation. Summary of Evaluation: The change was made to prevent premature failure of the Reactor Feed , f Pump (PEP) Seal Water Return Pump due to excessive cycling operation; therefore, the probability of equipment malfunction is decreased. i Additionally, operation of the RPF Seal Water Return Pump is not part 1 of a safety system previously evaluated in the UFSAR., l l DESIGN CHANGE
SUMMARY
3 Safety Evaluation No.: 87-0133 UFSAR Text Change Figure Change Implementation Document No.: EDP 7308 System No. G1:16 Title of Change: Extruder / evaporator modifications. Sammary: This change was issued to provide for several changes to the Solid Radwaste System Extruder / Evaporator G116D087. The changes are being made to correct maintenance reliability and stability problems ; encountered during the startup and operation of identical units. (Palisades and Hope Creek). Safety Evaluation Summary: , This system is not a safety related system nor does it interface with safety-related systems in any way. This safety evaluation was l l y .
- l
Enclorura to NRC-89-0044 l Page 9 performed due to a revision of a system diagram which resulted in an UFSAR change. DESIGN CHANGE SUlWARY Safety Evaluation No.: 87-0140 UFSAR Figure Change Implementation Document No.: EDP 6949 j Systen No.: G3300 l I Title of Change: Reactor Water Cleanup check valve addition. l Rn==ary: Addition of a second check valve in series with the existing check valve (G3300F059) in the air supply line to the Reactor Water Cleanup l system filter /demineralizers. Safety Evaluation Summary: The addition of a second check valve in series with the existing check valve in the air supply lines to the Reactor Water Cleanup filter /demineralizers constitutes a system enhancement, which further reduces.the potential for contamination of tne air system due to backflow from the filter /demineralizers. No unreviewed safety concerns exist. DESIGN CHANGE
SUMMARY
{ Safety Evaluation No.: 87 0150 UFSAR Figure Change Implementation Document No.: EDP 2873 Systen No.: E1100, P4100 l Title of Change: Sample point modifications. Rnamary: This modification added two (2) new " grab" type sample points and sinks to the Division I and II Residual Heat Removal Service Water (RHRSW) system, and also relocated the General Service Water (GSW) local grab sample point to an existing sample sink.
Enclo2urs to NRC_89 0044 Page 10 l, Safety Evaluation humar'y: I Incorporation'of grab sample points into the General Service Water and-Residual Heat Removal Service Water systems are considered plant enhancements which facilitate the ability of' plant personnel to.take samples in a timely, efficient, clean manner which improves the design- ., with~ regard to ALARA principles. l 1 DESIGN CHANGE SUtStARY I Safety Evaluation No.: 87-0161 UFSAR Figure Change
' Implementation Document No.: EDP 6424 System No.: A3000 Title of Change: Metrology laboratory. i Summary:
This. engineering design' change added a Metrology Laboratory to the Office Service Building Warehouse.
. Safety Evaluation Summary:
General arrangement drawing revision of the Office Service Building Warehouse to show Metrology Laboratory. The laboratory will be i located ~approximately 290 feet away from the nearest safety-related equipment. It will have no influence on the function or operation of any equipment-located in the plant. It will be used to house tools and equipment used to calibrate instruments. l DESIGN CHANGE
SUMMARY
o Safety Evaluation No.: 87-0174 UFSAR Figure Change Implementation Document No.: EDP 6468 Systes No.: G3300 Title of Change: Reactor Water Cleanup leak detection isolation logic modification. l
Enclo'ura to NRC-89-0044' Page 11 Summary: 1 This modification added two keylock selector switches (Division I and II) in the Control Room to bypass their respective temperature channels of Reactor Water Cleanup leak detection isolation logic to allow the surveillance / maintenance of temperature switches without-isolating'the system. Safety Evaluation h==a y: The switches added in this change do not have a history of frequent failures and were procured as qualified components. If a switch were to fail in the closed position, an annunciator will provide warning to the Control Room Operator. However, the other division will still be available to achieve system isolation if necessary. During surveillance testing or maintenance, failure of a switch in the open position would cause isolation of the Reactor Water Cleanup System. However, the probability of a switch failure in the open position during performance of surveillance / maintenance is very low and this isolation will be infrequent. This change overall will be an ; improvement over the present-design. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0193 UFSAR Figure Change Implementation Document No.: EDP 7684 System No.: G3300 Title of Change: Modifications to the opening circuitry of Reactor Water Clean-up system valves. Suasary: The opening operation of valve operators for G3352F100, G3352F101, G3352F102, G3352F106 and G3352F119 were revised to prevent torque switch backseating in the opening mode. Safety Evaluation Summary: Powell Globe and Gate valves were not evaluated for backseating using the opening torque switch to de-energize the motor starter. Because they were not evaluated for the effects of backseating with the opening torque switch, the opening control circuits for valves G3352F100, G3352F101, G3352F102, G3352F106 and 03352F119 were revised so that when the valves reached the full open position, power to the valve motor operator was de-energized by the limit switch opening.
Enclosure to NRC-89-0044 Page 12
' This ensured proper valve operation (s) when plant conditions required them.
DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0199 UFSAR Figure Change Implementation Document No.: EDP 7177 System No.: E4100 Title of Change: Modifications to the Opening Circuitry of High Pressure Coolant Injection system valves. Snamary: The opening operation of the valve operators for E4150F008 and E4150F011 were revised to prevent torque switch backseating in the opening mode. Summary of Evaluation: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de energizing the motor operator. Because of this, the opening circuit for High Pressure Coolant Injection valves E4150F008 and E4150F011 were revised so that when the valve reaches full open position, power to the valve motor is de-energized by the opening limit switch. This ensures proper valve operation (s) when plant conditions require. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87 0203 UFSAR Figure Change Implementation Document No.: EDP 7620 Systen No.: 01116 Title of Change: Radwaste asphalt system changes. Snamary: This engineering change completed the following: 1) replaced existing PCV G1116.F850 with a valve designed for this application and resets pressure at 170 psig. 2) Installed a new pressure switch for system high pressure alarm to be set at 180 psig. 3) Installed a local pressure gauge. 4) Replaced and reset pressure relief G116-F1050 to
Enclosure to ; NRC-89-0044 Page 13 210 psig and 5) replaced damaged, irreparable G116-F1101 (VR3-3439) with a new valve. Safety Evaluation h ==ary: 1 The principle safety concerns associated with the radwaste system are the potential for uncontrolled releases of radioactive fluids and 'l excessive exposure to operations and maintenance personnel (ALARA). The changes made by this change do not affect the previous design and j safety evaluation with regard to uncontrolled releases or ALARA i considerations. The changes were made to improve system operation for both the Asphalt System and the Asphalt System Reduced Pressure Steam System which are clean systems that would not become contaminated or cause other systems to become contaminated'during normal operations. These are auxiliary systems for non-safety _related Radwaste System 3 contained in the non-safety-related Radwaste Building and are l separated from all safety-related equipment. j DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0204 UFSAR Figure Change Implementation Document No.: EDP 7681 - I System No.: E2100 Title of Change: Modifications to the opening circuitry of Core l' Spray system valves. Su==ary: The opening operations of valve operators for E2150F004A, E2150F005A, > E2150F031A and E2150F036A were revised to prevent torque switch backseating in the opening mode. Safety Evaluation Summary: Powell valves were not completely evaluated for the effect of backseating using the torque switch to de-onergize the motor operator. l The Core Spray System valves E15050F004A, E2150F005A, E2150F031A and j E2150F036A opening circuitry was revised by providing an open limit f switch in series with a torque switch to de-energize the valve motor ; l starter such that the valve stops within 1/4 to 1/2 turn before j backseat contact is made. This ensured proper valve operation (s) when plant conditions required them. f
1
'Enclo;ure to NRC-89-0044' Page 14 DESIGN CHANGE SmMARY l
-UFSAR Figure Change -l Safety Evaluation No.: 87-0205 Implementation Document No.: EDP 7682 System No.: E2100 Title of Change: Modifications to the opening circuitry of Core 1 Spray system valves.
h==acy: The opening operation of valve operators for E2150F004B, E2150F005B, E2150F031B and E2150F036B were revised to prevent torque switch - backseating in the opening mode. Safety Evaluation Summary: j Powell valves had not been completely evaluated for the effect of ! backseating using the torque switch to de-energize the motor operator. The Core Spray system valves E2150F004B, E2150F005B, E2150-F031B and E2150F036B opening circuitry was revised by providing an open limit switch in series with a torque switch to de-energize the valve motor i starter such that the valve stops within 1/4 to 1/2 turn before backseat contact is made. This ensured proper valve operation (s) when plant conditions required them. ! DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0206 UFSAR Figure Change Implementation Document No.: EDP 7674 1 Systen No.: E1150 Title of Change: Modifications to the opening circuitry of a Residual Heat Removal system valve Summary: The opening operation for the valve operator of E1150F009 was revised to prevent torque switch backseating in the opening mode. Safety Evaluation Summary: Powell gate valves had not been evaluated for the effect of . backseating using the torque switch to de-energize the motor
Enclosure to NRC-89_0044 Page 15 i operator. The Residual Heat Removal (RHR) system valve E1150-F009 i opening circuitry was revised such that when the valve reached the i full open position, the valve motor is stopped by the opening limit switch to within 1/4 to 1/2 turn before backseat contact'is made. i Also, the RHR pumps relay logic "A" stop circuit was modified to make available the use of limit switch contact LS4 which is required for ' i incorporation of the modification. This ensured proper valva operation (s) when plant conditions required them. i l DESIGN CHANGE SUMARY Safety Evaluation No.: 87-0208 Implementation Document No.: EDP 7679 System No.: E1150 Title of Change: Modifications to the opening circuitry of Residual , Heat Removal system valves. l Summary: The opening operations of the valve openators for E1150F603A, E1150F604A and E1150F605A were revised to prevent torque switch backseating in the opening mode. Safety Evaluation Snamary: Powell valves had not been completely evaluated for effects of backseating with the opening torque switch de-energizing the motor operator. Because of this, the opening circuits for Residusl Heat Removal valves E1150F603A, E1150F604A and E1150F605A were revised so that when the valve reaches the full open position, power to the valve motor is de-energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. i l 1 I 1 l
~
I
Enclo:ure to
- NRC-89-0044 Page 16.
DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0209 Implementation Document No.: EDP 7676 System No.: E1150 Title of Change: Modifications to the opening circuit of a Residual Heat Removal system valve. Summary: The opening operation for the va..a operator of E1150F608 was revised to prevent torque switch backseating in the opening modo. f i Safety Evaluat?.on Suasary: Powell valves had not been completely evaluated for the effects of backseating with the torque switch motor cutout. Because of this, the opening circuit for the Reactor Recirculation Extractor Isolation to l RHR Bypass valve E1150F608 was revised so that when the valve reaches the full open position, power to the valve motor is de-energized by ; the opening limit switch. (E1150F608 is also known as the RHR Shutdown Cooling Inboard Inlet Isolation Bypass valve.) This ensured i proper valve operation (s) when plant conditions required them. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0210 Implementation Document No.: EDP 7680 System No.: E1150 Title of Change: Modifications to the opening circuitry of Residual Heat Removal system valves. l Sn==ary:
, The opening operations of the valve operators for E1150F603B, E1150F04B and E1150F605B were revised to prevent torque switch backseating in the opening mode.
l l f
Enclosuro to NRC_89 0044 Page 17 Safety Evaluation Summary: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de. energizing the motor operator. Because of this, the opening circuit for Residual Heat Removal valves E1150F603B, E1150F604B and E1150F605B were revised so that when the valve reaches the full open position, power to the valve motor is de-energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. DESIGN CHANGE SUMARY Safety Evaluation No.: 87-0212 UFSAR Figure Change Implementation Document No.: EDP 7678 System No.: E1150 Title of Change: Modifications to the opening circuitry of Residual Heat Removal system valves. Summary: The opening operations of the valve operators for E1150F010, E1150F015A, E1150F015B, E1150F017A and E1150F017B were revised to prevent torque switch backseating in the opening mode. Safety Evaluation Summary: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de-energizing the motor operator. Because of this, the opening circuits for Residual Heat Removal valves E1150F010, E1150F015A, E1150F015B, E1150F017A and E1150F017B were revised so that when the valve reaches the full open position, power to the valve motor is de-energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. l
~
1 Enclo;ure to NRC-89-0044 Page 18 j DESIGN CHANGE
SUMMARY
j Safety Evaluation No.: 87-0213 1 Implementation Document No.: EDP 7688 l 1 Systen No.: P4400 ; Title of Change: Modifications to the opening circuitry of Emergency Equipment Cooling Water valves. . hamary: The opening operations of the valve operators for P4400F601A and P4400F603A were revised to prevent torque switch backseating in the l opening mode. i Safety Evaluation Snamary: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de-energizing the motor operator. Because of this, the opening circuits for Emergency ! Equipment Cooling Water-valves P4400F601A and P4400F603A were revised so that when the valve reaches the full open position, power to the . valve motor is de-energized by the opening limit switch. This ensured i proper valve operation (s) when plant conditions required them. l DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0214 j g Implementation Document No.: EDP 7689 Systen No.: P4400 Title of Change: Modifications to the opening circuitry of Emergency Equipment Cooling Water system valves. Summary: The opening operations of the valve operators for P4400F601B and P4400F603B were revised to prevent torque switch backseating in the opening mode. l l l
Enclosura to NRC-89-0044 Page 19 Safety Evaluation %==acy: Powell valves had not been completely evaluated for the effects of backseating with the opening torque de-energizing the actor operator. Because of:this, the opening' circuits for Energency, Equipment Cooling Water valves P4400F601B and P4400F603B were revised so that when the valve reaches the full open position, power-to the valve motor is de-energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. , DESIGN CHANGE SUIMARY Safety Evaluation No.: 87-0220 UFSAR Figure Change Implementation Document No.: EDP 7675 l System No.: E1100 Title of Change: Modifications to the opening circuitry of Residual Heat Removal system valves. Snamary: 1 The opening operations of valve operators for E1150F003A,'4A, 4C, 6A, ' 6C, 7A, 16A, 21A, 27A, 28A, 47A and 48A were revised to~ prevent torque switch backseating in the opening mode. j Safety Evaluation Sumancy: Powell globe and gate valves were not completely evaluated for the effect of backseating using the torque switch to de-energize the motor operator. The residual heat removal system valves E1150F003A, 4A, 40, 6A, 6C, 7A, 16A,'21A, 27A, 28A, 47A and 48A opening circuitry was revised so when the valves reach the full open position, the valves are stopped by opening limit switches within 1/4 to 1/2 turn before backseat contact is made. This ensured proper valve operation (s) when plant conditions required them, l
~
_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ - _ - _ _ _ n-
[ Enclosura to NRC_89-0044 Page'20 DESIGN CHANGE SU19tARY Safety Evaluation No.: 87-0225 UFSAR Figure Change l Implementation Document No.: EDP 7677 1 System No.: E1100 Title of Change: Modifications to the opening circuitry of Residual Heat Removal valves. !
)
h= mary: ! The opening operations of valve operators for E1150F003B, 4B, 4D, 6B, 6D, 7B, 8, 16B, 21B, 26B, 278, 288, 47B, 48B, 73 and 75 were revised to prevent torque switch backseating in the opening mode, j Safety Evaluation Summary: Powell globe and gate valves were not completely evaluated for the effect of backseating using the torque switch to de-energize the motor operator. The Residual Heat Removal valves E1150F003B, 4B, 4D, 6B, 6D, 7B, 8, 16B, 21B, 26B, 27B, 28B, 47B, 48B, 73 and 75 opening circuitry was revised so that when the valves reach the full open position, the valves are stopped by opening linit switches within 1/4 to 1/2 turn before backseat contact is made. This' ensured proper valve operation (s) when plant conditions required them. DESIGN CHANGE SU19tARY Safety Evaluation No.: 87 0226 UFSAR Figure Change Implementation Document No.: EDP 7580 System No.: P4400 Title of Change: Emergency Equipment Cooling Water Heat Exchanger modifications. Summary: The Division I/II Emergency Equipment Cooling Water (EECW) Heat Exchanger (Hx) outlet water high/ low temperature alarms operated spuriously when the EECW Hx pumps were not running. Thermocouple located in a pipe leg filled with stagnant water caused the spurious alarms due to ambient temperature increases. This modification
Enclosura to NRC-89-0044 Page 21 defeated the alarm circuits when the respective pumps were not in operation. h==acy of Evaluation: This modification does not affect any safety functions of the Energency Equipment Cooling Water system. The high/ low temperature alarms are not required and are defeated only when the pump (s) are not in operation. DESIGN CHANGE SUMARY Safety Evaluation No.: 87-0233 UFSAR Figure Change Implementation Document No.: EDP 7790 Systen No.: E1150 Title of Change: Modifications to the opening circuitry of Residual Heat Removal system valves. Summary: The opening operations of the valve operators for E1150F022 and E1150F024A were revised to prevent torque switch backseating in the i opening mode. ! Safety Evaluation Summary: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de-energizing the motor operator. Because of this, the opening circuits for Residual Heat Removal valves E1150F022 and E1150F024A were revised so that when the valve reaches the full open position, power to the valve motor is de-energized by the opening limit' switch. This ensured proper valve operation (s) when plant conditions required them. l l l [ .
Enclo:ura to NRC_89-0044 Page 22 DESIGN CHANGE SUMARY Safety Evaluation No.: 87-0235 UFSAR Figure Change ,
- Implementation Document No.: EDP 7789 Systen No.: . E1150 Title of Change: Modifications to the opening circuitry of Residual Heat Removal system valves.
m==ary: The opening' operations of the valve operators for E1150F023 and E1150F024B were revised to prevent torque switch backseating in the opening mode. Safety Evaluation Sammary: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de-energizing the motor operator. Because of this, the' opening circuits for Residual Heat Removal valves E1150F023 and E11F024B were revised so that when the valve reaches the full open' position, power to the valve motor is , de_ energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0237 UFSAR Figure Change Implementation Document No.: EDP 7791 Systen No.: B2100 Title of Change: Modifications to the opening circuitry of the Main Steam drain inboard isolation valve. Sammary: The opening operation of the valve operator for B2103F016 was revised to prevent torque switch backseating in the opening mode. l l
~
Enclogure to NRC-89-0044 Page 23 Safety Evaluation Summary: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de-energizing the motor -operator. Because of this, the opening circuit for the Main Steam Drain Inboard Isolation valve B2103F016 was revised so that when the valve reaches the full open position, power to the valve actor. is de-energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. i DESIGN CHANGE 3129tARY Safety Evaluation No.: 87-0238 UFSAR Figure Change Implementation Document No.: EDP 7792 System No.: G3300 Title of Change: Modifications to the opening circuitry of Reactor Water Clean-up system valves. ; Summary: The opening operation of the valve operator for G3352F001 was revised to prevent torque switch backseating in the opening mode. Safety Evaluation Summary: Powell, valves had not been completely evaluated for the effects of backseating with the opening torque switch de-energizing the motor operator. Because of this, the opening circuit for the Reactor Water Clean-up valve G3352F001 was revised so that when the valve reaches the full open position, power'to the valve motor is de-energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. l 1
)
9 1
Enclosure to NRC-89-0044 Page 24 DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0240 UFSAR Figure Change Implementation Document No.: EDP 7846 System No.: G3300 Title of Change: Modifications to the opening circuitry of a Reactor
-Water Clean-up system valve.
Summary: The opening operation of the valve operator for G3352F004 was revised to prevent torque switch backseating in the opening mode. Safety Evaluation Summary: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de-energizing the motor operator. Because of this, the opening circuit for the Reactor Water Clean-up valve G3352F004 was revised so that when the valve reaches the full open position, power to the valve motor is de-energ hed by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0241 Implementation Document No.: EDP 7803 - Systen No.: GS100 Title of Change: Modifications to the opening circuitry of Torus Water Management System isolation valves. Summary: l The opening operations of the valve operators for G5100F600, G5100F602, and GS100F606 were revised to prevent torque switch backseating in the opening mode. Safety Evaluation Snamary: Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de-energizing the motor operator. Because of this, the opening circuits for Torus Water
Enclo;ura to NRC-89 0044 Page 25'
-Management isolation valves G5100F600, G5100F602, and G5100F606 were revised so that when the valve reaches the full open position, power to the valve notor is de-energized by the opening limit switch. This ensured proper valve operation (s) when plant.' conditions required them.
DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87 0242 UFSAR Figure Change I
- (
Implementation Document No.: EDP 7845 System No.: B2100 Title of Change: Modifications to the opening circuitry of a Main Steam drain system valve. S==ary: The opening operation of the valve operator for B2103F019 was revised j to prevent torque switch backseating in the opening mode. Safety Evaluation Snumary: Powell valves had not been completely evaluated-for t'he effects of backseating with the opening torque switch de-energizing the motor 9 operator. Because of this, the opening circuit for the Main Steam Drain outboard isolation valve B2103F019 was revised so.that when the valve reaches the full open position, power to the valve motor is de-energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. j
f
' Enclecura to.
i NRC-89 0044 Page'26- j i 1 i DESIGN CHANGE
SUMMARY
l Safety Evaluation No.: 87-0243 .) i Implementation Document No.: EDP 7847 System No.: G5100 , Title of Change: Modifications to the opening circuitry of Torus ! Water Management System isolation valves. mamary: The opening operations of valve operators for G5100F601, G5100F603, G5100F605 and G5100F607 were revised to prevent torque switch backseating in the opening mode. Safety Evaluation Summary: Powell valves had not been completely evaluated for the effects of backseating the the opening torque switch de_ energizing the motor operator. Because of this, the opening circuit for torus water management system isolation valves (G5100F601, G5100F603, G5100F605 and G5100F607) was revised so that when the valve reached full open , position, power to the valve motor is de-energized by the opening - limit switch. This ensured proper valve operation (s) when plant conditions required them. , l ______________________________________________________________________ ) DESIGN CHANGE
SUMMARY
q UFSAR Figure Change j Safety Evaluation No.: 87 0245 1 Implementation Document No.: EDP 7702 System No.: G1115 l l Title of Change: Concentrates feed tank ultrasonic level system i.
]
replacement. Summary: Due to design difficulties with the Radwaste system Concentrates feed tank and level sensor malfunctions, the system was replaced with an electronic d/p cell transmitter with remote pressure seal. The level system which was replaced utilized a sonic continuous type sensor / transmitter. f I
. -1
Enclotura to NRC-89-0044 Page 27 Safety Evaluation Snamary: The principal safety concerns associated with the radwaste system 'are the potential for uncontrolled releases of radioactive fluids and excessive exposure to operations and maintenance personnel (ALARA). The changes made do not affect the previous design and safety evaluation with regard to uncontrolled releases or ALARA i considerations. The changes were made to improve the level indication i system of the Concentrates Feed Tank. This system will improve i ' indication reliability and will provide an accurate reading. The concentrates system is part of Radwaste Building and is separated from
~
all safety-related equipment. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87 0249 UFSAR Text Change Figure Change Implementation Document No.: EDP 1720, Rev. C Systen No.: P4400 Title of Change: Isolation of Emergency Equipment Coo. ling Water Makeup Tanks - Summary: This design modification. took the motor operators off of the Emergency Equipment Cooling Water heat exchanger's inlet valves and installed
'them on the EECW makeup tanks outlet valves. Logic was added so that the makeup tank valves do not start to open until the division -
isolation valves have closed. Safety Evaluation Summary: The safety evaluation was written against the applicable UFSAR changes l which resulted from this modification. Neither the modifications nor the UFSAR changes degrade or prevent the EECW system from being operated locally or affect auto-initiation. In addition, since a full capacity, redundant EECW loop provides shutdown capability upon failure of one of the new motor operated valves (MOVs), no unreviewed , safety questions exist. l l l
. -1
i
' Enclosure to NRC-89-0044 l Page 28 l DESIGN CHANGE
SUMMARY
3 Safety Evaluation No.: 87 0266 UFSAR Figure Change. 1 Implementation Document No.: EDP 3419 Systen No.: G1100 Title of Change: Chemical feed modification to the Extruder Evaporator Auxiliary Boiler.
% ==a ry :
System circuitry was modified to allow manual operation of the Chemical Feed pump, Feedwater Make-up solenoid valve and Boiler Feed pumps when the Auxiliary Boiler is secured. Summary of Evaluation: The principal safety concerns associated with the Radwaste system are i the potential for uncontrolled releases of radioactive fluids and ' excessive exposure to operations and maintenance personnel (ALARA). The changes made by this modification did not affect the previous design and safety evaluation with regard to uncontrolled releases or ALARA considerations. The changes were made to improve operation of the Auxiliary Boiler and to facilitate addition of chemicals to the Auxiliary Boiler by Chemistry. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0275 UFSAR Figure Change { l ' Implementation Document No.: EDP 7906 f System No.: R1400 j Title of Change: Motor Control Center Modifications. Summary: ] Isolating contactors and associated cabling and controls at 480 volt circuit breakers 72C-3C and 72F-5C were installed. Additionally, AC undervoltage, lockout relays and DC voltage monitoring relays at 480 volt switchboard 72C were installed. . l l l l l
-l ,
Enclosura to ' NRC-89-0044 Page 29 Suasary of Evaluation:
'A review of the various accidents considered in the UFSAR was made 'l that consider the swing bus of the Low Pressure Coolant Injection !
(LPCI) system functional, and it was determined that for the various ! accidents, the swing bus and LPCI systea'would be available as ; previously assumed in the evaluations. 'Ihe modification made ensured correct transfer of the swing bus to either available AC. source for an accident which considers Loss of Offsite Power with various total or partial-losses of one of the divisions of DC as the' considered single failure. Also considered were the.various components associated with the normal and standby feeds to the swing bus and the affect of any single failure that could cause more than loss of the swing bus, such ,, as loss of other Energency Core Cooling System (ECCS) equipment. The results of this review indicated that the proposed change would not .{ increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR as the new design will ensure that the MCC 72CF functions as previously intended. DESIGN CHANGE SUMARY f 4 Safety Evaluation No.: 87 0286 UFSAR Figure Change { Implementation Document No.: EDP 7944 System No.: P4400 Title of Change: Core Spray piping modifications. Summary: The Reactor Building Closed Cooling Water (RBCCW)/ Emergency Equipment Cooling Water (EECW) system provides cooling to the Division I and Division II Core Spray pump motors for bearing cooling. The RBCCW/EECW piping was connected to the motors with 3/4" flex hoses. These flex hoses experienced some bulging and were replaced with 3/8" I stainless steel tubing. Safety Evaluation Snamary: Use of a stainless steel tubing instead of flex hosing will not result in reduction or loss of RBCCW/EECW cooling to the coolers or impact in any way the operation of RBCCW/EECW. The use of the tubing will provide greater cooling flow reliability than the hoses.
Enclorura to
.NRC_89 0044.
Page 30 3 1 l I DESIGN CHANGE SMEARY Safety Evaluation No.: 87-0346
' Implementation Document No.: EDP 6732 System No.: T2300 Title of Change: Primary Containment Suppression Chamber and Reactor Building Vacuum Breaker isolation valve rework. .
h==ary: The Primary Containment Suppression Chamber and Reactor Building Vacuum Breaker isolation valves limit switches were replaced with a like_for-like model. 1 Safety Evaluation Summary: l The limit switches for the Primary Containment Suppression Chamber and Reactor Building Vacuum Breaker isolation valves were replaced with like-for-like parts. The original parts in limit switches T24N406A & B and T23N408A & B were Namco Model No. EA170-31100 and were replaced ; by Namco Model No. EA180-31302. Replacement of like-for-like parts ensures proper valve operation (s). ;
.________________________.._____________.._____..______...___________. l DESIGN CHANGE
SUMMARY
l l Safety Evaluation No.: 87-0369 : Implementation Document No.: EDP 7683 System No.: G1100 Title of Change: Modifications to the opening circuitry of a Radwaste System Drywell Floor Drain isolation valve. Su==ary: 1 The opening operation of the valve operator for G1154F600 was revised to prevent torque switch backseating in the opening mode. Safety Evaluation Summary: } Powell valves had not been completely evaluated for the effects of backseating with the opening torque switch de_ energizing the motor
Enclosure to NRC-89-0044 Page 31 operator. Because of this, the opening circuit for the Radwaste System Drywell Floor Drain isolation valve G1154F600 was revised so that when the valve reaches the full open position, power to the valve motor is de. energized by the opening limit switch. This ensured proper valve operation (s) when plant conditions required them. DESIGN CHANGE SUlWARY Safety Evaluation No.: 87-0394 Implementation Document No.: EDP 8132 Systen No.: T4100 Title of Change: Control Center Air Conditioning system modifications. Sumu ry: Time delay added to annunciator circuits for the Control Room Pressure High/ Low alarms to reduce the probability of spurious alarms. Safety Evaluation Sumury: Two (2) Agastat time delay relays, one each in panels H21-P296A and B, were added to provide a thirty (30) second time delay in the annunciator circuits for alarm windows 8D49 and 17D55 The design modification to add time delay relays was completed to reduce spurious alarms received at those windows. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0010 UFSAR Text Change Figure Change Implementation Document No.: EDP 8043 System No.: T5000 Title of Change: Automatic isolation of the non-essential Primary Containment Radiation Monitoring system. Sumancy: l The purpose of the modification was to provide automatic isolation of the non-essential Primary Co.9.ainment Radiation Monitoring System (PCRMS) from the essential Primary Containment Atmosphere Monitoring
Enclosuro to NRC-89-0044
- Page 32 l
l System (PCAMS), therefore maintaining PCAMS as a closed loop system outside containment. This design change resulted in PCRMS having two (2) redundant and divisional automatic isolation valves on the inlet and outlet lines. The subject isolation valves consist of the existing Div. I (T50-F450 and T50-F451) and the addition of Div. II (T50-F455 and T50-F456). Each isolation valve " fails closed" assuring- 1 integrity of the extended containment. A Technical Specification amendment'was received _for this change. Safety Evaluation Summary: , The electrical actuator power for these valves will be derived from redundant portions of the RPS system. Two of these valves.(one: inlet and one outlet) will be air operated by the interruptible air system. In the previous design,, valves T50-F450 and F451 (Div. I) provided isolation of the PCRMS in the event of a LOCA. However, since this isolation was not redundant (Div. I only) a single failure of an j isolation valve during LOCA would have violated closed loop requirements. There'was a second barrier provided by manual isolation valves, however use of a remote manual isolation valve as a barrier in the case of a non-essential system is not an approved alternative. Therefore, the existing design of PCRMS was upgraded to provide redundancy by addition of Div. II isolation valves. Each division's automatic control logic will provide a diverse valve trip / closure signal resulting from "high drywell pressure" or " low reactor-vessel water level 2". Each valve will require manual operator. action to reopen, providing that the logic permissive exists. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0026 UFSAR Figure Change Implementation Document No.: EDP 8181 Systen No.: E5100 Title of Change: De-energization of a Reactor Core Injection Coolant injection test valve in the full open position. Snamary: De-energization of RCIC injection test valve E51F012 in full open position. The valve was evaluated to determine if de-energization, with the valve full open, is an acceptable alternative to motor-operator replacement. l
I Enclorura'to NRC.89-0044 Page 33 i Safety Evaluation Summary: l The E51F012 motor-operator was sized for a maximum capacity of 14,000 lbs_ force total thrust. IE Bulletin 85-03 differential pressure i requirements resulted in a calculated thrust exceeding 20,000 lbs-force. E51F012 was installed'as a block valve for E51F013 which . protects low pressure piping on the RCIC pump suction side. It is equipped with a motor operator, and associated circuitry, to allow auto recuery for RCIC initiation. This allows the RCIC discharge ! line outboard isolation valve, E51F013, to be stroke tested during
- power operation. Per SER Supplement No. 4, Appendix K - Attachment'1, 6.1, a relief request was approved which exempts E51F013 from stroke :
testing except during cold shutdown. Therefore, E51F012 is not required to be closed during power operation for testing E51F013 It may be treated as a locked open manual block valve by de-energizing the motor operator and maintaining continuous remote position indication. j DESIGN CHANGE
SUMMARY
)
Safety Evaluation No.: 88-0032 UFSAR Figure Change Implementation Document No.: EDP 6937 System No.: H1100 Title of Change: Reactor isolation valve mimic display. Summary: , Isolation valve mimic groups 1 to 18 isolation actuation and valve full closure input signals were added. Safety Evaluation Summary: l Failure mode analysis of this change demonstrated that the ability of I the IE systems, which interface with the Reactor Isolation Mimic display will not be impaired by any postulated electrical fault on the non_IE circuits associated with them, l l l r l
~
c t EncloCurG to'- NRC-89-0044 Page 34 DESIGN CHANGE SUMARY Safety Evaluation No.: 88-0038 Implementation Document No.: EDP 7907 Systen No.: B2100, C4100, E1100, E2100, G3300, P4400 , Title of Change: Replacement of soft seat material in various swing check valves. h aary: Various swing check valves furnished by Anchor / Darling are designed with soft seats in combination with metal-to-metal seats. One of the soft seats had been subject to failure. Safety Evaluation Sammary: The following swing check valves were furnished by Anchor / Darling and designed with soft seats in combination with metal-to-metal seats: Nuclear Boiler Feedwater Inlet Outboard Primary Containment Check Valves (B2100F032A & B),_ Nuclear Feedwater Supply Check Valves (B2100F076A & B), Standby Liquid Control Outboard and Inboard Check Valves (C4100F006 & 7), Residual Heat Removal Low Pressure Coolant Injection Line Check Valves (E1100F050A & B), Core Spray Division 1 & 2 Inboard Primary Containment Check Valves (E2100F006A & B), Reactor Water Clean-up to Feedwater Spring Assist Close Check Valve (G3300F120) and Emergency Equipment Cooling Water Division 1 & 2 Supply to Drywell Equipment Check Valves (P4400F282A & B). One of the soft seats was subject to failure. Corrective action consisted of replacing the original soft seat with a geometrically identical soft seat constructed of a new, improved ethylene propylene material (Parker Compound E692-75). This material did not change the physical, functional or operational characteristics of the valves and increased valve reliability. l l 0
Enclo:ura to i NRC-89-0044 Page 35 DESIGN CHANGE
SUMMARY
, Safety Evaluation No.: 88-0039 UFSAR Text Change Figure Change Implementation Document No.: EDP 8350 System No.: N3000 l
Title of Change: Moisture Separator Reheater modifications to vent element drain and vent piping. Summary: The purpose of the engineering design package was to implement the GEC Turbine Generators Limited (GEC) modification to the, Moisture Separator Reheater (MSR) vent element drain and vent piping. Safety Evaluation Summary: The modifications implemented by a EDP were designed to decrease unsatisfactory thermohydraulic oscillations in. the reheat section of the MSR which lead to failures of reheat tubes. During the start up i test phase of EF2 the thermohydraulic oscillations resulted in widely
~
fluctuating steam flou rates supplied to the reheater and unstable , operation of the Reheater Drain System. GEC reviewed plant operating data, the MSR design and the Reheater Drain and Vent Systems and ! concluded that the modifications of the vent element drain and vent will lead to more stable MSR operation. The root cause of the oscillations was believed, by GEC, to be the inability of the vent element to operate as originally designed. A less than ideal design of the Reheater Drain System prevented the vent element from operating as designed. Modification of the Reheater Drain System was eliminated as a viable solution due to the complexity, time required and limited space available for properly modifying the Drain Systen. The design pressure and temperature of the initial design are not exceeded and the nature of the modification did not change or increase the possible failure modes of the MSR.
Enclo;urs to NRC-89-0044 Page.36 DESIGN CHANGE SUlttARY Safety Evaluation Wo.: 88-0068 UFSAR Text Change Figure Change Implementation Document No.: EDP 8684 System No.: E2100, P4400 Title of Change: Modification of EECW and EESW system controls to
. provide an auto-initiation in the event of high drywell pressure.
Summary: A previously implemented engineering design change (EDP 55'14) allowed ! for the isolation of the drywell loads upon high drywell pressure. This modification changed the logic such that both Emergency Equipment Cooling Water (EECW) and Emergency Equipment Service Water (EESW) would initiate and the drywell isolate upon the high drywell pressure signal. Safety Evaluation Summary: The design intent of the RBCCW/EECW Systems is for the Reactor . Building Closed Cooling Water (RBCCW) to operate during normal , operation and the EECW System to operate during accident conditions. ! During a Loss of Coolant Accident (LOCA) without a Loss of Power Accident (LOPA) it is possible for RBCCW to operate without actuation of EECW. In these cases operator action is relied upon for the initiation of EECW. In order to reduce the impact on the operator during LOCA conditions the modification was implemented to automatically initiate EECW on a high drywell pressure and isolate the drywell. This added initiation signal provided for automatic initiation of EECW and isolation of the drywell portion of EECW, thus insuring that EECW will support an accident without operator action. A small, intermediate and large coolant line break inside the drywell, is indicated by a high drywell pressure signal. The added initiation ! signal is high drywell pressure, a safety related (1 out of two taken twice) signal that also results in a Reactor scram. At the sanie time with the new logic EECW would initiate and the drywell would isolate. The present logic requires operator action, LOPA, or loss of RBCCW to place EECW in its accident support configuration. The design change adds the automatic action capability. l l l
Enclosure to NRC-89-0044 Page 37 DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0070 UFSAR Figure Change Implementation Document No.: EDP 1809 System No.: P3300 Title of Change: Corrosion product sampler panel. Summary: This change document provides for installation of a corrosion product sampler panel for condensate and feedwater, installation of a new sample line for reactor feedwater, manifolding heater drain sample lines to the forward pumped drain sample line for remote monitoring, and installation of two new roughing coolers away from high personnel traffic areas. Safety Evaluation Summary: The modifications described in the ELP are strictly related to, non-seismic, non-safety-related, QA Level II and III components located in the Turbine Building. The changes provide a better tool to monitor water chemistry and thus are considered an enhancement of the sampling system. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88 0072 UFSAR Figure Change Implementation Document No.: EDP 8170 System No.: B2100 Title of Change: Removal of SRV air bleed line isolation valve. Summary: Nutech Engineers conducted a safety relief valve (SRV) discharge test. The test's instrumentation and associated air bleed system were installed under a Temporary Modification. This removes equipment installed to perform the subject test which is not required for normal i plant operation. The modification encompasses the following: 1) i removal of the air bleed isolation valve B2100F106 (V17 2582) and capping.the remaining pipe stub, 2) plugging the pressure sensor tap on the discharge line from SRV B2104F013J (at elevation 587'_10"), 3)
Enclosure to NBC-89-0044 Page 38 determinating and removing cables 200430A-0C and 200431A-20, and 4) determinating power supply cables 200442-0C and 200443-0C. Safety Evaluation Summary: This modification removes a temporary testing hardware configuration and restores plant to permanent plant configuration; as intended in the UFSAR. The only actual change to the UFSAR is the deletion of a 1" valve shown on the 2 P& ids. Capping this connection instead allows for enhanced pressure integrity by minimizing loading at the
- connection to the SRV discharge line. It also enhances ALARA by removing an unnecessary valve from inside the drywell.
DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0114 UFSAR Text Change Figure Change Implementation Document No.: EDP 9043 System No.: G1100 Title of Change: Drywell sump pump integrator modifications. Summary: Nuclear Production observed that when drywell sump pumps (G1101C001A/B
& G1101C006A/B) were not running, their respective flow integrators, G11-K601 and K603 would sporadically " click" introducing a false count. Although the false number of counts per hour (typically 4) was within DC-4563 tolerance of 4.5 CTS /HR, it was decided to eliminate the problem via this modification.
Safety Evaluation Summary: The work scope of this modification was considered QA Level III. However, the work packages was given a QA Level I, Seismic I designation only because the new auxiliary relays, G1101M001A/B & G1101M002A/B would be mounted in QA Level I, C0P H11-P602. The flow integrators (G11-K601 and K603) affected by this modification were classified QA III components and not required for plant safe operation and shutdown. They are used in conjunction with the Leak Detection System to monitor the total flow out of the drywell floor drain and equipment drain sumps, respectively. This modification was implemented to enhance the accuracy of flow integrators, G11. ' ' and K603 by eliminating the " false count" introduced whenever tr- .ywell sump pumps were not running. Four relays (one for each of the four auxiliary drywell sump pump control circuits) in COP, H11-602 was l
Enclosura to c NRC-89-0044 i Page 39 l' l' installed.- Appropriate contact from each relay was used'to bypass the respective flow integrator whenever the drywell sump pumps were not running.' The bypass is removed whenever one or both drywell sump pumps run. The changes do not alter the function or operational characteristic of the existing drywell sump pumps nor any components of the affected instrument loop. The original intent of the design remains unaffected. DESIGN CHANGE SIDE (ARY , l l Safety Evaluation No.: 88-0137 UFSAR Text Change j i Implementation Document No.: EDP 7577 ) Systen No.: C9100, E1100, H3000, T4100 ! Title of Change: Provide alarm window for E0P related secondary containment area high temperature alarms. ) Summary: An alarm window (3D34) in COP panel H11-P603 will be provided for the EOP related secondary containment area high temperature alarms. This window is Process Computer driven and has process Computer driven a reflash capability. A digital open-to-alarm output contact from the Process Computer will provide the signal to actuate the window. Also, the alarm set points for instrument numbers E11-N600A and B have been revised from 175 F to 148 F. Safety Evaluation Stimmary: The modifications in this design change involved the alarm functions , for the EOP related alarms only. No change in any safety-related j control functions were involved. The safety evaluation was written - for the temperature set point revision of the RHR Pump Rooms Leak Detection System Area Temperature Alarm from 175 to 148 F. Reducing these setpoints is conservative since it provides earlier indication of a possible leak in the RHR Pump Rooms.
s Enclorure to NRC-89-0044 Page 40 DESIGN CHANGE SUletARY Safety Evaluation No.: 88-0144 UFSAR Figure Change Implementation Document No.: EDP 8838 i
)
System No.: C9400, T4700' Title of Change: Addition of thermocouple input signals to the Emergency Response Information System (ERIS). 1
% === *y :
Twenty-eight (28) T47 thermocouple input signals were added to the Emergency. Response Information System (ERIS) to calculate weighted volumetric average of the drywell air temperature. These calculated , values are to be used as input to the ERIS Safety Parameter Display ) System's (SPDS) Containment Integrity Display. f Safety Evaluation %==ary: - Confirmatory testing was conducted to verify that the addition of the parallel ERIS circuitry did not degrade the performance of the existing T47 hardware. The affects on accuracy were minimal. The new signal conditioners that were installed are Validyne TC-292-02, which are high quality, qualified and reliable devices which are used at i Fermi 2 in other applications. However, since the present application l was not safety-related, the devices were not categorized as QAI. ] 1 1 t DESIGN CHANGE
SUMMARY
I Safety Evaluation No.: 88-0157 UFSAR Text Change Figure Change Implementation Document No.: EDP 8841 System No.: E4100, H1100, H2100 Title of Change: Addition of signal to the Safety Parameter Display System. Summary: This modification added a torus water level narrow range analog signal from level transmitter E41N062D to the' Safety Parameter Display System (SPDS) which is not safety-related. A signal isolator was required to separate the safety-related output of E41N662D from the SPDS. The change provided an analog signal identical to the existing level l l ___u__ _.__ ._ - ___._____ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - _ - - - _ _ - -
Enclorura to. NRC-89-0044 Page 41 transmitter E41N062B analog signal and was required to comply with NRC SPDS signal validation on a real time basis. Safety Evaluation humary: i The SPDS analog signal was connected to the buffered auxiliary analog ' output of the master trip unit in the torus water level instrument-loop via a qualified signal isolator. This connection to the auxiliary buffered output implied that a direct short across the auxiliary output will not affect the primary channel.. Manufacturer's test data for buffered output validates the above statement. 4 Therefore, though an extra channel has beers added via a qualified IE isolator - any failure 'of even the qualified isolator.will not increase the probability or the consequences of a malfunction of equipment important to safety. DESIGN CHANGE SUMARY Safety Evaluation No.: 88-0158 UFSAR Text Change Figure Change i . Implementation Document No.: EDP 9055 System No.: C9400, T5000 1 Title of Change: Torus pressure input to the ERIS computer. Summary: This modification provided for the addition of Torus Pressure input (ERIS Pt. #277) to the computer on a real time basis, and modification , of the corresponding software program. The change was necessary to j enhance the calculation and SPDS display of the following E0P curves: { t
- 1) Pressure suppression pressure.
- 2) Primary Containment pressure limit.
- 3) Maximum primary containment water level limit.
Safety Evaluation Summary: The torus pressure instrument loop (T50-N414A) affected by this modification is used only for indication / recording and not for direct control of safety related equipment or process. However, the torus pressure indication in the control room is used by the operator in assessing plant safety status during normal, abnormal and emergency operations and is, therefore, a safety-related, QA level I installation. The instrument loop modification was designed such that if any new unqualified components fail, the performance and function of the existing QA level I instruments will not be affected. This is
L l I Enclosura to NRC-89-0044 Page 42 done by using a qualified, QA level I signal isolator (modulator) which provides isolation between the Class 1E and the new non-class 1E j wiring. This means of isolation meets the requirements of Regulatory Guide, 1 75 (Not a requirement for Fermi 2, however, the' isolation technique is the same as used for other ERIS signals and has been . l approved by the NRC).
...________.._ ....... ___......_______ ....__________________........ i I
DESIGN CHANGE SUBetARY Safety Evaluation No.: 88-0173 UFSAR Figure Change l l Implementation Document No.: EDP.8237 System No.: C9400 Title of Change: Software and hardware modifications to the Safety Parameter Display System. I Summary: j To comply with NRC requirements, corrective actions involving system software and hardware modifications were made to the Safety Parameter Display System (SPDS). These modifications involved the replacement of the 4MB bulk storage with a 160MB fixed storage drive. Safety Evaluation Summary: The modification referenced above was made to non-safety-related ' components of the SPDS as an enhancement of the bulk memory (i.e., additional memory). The new memory unit did not affect the performance or functions of QAI level instruments. The replacement of the SPDS/ERIS 4 MB bulk storage with a 160 MB fixed storage drive did not involve hardware changes. The upgrade of the bulk memory unit for increased capacity was considered a system enhancement internal to the computer system and afforded increased system reliability.
~
b
- Enclosura to g NRC-89-0044 p Page 43 DESIGN CHANGE SUlttARY Safety Evaluation No.: 89-0002 UFSAR Figure Change Implementation Document No.: EDP 4828 System No.: G1135 L Title of Change: Centrifuge torque records and' control limit switch l .
changes. l hamary:- This change document was issued to accomplish the following: 1) assign PIS numbers to the radwaste centrifuge torque. control switch and zero speed switch, 2) change the pen-1 scale of the centrifuge torque recorder from the range of 0-100% over to 0-30 lbf. Safety Evaluation Summary: No' field work or system changes were involved. The engineering design package was completed and approved with'a preliminary safety. evaluation performed which indicated that "no safety evaluation was required". However, minor change was made to.the system P&ID which ' affected a figure in the UFSAR. This safety evaluation was performed-to maintain procedural compliance. These paper changes represent no unresolved safety concerns. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 89-0033 UFSAR Figure Change Implementation Document No.: EDP 3784 System No.: N/A ; Title of Change: UFSAR Figure changes. Summary: A Figure in the UFSAR was revised to reflect plant conditions. The revision involved changing the number of concrete plugs from seven (7) to five (5) due to the replacement of the two torus hatches concrete plugs with. steel plugs. Safety Evaluation Summary: The safety evaluation was performed to evaluate the impact of plant conditions on Figures in the UFSAR. The physical plant modifications that were made did not affect equipment. Structural loading considerations showed that no unreviewed safety questions existed. l
l Enclosurs to < NRC-89-0044 Page 44 1 i POTENTIAL DESIGN CHANGES (PDCs) DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0298 UFSAR Figure Change Implementation Document No.: PDC 7986 l System No.: N2100 Title of Change: Reactor Feed Pump Seal Water Return Pump Discharge piping check valve.
'l Summary: ,
j i A check valve was installed in the Reactor Feed Pump Seal Water Return l Pump Discharge piping to prevent draining water from overhead piping ! to the Seal Water Return tank. Snamary of Evaluation: In the original design, the seal water return pump automatically operated when the condenser pressure was higher than 7 psia and seal water return tank was at high level. When the seal water return pump tripped, based on level signal, water in the pump discharge piping could drain back to seal water return tank. Installation of the check valve prevents water draining to the seal water return tank and thereby reduces frequency of seal water return pump operation. When the reactor feed pump is in operation the condenser is in vacuum. The seal water return pump is not required to transfer seal water from the seal water return tank to the condenser. The condenser vacuum will be i the driving force to drain water from this tank. A failure of check valve will not impact the operation of the reactor feed pump. l;
)
']
Enclorura to NRC-89-0044 Page 45 DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0326 UFSAR Figure Change Implementation Document No.: PDC 7814 System No.: E5100 Title of Change: Reactor Core Injection Coolant modification.1. Summary: During High Pressure Coolant Injection (HPCI) system testing for the Startup Test Phase, it was noticed that the Reactor Core Isolation Coolant (RCIC) system would receive an automatic trip on low suction pressure occasionally. This phenomenon would only occur during a HPCI starting transient. The trip signal lasted only for a fraction of a second and cleared. This minor modification was dispositioned to install a two second time delay to prevent RCIC trip due to a spurious 4 negative pressure spike caused by an HPCI starting transient. Summary of Evaluation: 4 The installation of a two second time delay relay in the Reactor Core Isolation Coolant (RCIC) system low suction pressure trip logic will prevent spurious trips due to negative pressure spikes and will improve availability of RCIC system, since the logic circuitry that controls the valves which are automatically closed on turbine trip signals is equipped with manual reset devices so that the valves cannot be reopened without operator action. Requirements for the RCIC system to provide full flow within 30 seconds is not affected by this change. The previously mentioned low suction pressure trip is provided for equipment (RCIC) protection. A two second time delay does not increase the vulnerability of RCIC turbine pump to a loss of suction event. i e
1 I Enclosure to l i NRC-89-0044 j Page 46-DESIGN CHANGE SUlWARY Safety Evaluation No.: 87 0335 UFSAR Figure Change ; Implementation Document No.: PDC 8100 System No.: G1116 Title of Change: Steam trap installation. I m==a ry: This minor modification provides for the installation of a steam trap upstream of Pressure Control Valve G1100F850 to eliminate the presence of condensed steam in the piping. This accumulation of condensed steam caused difficulty in maintaining the required system pressure of 70 psig; this condition was being. remedied by periddic manual drainage of the piping via the upstream Y strainer. Safety Evaluation Summary: The principal safety concerns associated with the Radwaste System are the potential for uncontrolled releases of radioactive fluids and excessive exposure to operations and maintenance personnel. The modifications made by this PDC have no effect on these concerns mainly , because the changes are being made on a non-radioactive system located ! in a clean, non-radioactive area of the plant. These changes will have no impact on the only UFSAR evaluated radwaste related accident ' of simultaneous seismic failure of all tanks, since no direct or indirect interface with these tanks exists. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0392 UFSAR Figure Change Implementation Document No.: PDC 8294 System No.: B3100 Title of Change: Limiter setpoint reduction. Summary: The setpoints of Limiter #2 (B31K621A and B) and Limiter #3 (B31K812) were reduced to 42% and 48% respectively to reflect the actual correlation between reactor recirculation pump A and B speed and reactor power at 100% of rod line. i O
.1 Enclo:urs to NRC 89-0044
~
Page 47._ i j q Safety Evaluation Summary:' The changes are conservative in nature since it increases the margin of feedwater inventory during both transients (loss of one reactor
' feed pump'and heater drains pumps) required ~to maintain reactor water ,]
- level above level 3 (scram).
DESIGN CHANGE SUlWARY ,
, I
- Safety Evaluation No.: '88 0006 Implementation Document No.: PDC 8315 l System No.: E1100 i
Title of Change: - Leakage pressure monitor alarm setpoint' changes. ! Summary: The alarms associated with the leakage pressure monitors (LPM) were set at values above the setpoints of.the pressure relief valves in the-RHR system.. Therefore, if the condition being monitored by the LPH was exceeded, the alarms would not actuate and the operators may not . l have been aware of.the situation. The modifications set the setpoints to values that were below the relief valve setpoints.- Safety Evaluation Summary: The setpoint changes merely assure that the operator would be alerted to leakage at the high/ low pressure valve interface in accordance with l the original design intent. The failures of the alarm to actuate .! would allow the low pressure piping to achieve the pressure relief-valve setpoint without operator awareness. The failure of the RHR piping was previously analyzed. The change of alarm setpoints does not affect that previous analysis. This modification was implemented .i upon NRC approval of a required Technical Specification change. i ___. _.___________-______-_m_m____________m_.___________m.__ _ _ _ _ _ _ _ _
Enclo:;urs to NRC-89-0044 ' Page 48 DESIGN CHANGE
SUMMARY
I 1 Safety Evaluation No.: 88_0014 UFSAR Figure Change Implementation Document No.: PDC 7829 l 1 Systen No.: G5100, H1100 ] Title of Change: Installation of restraint block in Control Room chart recorder. Summary: Installation of restraint block to prevent cover of chart recorder. l C51R603 from depressing Back_up Manual scram button on Control Room i panel H112P603 Safety Evaluation Summary: The minor modification installed a restraint block to prevent the-cover of chart recorder C51R603 from depressing the back-up manual scram button on Control Room panel H112P603 The addition of the restraint block had no impact upon the function / operability of the , chart recorder; saismic qualifications were not impacted. l l ______________________________________________________________________ l DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0033 UFSAR Text Chatge Figure Change Implementation Document No.: PDC 8511 System No.: E5100 Title of Change: Removal of a Reactor Core Injection Coolant steam admission bypass valve function. Snamary: Target Rock valve E5100F095 required frequent maintenance which affected Reactor Core Injection Coolant (RCIC) availability. This modification removed the RCIC steam admission bypass valve function provided by the small one inch (1") E5100F095 valve.
Enclorura to 1 NRC-89-0044 j Page 49 J Safety Evaluation Summary: Removal of the RCIC steau admission bypass valve function provided by j E5100F095 was evaluated for its effect on RCIC system operability. ! The small one inch (1") E5100F095 valve was originally added to limit steam admission to the turbine during startup, and preclude overspending of the unit which.can occur if the admission valve opens before the control valve has closed sufficiently. Removal of the steam admission bypass function caused the first peak of turbine speed to be higher than it would have been otherwise, but the peak was less 1 than the acceptance criteria values. DESIGN CHANGE SUltiARY . Safety Evaluation No.: 88-0034 UFSAR Text Change Figure Change. , Implementation Document No.: PDC 8397 System No.: P3300 Title of Change: Installation of a sample point in'the common j feedwater header Summary: The purpose of the minor modification was the installation of Sample Point 22 in the thirty-six inch (36") common feedwater' header. Safety Evaluation Summary: Based on proposed changes to the feedwater sampling system, a final . feedwater sample point was required to monitor corrosion products and ! feedwater quality. Since no connection existed, a new connection was made to the thirty-six inch (36") common feedwater header. The new connection, Sample Point 22, consisted of a sample probe nipple and capped isolation valve. The tubing and associated instrumentation will be installed under EDP-1809
L ! i Enclorura to NRC-89-0044
~Page 50 l 1
DESIGN CHANGE SUMARY l Safety Evaluation No.: 88-0036 l Implementation Document No.: pdc 8356 System No.: R3000 1 Title of Change:. Energency Diesel Generators 12 and 14 fuel oil ; transfer pump motor foundations. l Summary: Reinforced concrete piers R3000c004.and R3000co12 support the electrical pump'notors for EDG 12 and EDG 14. Adjacent to these . piers, but not mounted to them, are the. fuel oil lines for the ! transfer pumps.- These lines have strainers which require periodic. l cleaning or replacement. The fuel oil lines were routed such that the strainers'could not be removed because they hit the subject concrete piers at their top southwest corners. To facilitate removal of these strainers, a small area of the piers was chipped away. It was determined that this would be more cost effective than re-routing the fuel lines. 1 Safety Evaluation Summary: ! The seismic qualifications of the motors mounted to these piers were reviewed and determined to be adequate. The insignificant cosmetic change would not have any impact on the capability of the pump motors or structural integrity of the piers.
...................................................................... l DESIGN CHANGE
SUMMARY
i Safety Evaluation No.: 88 0045 Implementation Document.No.: PDC 8327 System No.: N3000 Title of Change: Elongate mounting holes for main turbine stop valve limit switches to facilitate setpoint adjustments. l humary: Modifications were made to the mounting plates for the main turbine stop valve position limit switches. These switches input to the RPS, and cause a scram if the valve is less than a nominal 95% open, (93% O _ - _ . - - _ _ _ _ _ _ _ . _ - . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Enclosura to
- NRC-89-0044 Page 51 (
t
'l l
is the Tech Spec " allowable limit"). Prior to this modification, the j setpoint was restrained to finite increments of the serrated shaft to l which the limit switch moving arm is attached. This restricts the l ability to " fine tune" this adjustment. The modification, i.e. l elongating the mounting holes, allows infinite resolution within the j limits of the slotted holes. ! Safety Evaluation Summary: These was no change to the mounting as compared to the original (i.e. . two (2) bolts and washers (DECO File T1-875)). The new mounting offers essentially the same friction force. The elongated mounting holes were only to facilitate setpoint adjustment; the holes serve no ( structural function. The minor modification also did not move the actual setpoint but only permitted " fine tuning" of the setpoint to obtain higher resolution. 1 DESIGN CHANGE SUltiARY ; Safety Evaluation No.: 88-0060 Implementation Document No.: PDC 7982 Systen No.: B2100 f l Title of Change: Locating thermocouple at "AVC0" MSIV manifold. I assemblies and Target Rock SRV solenoid valves. Summary: l
)
Drywell temperature data obtained from drywell cooling steady state data tests provided ambient temperatures at the Main Steam Isolation Valve (Elevation 589-6") and Safety Relief Valve (Elevation 612'_9") locations. These temperatures did not reflect the additional temperature contribution that the MSIV manifold assemblies and SRV solenoid valves saw due to main steam line " process" conditions under plant operation. The qualified life of these items were based on 135 F (Technical Specification average air temperature requircuient). Actual device temperatures had to be obtained to calculate the qualified life at service conditions (ambient and process temperatures) in order to maintain environmental qualification. Safety Evaluation Summary: One small thermocouple was mounted at each of the following locations: the MSIV manifold assembly and the SRV solenoid. In each case, the lightweight (2 ounce) thermocouple lead was held in place by
' 4
Enclorure to NRC-89 0044-Page 52 _; i an adhesive which was qualified to maintain its integrity in the harsh environment of the drywell. The thermocouple extension wires were supported from Seismic I structures via.ty-raps which supplied a secure but flexible mounting fron the wires. . The functions of the ; affected solenoids was not affected by the modifications. 1 t DESIGN CHANGE
SUMMARY
'l Safety Evaluation 50.: 88 0073 :
Implementation Document No.: PDC 8750 i System No.: T2100 Title of Change: Rework of watertight door RSB-2 Summary: The thrust bearing which supports the 16,000 pound RSB-2 door (SE. .
' Quad) wore out and hindered the opening and closing of the door. This !
minor modification was issued to replace the broken bearing; some of the original design thrust bearing housing was modified.
-Safety Evaluation Summary:
The replacement of the thrust bearing for door RSB-2 had no impact ; upon original design intent. The replacement was a like-for-like- t replacement with the exception of minor base plate dimensional changes to accommodate installation with the door in the closed and secured position. ______________________________________________________________________ 2 DESIGN CHANGE
SUMMARY
1 Safety Evaluation No.: 88 0075 UFSAR Text Change Implementation Document No.: PDC 8552 System No.: G3300 Title of Change: Reactor Water Cleanup system annunciator change. Summary: Change annunciator 2D102 inscriptions from "RWCU filter demineralized discharge pressure high/ low" to "RWCU blow down pressure high/ low". The previous inscriptions did not accurately describe the plant i i 5
. I h
. -1
Enclogure to NRC-89-0044 Page 53 l l parameters that are monitored. The annunciator 2D102 is actuated by l instruments PSL-G33-N013 and PSH-033-N014 which sense low and high blow down line pressure respectively. The previous inscriptions were misleading. Safety Evaluation m==acy: This change modified the alara 2D102 description to one which better describes affected equipment. This alarm alerts the operator of abnormal blow down line pressure and prevents pulling a vacuum from the condenser on low pressure and overpressurization of low pressure downstream piping. The high/ low pressure setpoints remain unaffected. Therefore, no unreviewed safety question exists. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0105 Implementation Document No.: PDC 8987 System No.: R3100 Title of Change: MPU numbers 1 and 2 time delay auto retransfer disablement. Summary: Line fuses of MPU #2 randomly bleu during auto transfers of the MPU from its alternate power source to its normal power source. It was determined that if the transfer took place within a cycle and the normal source was out of phase (-90 ), a high inrush current resulted. With the power being out of phase, the result was that of having an EDG supplying power to the normal source. The MPU logic was changed so that the power seek mode remained but the 2 minute auto throwback was defeated for MPU #1 and #2 only. Safety Evaluation Summary: Due to the pattern which emerged that appeared to tie the blowing of fuses to automatic re-transfer of supply to an Emergency Diesel Generator-supplied normal source, it was proposed to disable this automatic re-transfer function. Manual transfer to the normal source was still possible, but this is slower than the one cycle automatic re-transfer and there is less likelihood of transient disturbances being caused by the voltage regulators or other equipment. The other automatic functions of the relay logic remained, i.e. seeking normal supply voltage if switched to the standby supply and it fails. The
. g,
1 I Enclo;ura to NRC.89-0044 l Page 54 modifications in connection to time relay will achieve this disablement of automatic re-transfer without other complications. DESIGN CHANGE SIDMARY ) l Safety Evaluation No.: 88-0123 Implementation Document No.: PDC 8994 System No.: R3000 Title of Change: Addition of o-ring seals to EDG emergency stop ] buttons. Summary: 1
-The Emergency Diesel Generators (EDGs) are' equipped with an emergency l stop button that when pushed terminates the flow of fuel oil, thus stopping the unit. Original design configuration of these push button mechanisms indicated that'through normal wear between the metal-to-metal components (stainless plunger and brass line housing) -
that both became scored, thus providing a potential oil leak path and fire hazard. Based upon vendor recommendations, a snug fitting 0-ring i was placed on the plunger shaft so that the seal would no longer be metal-to-metal but metal-to-o-ring to metal. ! Safety Evaluation Snmancy: The actual change, addition of the 0-rings, reduces the potential of-oil leakage and the resulting possibility of a fire. Failure of the 0-ring would return the pushbutton mechanism to the metal-to-metal seal which had originally been analyzed to be acceptable. The 0-rings are considered an enhancement which improve system availability. l l I
i f Enclosurs.to NRC-89-0044 Page 55 l o l DESIGN CHANGE SUletARY Safety Evalunt. ion No.: 88-0130 UFSAR Text Change Implementation Document No.: PDC 9187 System No.: N3000 Title of Change: Main turbine trip setpoint change. Summary: This modification changed the setpoint on Main Turbine to trip at-5 5 inches Hg absolute (increasing) main condensce pressure instead of 4.5 inches. Safety Evaluation Sunesry: The change in trip setpoint reduces the availability of the turbine bypass system by 0.5 seconds. This small change in the bypass system l availability has no impact on transient analysis, because the bounding J transient for this event (turbine trip) is without bypass. DESIGN CHANGE
SUMMARY
-l Safety Evaluation No.: 88-0147 Implementation Document No.: PDC 7421 System No.: G1100 i Title of Change: Waste sample pumps seal water supply.
Summary: The three (3) Waste Sample Pumps were receiving their seal water from ' the condensate header which introduced " unmeasured" radioactivity. The modification disconnected this water supply and provided seal water from a tap on the pump discharge line. Safety Evaluation Sumanry: The modification was a minor change that dealt only with the connection of (small-bore) properly sized seal water lines to the pumps. This minor piping change to a non-safety system had no direct or indirect interfaces with safety equipment or systems. Releases of
_ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ c Enclosure to NRC-89-0044 Page 56 radioactivity, if any, would remain well within the limits allowable by 10CFR20, Table 2. ____________7________________GN DESI CHANGE SUle(ARY Safety Evaluation No.: 88-0164 UFSAR Figure Change Implementation Document No.: PDC 9384 System No.: G1154, T4500 Title of Change: Drywell equipment drain sump discharge valve modification. Suasary:
.The circuit associated with the Drywell equipment drain sump discharge valve G1154F018 was modified'to provide remote manual isolat. ion capability from the Relay Room Panel H11-P915.
Safety Evaluation Snamary: Addition of backup manual isolation assures valve closure in the event that the automatic isolation fails. Upgrading the BOP manual closure feature ensures system operability and poses no unreviewed safety question. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0170 UFSAR Text Change Implementation Document No.: PDC 9403 System No.: N3000 Title of Change: Main turbine high vibration trip. Snamary: The main turbine generator high vibration trip relay was replaced with a time delay relay (10 seconds) to eliminate unnecessary reactor scrams due to spurious high vibration spikes. l l I _ _ - - - _ - _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ __ __ _ J
Enclo:ura to
'NRC.89-0044 Page 57 1
Safety Evaluation Summary: i Adding a 10 second time delay relay to the main turbine high vibration trip will reduce spurious trip and hence reduce reactor challenge. The main turbine is located in the Turbine Building; there is no " safety" i equipment in the immediate surrounding area. Should this relay fail, , the worst accident condition which could result would be a thrown ) blade (turbine missile) and this situation has been previously evaluated and addressed in the UFSAR. DESIGN CHANGE S(DMARY ] i Safety Evaluation No.: 88-0177 UFSAR Figure Change Implementation Document No.: PDC 9423 System No.: C3200 . Title of Change: Revise the post scram setdown setpoint. Sumanry: ] The Feedwater Control system post scram setdown setpoint circuit (analog) was revised to reduce the affect the input demand signal (i.e., reactor pressure vessel level) had on the setpoint. Safety Evaluation Susancy: The revised circuit reduces the probability of a " Level 8" trip after a scram. This ensures that " safety" equipment such as HPCI and RCIC remain available if required. i DESIGN CHANGE
SUMMARY
)
Safety Evaluation No.: 88-0184 UFSAR Figure Change ! l Implementation Document No.: PDC 9646 l
/
{ System No.: C7100 l Title of Change: Disable RPS-MG set overvoltage protection relay. j i l Sumancy: ; I The overvoltage protection relay was disabled to avoid spurious RPS.HG set (PIS No. C7102S001A&B) overvoltage trips. The change was required 1 l l l
- l
._ - ___ . - _ - - - - ._____________-_______L
I i
~
Enclosura to NRC-89-0044 Page 58 _ 1 t due to the sensitivity of the overvoltage relay to the MG-set vibration and/or mechanical shocks which caused relay contact' status i change and initiated the trip. 1 Safety Evaluation Summary The removal of the MG set overvoltage protection relay did not-adversely affect the function of the MG Set and the removal prevented- l inadvertent tripping. The overvoltage protective function is provided 1 by the safety-related EPA units. These EPA units were added after initial system design to improve the quality of the power protection I for the RPS circuits via use of qualified safety-related protective i devices. DESIGN CHANGE SIAMARY Safety Evaluation No.: 88-0190 UFSAR Figure Change ! Implementation Document No.: PDC 9535 : System No.: G3300 Title of Change: Reactor Water-Clean-up valve modification. Summary: The G3352F119 was changed from a open/close valve to a jog throttle open valve, stop close control. ! ' Safety Evaluation Summary: Valve G3352F119 is the recirculation suction valve for the Reactor Water Cleanup system. This gate valve isolates the RWCU recirculation suction but is not'a containment isolation valve. The safety function of this BOP powered valve is to maintain pressure integrity in the ASME piping system but doos not perform an active safety function. The control circuit change which made it a stop close and a jog / throttle open function had no effect on the pressure boundary integrity. This does not present a new or unique design in the facility. i _ . _ _ _ . _ t ______ _ ______ . i
Enclorura to NRC-89-0044 Page 59 ; DESIGN CHANGE SUMARY Safety Evaluation No.: 88-0200 UFSAR Figure Change Implementation Document No.: PDC 9071,.ABN 9071-1 System No.: G4100, P4200 Title of Change: Fuel Pool Cooling and Cleanup system and Reactor Building Closed Cooling Water system changes. h==acy: The system P&ID for the Fuel Pool Cooling and Cleanup system was updated to delete two drain valves, two relief valves, two vent valves and associated piping and drains. The system P&ID for the Reactor
. Building Closed Cooling Water system was updated to add two drain valves, two relief valves, two vent valves, and associated. piping and drains.
Safety Evaluation Snamary: i This as-built corrects valve number designations in the affected documents to "As-Built, As-Designed" plant configuration. There were no design or hardware changes; document changes only. All of the valves addressed are currently shown in the UFSAR. This notice merely relocated the valves associated with the shell side of the FPCCS heat exchangers to the correct P&ID (i.e., RBCCW, which provides shell side cooling for the FPCCS heat exchangers). There is no impact on any safety-related component, equipment or system required for safe shutdown. DESIGN CHANGE
SUMMARY
l Safety Evaluation No.: 88-0241 UFSAR Text Change Implementation Document No.: PDC 8416 System No.: E5100 i Title of Change: Reactor Core Injection Coolant Discrepancies. i Snamary: l Various sections of the UFSAR which discussed Reactor Core Injection ; Coolant system parameters were updated to clarify existing information or correct grammatical errors. O g
i Enclorure to NRC-89-0044 ! Page 60 i Safety Evaluation %==ary: The operation or function of plant equipment was not changed, and the l
^
intent of each section revised remained the same. The changes made were to clarify or improve the quality of information in the UFSAR.. 1 i 1
. .l
i i Enclosure to I NRC-89-0044 Page 61 AS-BUILT NOTICES (ABNs) DESIGN CHANGE
SUMMARY
l l Safety Evaluation No.: 87-0125, 87-0320 , Implementation Document No.: ABN 3529-1 System No.: P4500 Title of Change: Emergency Equipment Service Water system isolation valves. Snamary: This notice was issued to reflect the as-built configuration of the Emergency Equipment Service Water System to include series isolation valves on five (5) vent and drain connections associated with the supply and return piping to/from the EECW Heat Exchanger. Safety Evaluation Snmanry: This design change reduces the probability of an accident or malfunction by establishing in series two (2), normally closed manual isolation valves on the subject vent and drain connections. The failure of the additional valve could occur either by the valve failing to close or by the valve not maintaining its integrity. In either case, the safety'of the system is not compromised because the valve upstream will maintain system integrity. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87 0146 UFSAR Figure Change Implementation Document No.: ABN 4594-1 System No.: 01100, P7000 Title of Change: Radwaste system and Waste Oil system changes. Sumancy: This as-built notice revised drawings and sections of the Master Instrument List (MIL) associated with the radwaste system and the
7 ( Enclorura to-NRC-89-0044 Page 62 waste oil system to reflect the current plant configuration. .It. changes'the location of G1100L513 (radwaste chemical-tap) and deletes the tubing going to sample sink #1 making this point a local sample point. Pressure taps P7000L404'and L406 are also relocated. The i suction pressure tap'for pump P7000C041 is also deleted.: The MIL is-also updated with respect to reference drawings and design temperature.
' Safety Evaluation %= mary:
These changes deal with the piping connecting the oil / water separator waste water pump discharge with the radwaste floor drain tank.. The function of the oil / water separator is to reduce the oil in the water coming from the oily waste sumps before it'goes to the radwaste floor drains tanks for further processing. The systes' operation is not safety-related nor is there any impact on any safety-related. . function.. Because of this fact and the fact that the changes have no impact on the function of the systems, there is no impact on facility or. procedures described in the text of the UFSAR._ In fact, even if the system was to be inoperational, there would be no impact on the safe shutdown of the plant.- DESIGN CHANGE SUMARY Safety Evaluation No.: 87-0279 UFSAR Figure Change Implementation-Document No.: ABN 4829-1 System No.: N2100 Title of Change: Feedwater. pumps lube oil system pressure switches. Summary: 1 1 This ABN was issued to reflect the "as-built" condition for pressure switches, N2100N502 thru N2100N509 of the Standby Feedwater Pumps Lube . Oil System. The change affected the instrument service description and reference drawings in the Master Instrument List. Also PSE-N2100N503 and N507 were installed as spares. Safety Evaluation %==ary: l Pressure switches N2100N503 and N2100N507 are redundant instruments and are not required for the operation of standby feedwater pump A and B, respectively. The pressure switches are installed "in place" but not electrically wired. These actions were taken based upon a lube oil system evaluation which indicated that the pressure switches would not be required.
- _ - . _ _ -________l______-_-_-
EncloIura to NRC-89-0044
- Page 63 DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87 0317 Implementation Document No.: ABN 7537-1 Systen No.: G3300 Title of Change: Reactor Water Cleanup System revisi6ns. Susanry: Revise piping drawings to show as-built pipe sizes and caps on lines. Add de6a for valves. Revise seismic level. Applicable changes locaton on the three inch (3") suction side piping of Recirculation Pumps .4 and B in the Reactor Water Cleanup system. Seismic level revised from II to II/I. Safety Evaluation Summary: According to Chapter 3 7 of the UFSAR, seismic II/I as applicable to non-safety-related components, has the same meaning as level II. The affected co9conents serve no safety functional requirements. ___________________________________________..____,_____________________ a, DESIGN CHANGE
SUMMARY
l Safety Evaluation No.: 87-0323 UFSAR Figure Change Implementation Document No.: ABN 6dC.-1 System No.: N2100 1 Title of Change: Elimination of reactor feed pump system diagram duplication. Summary: i
)
The primary purpose of this as-built notice is to eliminate duplication of Reactor Feed Pump Suction Piping currently appearing on two different system diagrams; this resulted in changes to seven (7) figures in the UFSAR. Safety Evaluation Sumanry: 1 The changes associated with this as-built notice required moving items l from one drawing to another and eliminating items from being shown on multiple P& ids; these are drawing changes only. These changes have 4
=
v Enclosura to FRC-89-0044 Page 64 l
. 1 been made to prevent possible. confusion. .The operational I characteristics of the Reactor Feed Pump system have not been altered nor have any system design parameters been changed.
DESIGN CHANGE SUlWARY
. Safety Evaluation No.: 87-0342
_ Implementation Document No.: ABN 7536-1 ) J System No.: G3300 i Title of Change: Change of Reactor Water Cleanup valve designation. Summary: i Revise system prints to change G3300F101 valve status from normally closed to normally open. G3300F101 is the-bottom head drain valve and , is used to minimize temperature stratification across the RPV. ! i Safety Evaluation Summary: l Keeping the G3300F101 valve normally oben marginally increases the' l fraction of water that the RWCU system draws off the bottom of the , vessel. It does not affect the total flow or the functioning of this l or other systems. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0366 UFSAR Figure Change Implementation Document No.: ABN 5533-1 Systen No.: P4400 Title of Change: Emergency Equipment Cooling Water system piping changes. Snamary: A previous as-built notice made changes to the Emergency Equipment Cooling Water system function operating sketch but failed to include l the changes on the system P&ID or corresponding piping isometrics. 1 This ABN incorporates those changes.
Enclosure to NRC-89-0044 Page 65 Safety Evaluation Summary: 3 The changes involved included: correcting piping isometrics, switching valve numbers, indicating double globe valve configuration, adding pipe caps, properly identifying types of valves, removing valves from prints that do not la reality exist in the plant. These changes did not require design changes to the system (s) and were evaluated in depth in other documentation (i.e., as-built notices, engineering design packages, design calculation). All changes were made to update plant drawing to represent plant configuration. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0373 UFSAR Figure Change Implementation Document No.: ABN 7624-1 System No.: B3100, C3600, E1100, H2100, P4400 Title of Change: QA level and seismic classification change of level transmitter. Summary: This as-built notice was required to correct a number of deficiencies encountered in EDP 1702. The changes were general in nature, predominately minor miscellaneous corrections, additions, and deletions on design drawings. The QA level and seismic classification of level instrument C36N404 was changed from QAI, Seismic I to QA level IM and non-seismic. Safety Evaluation Sumancy: Level instrument C36N404 is installed on Condensate Storage Tank for level indication only as part of the Dedicated Shutdown System. The said instrument has no safety-related function nor is it interlocked with safety-related equipment or components. Per Detroit Edison letter NE-PJ-88-0163, QA level designation for this instrument should be 1M. Seismic classification is non. seismic since the Condensate Storage Tank itself is non-seismic.
^
1 Enclosure to NRC-89-0044 Page 66 DESIGN CHANGE SIEMARY Safety Evaluation No.: 87-0382 Implementation Document No.: ABN 7066-1 System No.: T4100 Title of Change: Main Control Room HVAC chiller unit changes. Summary: The following changes were documented in this as-built notice: s revision of purge compressor motor ratings from 1/4 HP to 1/3 HP, reduction of control circuit power requirement from 3KVA to 2.47 KVA, miscellaneous editorial document changes to reflect this information. Safety Evaluation Snamary: The change involved the increase in recorded horsepower rating of the purge compressor motors in the Division I and II Control Room HVAC chiller units. The systems to which these motors belonged remained unchanged. 5 DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 87-0383 Implementation Document No.: ABN 8046-1 System No.: N6100 Title of Change: Condensate system level indicator. Summary: This as-built notice dealt with the addition of a level indicator, N61R406, to the South condenser to monitor the water level. Safety Evaluation Snumary: The level indicator, which is a sight glass, is connected to the condenser via an instrument line. The entire system is non-safety related, hence no credit has been taken in accident evaluation of the integrity of the pipes or instruments. A break or leak in the sight glass may cause the loss of condenser vacuum, which may lead to a scram, but will not result in any accident or fuel damage. A leak or
~
1
-i Enclorura to j
-NRC-89-0044 !
Page 67 q l complete break in the sight glass will not result in condenser vacuum j loss at a rate greater than 2 inches per second. The leak in the ' level indicator will, however, cause the release of radioactive condensate. The sight glass under normal circumstances performs a l passive function which has no impact on any other systems. Loss of pressure integrity in the sight glass is its only failure mode and was. j shown to be bounded by present UFSAR evaluations. l I DESIGN CHANGE SUMARY l i Safety Evaluation No.: 88-0013 UFSAR Figure Change j Implementation Document No.: ABN 7438-1 ; Systen No.: N2100 ) 9 Title of Change: Reactor feed pump seal water control loop ] configuration. l i Summary. ; The condensate system P&ID incorrectly showed the control signals for loops N21N427AS and N21N427BS as coming from the test thermowells TWT-N21L455 and TWT-N21L456, respectively. The signals should have. i been shown, instead, as coming from the primary instruments TEW-N21N427 and TEW-N21N427B. Safety Evaluation Summary: , The change represents correction of an error and implementation of the )' original design intent. Besides, both the test therscwells and the primary instruments sense the temperature of the same section of the system pipe. Therefore, there is no change in the real source of the signal that is fed into the control loop. This assures that there is no impact of the functioning of any of the systems. ; I \ \ f I l 1 l. L i t
Enclosure to' NRC-89-0044 Page 68 DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0052 UFSAR Figure Change Implementation Document No.: ABN 8660-1 System No.: T4102 Title of Change: Control Center HVAC system setpoints. m==ary: This as_ built notice revises the master instrument list and Control Center HVAC mechanical system diagrams to reflect the as-built control setpoint for the normal and recirculation modes of the Control Center HVAC system (+ 1/4" W.C.). This control setpoint for both modes of CCHVAC is consistent with the control center pressure requirement (+ 1/4" + 1/8" W.C.) provided in the UFSAR. Safety Evaluation Sammary: This as-built notice revises the Master Instrument List and Control Center HVAC drawings to reflect the as-built control setpoints and
-control pressures identified in the UFSAR. Consequently, this is a i documentation change only. The context of this notice does not impact the safety significant mode of control room pre.ssurization under the recirculation mode of CCHVAC.
I ______________________________________________________________________ j DESIGN CHANGE
SUMMARY
l 1 Safety Evaluation No.: 88 0056 ) l Implementation Document No.: ABN 8610-1 System No.: R3400, R3500 l Title of Change: Reworked wire splice terminations. 1 Si==a ry : 1 This notice proposed an alternative method to reduce the overall size of the splice termination to accommodate existing raceway systems for reworked safety-related instruments. A crimped connection and a small raychem heat shrink were chosen over a bolted connection. i l
Enclosure to NRC-89-oo44 Page 69 Safety Evaluation Sn==ary: Electrical and mechanical characteristics for bolted and crimped connections are the same due to the fact that the ring terminal lugs used for the bolted connection are also crisped (at both ends). The only difference is in the crimped connection of two different sizes of wires, where the wires will be soldered and then crimped. Based on engineering judgement, the crimped sc1dered connection provides a better electrical and mechanical connection. ' LESIGN CHANGE SIDE (ARY Safety Evaluation No.: 88-0069 Implementation Document No.: ABN 7476_1 System No.: R3000 Title of Change: EDG control panel fuse holder replacement. 1 Snamary: The Colt Industries' vendor manual was updated to document installation of Bussmann fuse holders. The Bussmann fuse holder is supplied as a like-for-like replacement of the original, obsolete (Graybar) fuse holder with the same Colt Industries part number. The fuse holder is mounted inside emergency diesel generator (EDG) control panel R3000soo7 Safety Evaluation Snamary: ; Replacement of the fuse holder is considered equivalent or better that ! the original fuse holder (in performance and reliability). The difference in weight between the two is minimal, therefore there is no significant impact upon seismic qualifications. In addition, Colt Industries provided a certificate of conformance for the replacement fuse holder stating that the new fuse holder meets the requirements of the original Detroit Edison order.
. 1 l
l 1 I
Enclosura to l NRC-89 0044 j Page 70 DESIGN CHANGE SUlefARY 1 Safety Evaluation No.: 88-0101 Implementation Document No.:~ABN 9017 1 System No.: R3000 Title of Change: Emergency Diesel Oenerator exciter units replacement relay. humary: At the Emergency Diesel Generator Units the K1 relay was replaced.- The original ITE Catalog A123E relay was replaced with an equivalent ITE Catalog A143E relay which was provided by the vendor as an equivalent replacement. Safety Evaluation Summary: The replacement contactors are identical in design, materials used, construction, rating and have adequate voltage for the coils to perform their intended function. Therefore, the change in contactor , from A123E to A143E is acceptable. . DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88-0108 Implementation Document No.: ABN 5366_1 System No.: T2300, T4600, T4800 Title of Change: Ethylene propylene terpolymer (EPT) soft seats. i Summary: l ABN 5366_1 Rev. o was written to revise design drawings and Equipment Qualification Manual to reflect TEFZEL (w/ Silicone o-ring) as an alternate seat material for T2300F407, T2360F410, T4800F404, T4800F405, T4600F400 and T4600F401 and also to update the design ; drawing for T2300F410 to reflect material change (carbon steel to ! stainless steel) for the valve wafer and shaft. Rev. A of ABN 5366-1 revises the recommended material for the 2 piece TEFZEL soft seat with o-ring backing ring from Silicone to Ethylene
.1
- Enclosuro to.
. NRC_89-0044 ,
l Page 71 Propylene Terpolymer (EPT) and provide a corrected. alternate detail indicating the correct installed location for the.o_ ring backing ring. Safety Evaluation %= mary: The replacement of the carbon steel wafer and shaft, and the use'of .- EPT o-rings improve the valves reliability and enhance each valve to perform its safety function; The changes described improve the functional capability of the valves and further enhance their ability; 1 to perform properly. These changes lessen the possibility of a valve. ) malfunction caused by excessive binding of.the valve shaft, or , j degradation of the seat materials. ] ______________________________________________________________________ 1 DESIGN CHANGE SUletARY j Safety Evaluation No.: 88-0109 UFSAR Figure Change'- Implementation Document No.: ABN 8368 1 System No.: B2100 Title of Change: Main Steam Isolation Valve drain detail. Summary: As-built notice adds Main Steam Isolation Valve drain and incoming heater drain detail'to the Nuclear Steam Supply System P&ID drawings. Drawing cross-reference notes are also added. In addition, this notice incorporated changes reflected in Minor Modification PDC-8245, Rev. A which showed correct open/ closed valve positions and' flow directions. Safety Evaluation Summary: The revisions to the drawings reflect routing and modifications to agree with other drawings and will improve the ability to obtain and l upgrade all pertinent information for this system. Cross-referencing the drawings will enhance future system documentation updating and ! information retrieval. t I i
1 Enclorura to NRC-89-0044. Page 72-
-DESIGN CHANGE SUlstARY Safety Evaluation No.: 88-0115-Implementation Document No.: ABN 8829-1 y 1
System No.: T4600 Title of Change: Cardox CO2 fire suppression system... Summary: This as-built notice revised the source valve for actuation of.the 1
.CO2 fire suppression system, revised piping and valve configuration and.added PIS numbers for. valves and-' instruments.. Flow diagrams and functional operating sketches (FOS) were revised to show as-built-configuration. i l
Safety Evaluation Summary: The CO2 fire suppression system is a vendor (Chemetron) engineered-and supplied package system to meet Fermi 2 requirements. The drawing revisions associated with this ABN reflect the design specified by the - vendor. The current-drawings do not properly represent the vendor's. design. This ABN correlates and reconciles Fermi 2 design drawings with the vendor's design. Also,'the system FOS is utilized for' system
.line-up verification and locked' valve positions as applicable. ~The FOS revisions contained within this ABN reflect locked open valve-positions. These valves are in the flow path of the CO2 discharge j system and required to be open all the time. !
..._____..___..._____..._______________________________________.______ 1 DESIGN CHANGE
SUMMARY
l Safety Evaluation No.: 88-0128 UFSAR Figure Change Implementation Document No.: ABN 9138 1 System No.: G5100 Title of Change: Torus Water Management pump suction strainer.
%= mary: ;
This notice was issued to update the design documents for the TWMS to l: indicate that strainers G5100D001 and G5100D002 (TWMS Pump Suction Strainers) had their' strainer screens removed and have no straining type of function. This was completed as one of a number of items t 4 I 1 I
.. _ _ - _ _ - _ _ _ _ _ _ . _ _ _ 1
1
'Enclosura to NRC-89 0044' Page.73 i
recommended by a special task force reviewing the Three Mile Island incident. Safety _ Evaluation Suasary: . The TWMS has twin pumps for reliability and the elimination ~of the . screens eliminates the: possibility of the strainers becoming clogged.. The ABN was issued to place a note on the relevant design documents to alert plant personnel that the strainers internals are not installed. This portion of the TWMS is a QA III level system and.coes not affect safe shutdown of the plant nor is it used in any basis for the margin of safety. DESIGN CHANGE SUltlARY ._ l i Safety Evaluation No.: 88-0135 UFSAR Text Change j Figure Change l Implementation Document No.: ABN 5936-1 i System Nc.: C4100 Title,of Change: Determination of QA levels for the Standby Liquid Control system'. Summary: 1 ' Standby Liquid Control System (SLCS) components that were supplied as QA Level I will be identified as QA Level I in CECO. This applies to , the containment isolation check valves and manual valves that form the SLCS Boron injection path from the storage tank to the Reactor vessel and manual valves that maintain the flow path boundary. QA Level 1H will apply to storage tank level and-temperature instrumentation. Safety Evaluation Summary: The nature of the revisions (revision of QA level) does not alter the design or function of either the SLCS or the plant. The assignment of CA Level 1H to SLCS components and their inclusion in CECO enhances the attendant maintenance activities to ensure more reliable operation of the SLCS. l i l i i l i
)
~
a
Enclo2ura to NRC-89-0044
" Page 74 DESIGN CHANGE SUle(ARY Safety Evaluation No.: 88-0136 UFSAR Figure Change Implementation Document No.: ABN 8390-1 System No.: B2100 Title of Change: Main steam safety relief valve numbering.
i Summary: In order to minimize the impact of Safety Relief Valve (SRV) i maintenance on design documents, this change document was prepared to: j a) establish a "V' number to PIS number cross-reference on one P&ID j (6M721-5538) for SRVs, b) remove the "V" number from other design documents and either add the PIS number or a' note referring to the matrix on the P&ID as appropriate. j f Safety Evaluation Summary l This design change will greatly simplify the task of keeping SRV l design documents up-to-date, since only one drawing will need to be j revised following maintenance involving SRV assembly change-out. l There is no impact on any design basis. The design bases for the SRVs '! are a function of the installed location, which is a function by the !' PIS number. No PIS numbers are being changed, rather PIS numbers are being used to identify which particular installed location a particular design document applies to.
---------------------------------------------------------------------- \
DESIGN CHANGE
SUMMARY
l!
-l Safety Evaluation No.: 88-0168 UFSAR Text Change f Implementation Document No.: ABN 8699-1 l Systen No.: R3400 !
1 Title of Change: Cable and wire changes. ) i Summary: Revised the QAL I cable specification 3071-80 to conform to the latest l IEEE, ANSI and ICEA st6ndards, especially flame test revisions per j IEEE 383-1974. Added to specification 3071-80 the " Quality Assurance Program Specification for Class 1E cable". Included scope to include ! i high temperature (rated 125 C) cable. i I i
~
Enclorura to NRC-89-0044 r Page 75
' Safety Evaluation Summary:
The cable purchased to the revised specification will'aeet;or exceed that which is already qualified, purchased, and installed in the plant and will-not pose an unreviewed safety question. < j l DESIGN CHANGE SUINIARY l i Safety Evaluation No.: 88 0174 UFSAR Figure Change a Implementation Document No.: ABN 9393-1 i System No.:: B2100, E2100, C1100, G5100, P1100, P4100, P4400, P5000, T4800, U4200 1 Title of Change: Locked valve program. h==acy: This as. built notice revises P& ids and FOSs to properly identify those j valves which are required to be in a locked position. Letters l NE-PJ-88-0449 and procedure POM 21.000.14, identifies the valves-that ,
'~ are in.the plant. These letters and procedure provided;the' criteria i and administrative. controls for the-locked valve program.- This program ensures that'the applicable valves remain positioned in a manner that will maintain proper operation of the engineered safety '
features and prevents the unmonitored release of radioactive
. material. The locked position is consistent with-the normal plant
-operation modo. q 1
Safety Evaluation Summary: No hardware, design bases or functional requirements are introduced or j modified by this document. DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88 0204 UFSAR Figure Change Implementation Document No.: ABN 8413-1 System No.: B2100, B3100, C3200, C3500, C3600, C9400, E110, P4400, , P4500, R1400, T4700, T5000 Title of Change: Technical Specification Review Group , responsibilities.
- - - - - - _--____1___-- ___________ 1
Enclosurs to NRC-89-0044 Page 76 I l
%==a cy: ,
Nuclear Prodt$ction has implemented a program to identify instruments and control devices associated with Technical Specification surveillance requirements. These devices and instruments must be properly identified and classified to ensure that corrective and preventive maintenance is performed as required. The Technical Specification Review Group (TSRG) has been assigned the responsibility of identifying these components and associated deficiencies / improvements. During the TSRG review of surveillance procedures, various drawings have been identified as having drafting errors, missing, incomplete or misleading information. This ABN corrects these deficiencies by revising twenty _four (24) Basic Configuration Design Document (BCDD) identified by TSRG. j Safety Evaluation Summary: All the above changes are strictly drawing revisions for providing ' correct and~ complete information to improve the quality of drawings. There exists no requirement for any plant modification. All information used in these drawing revisions are obtained from existing ! design documents and from the as-built plant configuration. No system design or equipment is physically or functionally changed by this ABN. It does not alter the original intent pf system design or' function j DESIGN CHANGE
SUMMARY
Safety Evaluation No.: 88 0237, 86-0217 UFSAR Figure Change j Implementation Document No.: ABN 6453 System No.: P4400, P4500 Title of Change: System drawing updates. Summary: The actual changes made to the two P& ids (6M721-5357 and 6M721-5444) can be categorized as follows:
- 1. Changing valve symbols from Gate Valves to Globe valves.
- 2. Adding a second Vent Valve or Drain Valve to vent or drain lines.
3 Delete a vent valve.
- 4. Show valve identification tags on the P& ids for various vent or drain valves which are not identified with a valve number.
- 5. Add Pipe Caps to the ends of vent and drain lines.
l l
Encloruro to 4 NRC-89-0044 i l
- Page 77 Safety Evaluation S - y:
The changes ellainate discrepancies between the P& ids,.FOSs, corresponding Isometrics Tria Sketches, CECO.and the actual field configuration. 'These changes do not constitute any design changes to the system nor chane any procedures or assumptions described in the UFSAR. The double vent and drain valves being added to the P&ID and FOS are not a result of new valves being'added to the system. . This is a case where the valves' originally existed in the system, but the P&ID , i and FOS neglected to indicate them. These double vents and drains are already used at various other locations in the EECW and EESWS. Therefore,-addition of these other vents and drains is in agreement with general design of vent and drain connections for'this system. Therefore, since it does not involve hardware changes, new equipment, i design change, affect technical specifications nor constitute a special test, this is not an unreviewed safety question. This ABN documents the "as-built", "as designed" configuration. i l I i
)
I m____--________ _ _ _ _ _ _ _ _ _ _ _ _ _ _
-)
Enclosura to NRC-89-0044 Page 78 1 The following As-Built Notices (ABNs) resulted in drawing changes; these changes were reviewed for potential safety consequences. Since. 1 the modifications made to the affected drawings were so slight; a l summary for each was not prepared. Instead these ABNs, along with. l their associated safety evaluations, have been listed for future ; reference. Safety Evaluation No.: 86-0198 UFSAR Figure Change { I Implementation Document ABN 7624-1 Safety Evaluation No.: 86-0232, 88-0221 UFSAR Figure Change Implementation Document ABN 6514-1 Safety Evaluation No.: 86-0244 UFSAR Figure Change Implementation Document ABN 5743-1 Safety Evaluation No.: 86-0271 UFSAR Figure Change , Implementation Document ABN 6336-1 Safety Evaluation No.: 87-0104 UFSAR Figure Change Implementation Document ABN 6849-1 Safety Evaluation No.: 87-0121 UFSAR Figure Change Implementation Document ABN 6077-1 Safety Evaluation No.: 87-0158 UFSAR Figure Change Implementation Document ABN 7109-1 Safety Evaluation No.: 87-0179 UFSAR Figure Change Implementation Document ABN 6786-1
~
Safety Evaluation No.: 87-0183 UFSAR Figure Change < Implementation Document ABN 3189-1 Safety Evaluation No.: 87-0216 UFSAR Figure Change Implementation Document ABN 2979-1 Safety Evaluation No.: 87-0217, 87-0395 UFSAR Figure Change Implementation Document ABN 2775-3 Safety Evaluation No.: 87-0221 UPSAR Figure Change Implementation Document ABN 6471-1 i Safety Evaluation No.: 87-0278 UFSAR Figure Change Implementation Document ABN 5761-1 Safety Evaluation No.: 87-0279 UFSAR Figure Change Implementation Document ABN 4829-1
Enclo ure to
'NRC-89-0044 Page 79 Safety Evaluation No.: 87-0288 UFSAR Figure Change Implementation Document ABN 6481-1 Safety Evaluation No.: 87-0297 UFSAR Figure Change Implementation Document ABN 7199-1 Safety Evaluation No.: 87-0302 UFSAR Figure Change Implementation Document ABN 3292-1 Safety Evaluation No.: 87-0307 UFSAR Figure Change Implementation Document ABN 6340-1 Safety Evaluation No.: 87-0309 UFSAR Figure Change Implementation Document ABN 7605-1 Safety Evaluation No.: 87-0310 UFSAR Figure Change Implementation Document ABN 8033-1 Safety Evaluation No'.: 87-0313 Implementation Document ABN 8032-1 Safety Evaluation No.: 87-0316 .
Implementation Document ABN 6472-1 Safety Evaluation No.: 87-0317 UFSAR Figure Change Implementation Document ABN 7537-1 Safety Evaluation No.: 87-0318 UFSAR Figure Change Implementation Document ABN 6441-1 Safety Evaluation No.: 87-0323 UFSAR Figure Change Implementation Document ABN 6866-1 Safety Evaluation No.: 87-0331 UFSAR Figure Change Implementation Document ABN 5948-1 Safety Evaluation No.: 87-0339 UFSAR Figure Change Implementation Document ABN 6775-1 Safety Evaluation No.: 87-0342 UFSAR Figure Change Implementation Document ABN 7536-1 Safety Evaluation No.: 87-0354 UFSAR Figure Change Implementation Document ABN 7412-1 Safety Evaluation No.: 87-0357 UFSAR Figure Change l Implementation Document ABN 6707-1 Safety Evaluation No.: 87-0358 UFSAR Figure Change ' Implementation Document ABN 3535-1 ! l l
'Enclosura to s NRC-89-0044-Page 80-Safety Evaluation No.: 87-0359 UFSAR Figure Change Implementation Document ABN 4814-1 Safety Evaluation No.: 87-0360 .
UFSAR Figure Change Implementation Document ABN 7830-1 Safety Evaluation No.: 87-0361 UFSAR Figure Change -! Implementation Document ABN 4815-1 Safety Evaluation No.: 87-0362 UFSAR Figure Change Implementation Document ABN 6267-1 q Safety Evaluation No.: 87-0371 UFSAR Figure Change ; Implementation-Document ABN 7388-1 Safety Evaluation No.: UFSAR Figure Change 87-0372 Implementation Document ABN 7642-1 Safety Evaluation No.: 87-0376, 88-0159 UFSAR Figure Change Implementation Document ABN 7909-1 .! Safety Evaluation No.: 87-0379 UFSAR Figure Change ; Implementation Document ABN 6722-1 Safety Evaluation No.: 87-0383 UFSAR Figure Change Implementation Document ABN 8046-1 Safety Evaluation No.: 88-0002 UFSAR Figure Change l Implementation Document ABN 8145-1 l Safety Evaluation No.: 88-0023 UFSAR Figure Change-Implementation Document ABN 8072-1 i Safety Evaluation No.: 88-0051 UFSAR Figure Change ! Implementation Document ABN 8596-1 Safety Evaluation No.: 88-0071 UFSAR Figure Change Implementation Document ABN 8393-1 Safety Evaluation No.: 88-0076 UFSAR Figure Change Implementation Document ABN 8605-1 Safety Evaluation No.: 88-0081 UFSAR Figure Change Implementation Document ABN 8737-1 Safety Evaluation No.: 88-0088 UFSAR Figure Change Implementation Document ABN 8070-1 Safety Evaluation No.: 88-0100 UFSAR Figure Change l l Implementation Document ABN 8493-1 l _ _ _ _ _ . _ _ _ _ _ _ _ _ ________ _ __-__ _ _ _ -___-_ 2
. Enclosure to NRC-89-0044 sPage 81 j
;i Safety Evaluation No.:. 88-0106 UFSAR Figure Change ,
Implementation Document ABN 8347-1 Safety Evaluation No.: 88-0111 UFSAR Figure Change i Implementation Document' ABN 8936-1 3 Safety Evaluation No.: 88-0115 UFSAR Figure Change
-Implementation Document ABN 8829-1 Safety Evaluation No.: 88-0116- UFSAR Figure Change Implementation Document ABN 8551-1 Safety Evaluation No.: 88-0117 UFSAR Figure Change Implementation Document ABN 8794-1 Safety Evaluation No.: 88-0121 'UFSAR Figure Change Implementation Document. ABN 6815-1
}
Safety Evaluation No.: 88-0125 UFSAR Figure Change Implementation Document ABN 9061-1 ]; Safety Evaluation No.:- 88-0132 Implementation Document ABN 9191-1 Safety Evaluation No.: 88-0141 UFSAR Figure Change Implementation Document ABN 6090-1 4 Safety Evaluation No.: 88-0166 UFSAR Figure Change Implementation Document ABN 9031-1 l j Safety Evaluation No.: 88-0174 UFSAR Figure Change l Implementation Document ABN.9393-1 -l Safety Evaluation No.: 88-0176 UFSAR Figure Change f Implementation' Document ABN 8424-1 l Safety Evaluation No.: 88-0204 UFSAR Figure Change Implementation Document ABN 8413-1 Safety Evaluation No.: 88-0207 UFSAR Figure Change Implementation Document ABN 9516-1 Safety Evaluation No.: 88-0211 UFSAR Figure Change 1 I Implementation Document ABN 7711-1 Safety Evaluation No.: 88-0212 UFSAR Figure Change Implementation Document ABN 9227-1 Safety Evaluation No.: 88-0214 UFSAR Figure Change Implementation Document ABN 9354-1 4 --___m_-_-..m_ _ . _ . . _ _ . _ _ _ . _ _ _ - _ . _
F- i l i Enclosuro to. l NRC-89-0044 Page 82 , Safety Evaluation No.: 88-0221 UFSAR Figure Change Implementation Document' ABN 6514-1 Safety Evaluation No.: 88-0222 UFSAR Figure Obange . Implementation Document ABN 4868-1 j Safety Evaluation No.: 88-0225 UFSAR Figure Change- -j Implementation Document ABN 9579-1 l l Safety Evaluation No.: 88-0226 UFSAR Figure Change ') Implementation Document ABN 6819-1 { l Safety Evaluation No.: 88-0230 UFSAR Figure Change j Implementation Document ABN 9082-1 ! Safety Evaluation No.: 88-0234 UFSAR Figure Change j Implementation Document ABN 7715-1
-t
' Safety Evaluation No.: 88-0237 UFSAR Figure Change j 1
Implementation Document ABN 6453-1 Safety Evaluation No.: 88-0244 UFSAR Figure Change Implementation Document ABN 9907-1 i UFSAR Figure Change j Safety Evaluation No.: 89-0004 1 Implementation Document ABN 9155-1 Safety Evaluation No.: 89-0006 UFSAR Figure Change 4 3 Implementation Document ABN 10023-1 i Safety Evaluation No.: 89-0009 UFSAR Figure Change i Implementation Document ABN 8289-1 Safety Evaluation No.: 89-0012 UFSAR Figure Change Implementation Document ABN 9941-1 Safety Evaluation No.: 89-0018 UFSAR Figure Change l Implementation Document ABN 9142-1 Safety Evaluation No.: 89-0026 UFSAR Figure Change Implementation Document ABN 7082-1 b
-.-,------___._,______,___=_3
i. Enclosur*e to NRC-89-0044 Page 83 1
. SECTION 2 PROCEDURES, TESTS AND EXPERIMENTS 4
0 .
Enclorurs to NRC-89-0044
'Page 84- ,
-)
i There wer'e no procedure changes, tests.or experiments during 1988 which introduced an unreviewed safety' question. .The Technical Specification change.for the Primary Containment' Average Temperature
'(Safety Evaluation Nos. 86-0167 and 88-0210) did however require NRC approval prior to' implementation. The following is a summary of the procedures, tests and experiments.which were reviewed for any safety significance. ,
i l I Safety Evaluation No.: 87-0215 UFSAR Text Change- ] Procedure, Test or Experiment No.: DER 87-0271 i
%==acy:
Modification of the UFSAR to allow for flexibility in scheduling for-fire brigade training. The change will allow for up to a 25% , extension of the specified time, but the combined time interval for any three (3) consecutive intervals shall not exceed 3 25 times the. specified time interval. The fire brigade training at-Fermi meets or-
" exceeds all training requirements and guidelines. This change is consistent with the provisions of the Technical Specifications.
Safety Evaluation No.: 87 0258 UFSAR Text Change Procedure, Test or Experiment No.: N/A Snamary: Revising the instrument numbers for the Hydrogen and Oxygen sensors listed in the UFSAR will indicate the as-built condition of the plant. The Division I and II sensors were provided by Exo-Sensor and physically installed in the plant under EDP-1422. Therefore, correcting the sensors listed in the UFSAR (paper change.only) will' not affect the safety of the plant. Safety Evaluation No.: 87-0351 UFSAR Text Change Procedure, Test or Experiment No.: 23 701.11 m==ary: Change deleted reference to and thus removed the requirement for having Reactor Water Cleanup filter demineralized sludge hold-up for 60 days in the RWCU phase separator to allow for decay of short lived
~
j
Enclosura to NRC-89-0044 Page 85 radionuclides for ALARA reasons. .This change does not pose an unreviewed safety question since no physical changes to the facility were made. The change affects the resin. wastes in the RWCU phase separators and their transfer and processing by the Radwaste system
. sooner than previously evaluated.
l Safety Evaluation No: 88 0001 Procedure, Test or Experiment No.: STUT.06B.033 Summary: During performance of the startup procedure STUT.06B.033, a Level 1 acceptance criteria violation for Hain Steam transient vibration, loop pipe sensors F006 and D017 was documented. Subsequent analysis of the results by Sargent & Lundy determined that the measured piping respense by the two transducers was acceptable. Sargent & Lundy calculated revised Level I and Level 2 limits for the two sensors; the basis for these limits was documented in Sargent & Lundy calculation number EMD-064476. i Safety Evaluation No: 88-0008 Procedure, Test or Experiment No.: EF2-71531 Summary: This safety evaluation was performed to determine the consequences of plant operation with Recirculation Pump Seal Purge Line valves B2100F090A & B fully open. Original design intent was to set these two valves in a throttle position to limit break flow (postulated double ended break outside containment). The basis for limiting flow ' was to assure that the consequences of a break in these lines were bounded by an instrument break analysis. However, these-lines are 3/4" diameter process lines and are therefore exempt from break analysis relative to system function. 1
=
Enclo uro to l NRC-89-0044 l Page 86
-l 1
Safety Evaluation No: 88-0012 l 1 Procedure, Test or Experiment No.: STUT.06A.033 Anamary: l 1 During power ascension, it was observed at 85% power that the Level I l' vibration criteria for two Main Steam steady state vibration, loop pipe sensors (D-015 and D-016)_were exceeded. As a result of detailed stress analysis (required by the UFSAR) of the Main Steam piping, it was determined that the maximum neasured values were acceptable and below the new revised Level 1 acceptance criteria. The previous acceptance values were based upon a more conservative and simplified 4 analysis. Safety Evaluation No: 88-0027 Procedure, Test or Experiment No.: N/A { Snamary: A safety evaluation was performed to evaluate the possible options for implementing low-power main steam flow control by partial-arc _ steam admission through the Turbine Control Valves (TCV). This would reduce vibration in the Main Steam System and minimize damage that occurred previously from extended low power operation. A detailed review indicated that flow control of main steam admission to the high-pressure turbine by the TCV partial-arc opening method is achievable. Safety, Evaluation No: 88-0030 Procedure, Test or Experiment No.: N/A i Snamary: Placement of corrosion coupons into the RHR reservoir such that the corrosion rate of the system can be monitored to meet General Electric's corrosion rate specifications. This temporary method of monitoring will be evaluated to determine future, permanent method.
~
Enclosura to H i NRC-89-0044 Page 87 l l Safety Evaluation No: 88-0043 Procedure, Test or Experiment No.: Procedures 23 208, 24.205.5, 24.205.6 Summary: These procedures for the operation and testing of the RHR Service Water systems were changed such that the Division 1 and 2 flow control valves (E1150F068A and B) would be given an "open" signal up to 5 seconds before the respective pumps are started. This was necessary since the valves would not open if.the pumps were started first. Safety Evaluation No:. 88 047 Procedure, Test or Experiment No.: SOE B3100 88-02 23 138.07 Sn==ary: Safety evaluation was written to ensure that there would be no safety significance when Reactor Recirculation MG Set A & B speeds are increased in 1% increments from 51% to 56% while running GETARS traces and process computer traces to determine the effects of changes in recirculation speed on Linear Heat Generation Rate (LHGR) and on Pellet Clad Interaction Operating Margin Requirement (PCIOMR) implementation. Safety Evaluation No: 88-0048 Procedure, Test or Experiment No.: DER 88-0095 Summary: The intent of this safety evaluation was to evaluate conditions created when utilizing " Star _ Lug" type termination lugs on safety _related equipment. Identified in DER 88-0095, the extent of adequate considerations and lack of OSRO approvals are in question for the design change document (EDP 1706) which allowed the use of these termination lugs. This safety evaluation is in support of the design verification and the original preliminary safety evaluation. l l f L1______
1 Enclo:ure to NRC-89-0044 l Page 88 Safety Evaluation No: 88-0049 I Procedure, Test or Experiment No.: SOE R1102-88-01, R1102-88_02 l I Summary: These two SOE procedures performed logic system functional tests of the " Loss of Power" undervoltage relay schemes on Buses 64B and 64C as required by Technical Specification Surveillance requirements. The SOEs will verify the logic for the unde'rvoltage schemes from the undervoltage relay through the actuated device. Safety Evaluation No: 88-0054 Procedure, Test or Experiment No.: EPA-99-B-SQ Sn==ary: The revision to EPA-99_B-SQ was made to incorporate organizational changes in the Nuclear Organization that, in accordance with 10CFR50, Appendix E, and NUREG-0737 Supplement 1, must be reflected in the Emergency Plan. The change is administrative in nature. Safety Evaluation No.: 88-0064 UFSAR Text Change Procedure, Test or Experiment No.: FMR 7428, FDDR KH1-1074, NEDE 22109. Snamary: Revision 2 of the UFSAR will update the sections that describe operation of the HPCI and RCIC turbines barometric condenser equipment when the system is automatically initiated as required. The initiation logic was modified by FMR 7428 in 1984, with subsequent revision to the FSAR to reflect this modification, but the two sections mentioned above were inadvertently omitted from the revision. The safety evaluation reviewed the changes and determined that the changes were enveloped by the existing design basis.
Enclorure to
;NRC-89-0044 Page 89 Safety Evaluation No: 88-096 Procedure, Test or, Experiment No.: DER 88-0863 Summary:
Acceptability of Available-Stem Thrust "MOVATS" testing was completed-on Reactor Core Injection Cooling motor-operated valve (HOV) E5150F022 with final available thrust less than target value specified by PDC j 8505 Per the PDC, this safety evaluation was written to accept the ] performance of the MOV for continued operations. l 1 1 ________________________________________._____________________________ j Safety Evaluation No: 88-0099 Procedure, Test or Experiment No.: N/A Summary: Safety evaluation takes credit for Division 2 of Residual Heat Removal (RHR)' shutdown cooling as an operable alternate decay heat removal 1 method when Div. 2 shutdown cooling is declared inoperable due to the l inoperability of one RHR mechanical draft cooling tower fan. ! i Safety Evaluation No.: 88-0118 UFSAR Text Change Procedure, Test or Experiment No.: N/A Summary: The Reactor Building Ventilation Radiation Monitor write up as reflected in subsection "M" of UFSAR Section 7 3 2.2.8 is not correct. It is less conservative than the plant license basis as reflected in other UFSAR write ups on the same subject and the plant j schematics. The proposed write up is more conservative because the. four fuel pool ventilation detectors presently written in the UFSAR are not the only detectors that provide the isolation trip. In addition, there are two more detectors located on the Reactor Building ventilation exhaust. These two detectors look at the exhaust from the Reactor Building which also includes the exhaust from the fuel pool. i The proposed write up is the licensing basis and basis of the plant accident analysis. By correcting the mistake as it exists in the UFSAR the Chapter is brought to conformance with the plant's accident analysis as reflected in Chapter 15 and the basis of the plant. , i l
Enclo2ure to NRC-89-0044 Page 90 Safety Evaluation No.: 88-0120 UFSAR Text Change Procedure, Test or Experiment No.: STUT.06B.027 Summary: In compliance with the UFSAR and the internal test reduction program, credit was taken for an inadvertent turbine / generator trip (12/31/87) from Test Condition,6 for the required start',ip test Turbine Stop Valve and Generator Load Rejection - Generator Load Reject (which was to be
' performed deliberately). Per NRC commitment #7819 a safety evaluation was required to evaluate the event.
Safety Evaluation No: 88-0133 Procedure, Test or Experiment No.: DER 88 1198 Summary: Emergency Diesel Generators (EDG) 12 and 13 Jacket Coolant Heater Pressure Relief valves R30F018B and R30F018C are "Kunkle Model 6000". The origin of these two Valves was indeterminate (i.e., not an approved Colt replacement part, no purchasa information). The Kunkle 6000 is similar to the approved replacement part (Kunkle 84 4) in size, weight and operation. The EDGs reliability is not compromised by using these substitute relief valves. Safety Evaluation No: 88-0150 Procedure, Test or Experiment No.: DER 88-1389 Sunnary: This safety evaluation was performed to evaluate the potential impact on the Emergency Diesel Generators (EDG) due to the possible non-shed loads being applied to the EDGs during the load sequencing process following a Loss of Offsite Power (LOSP) and Loss of Coolant Accident (LOCA) as well as potential long time operation of the EDG with the added loads. The objective of this evaluation is to determine that, although the surveillance were not completed as required, the EDGs would have performed as required should a LOCA and LOSP have occurred. l l l l b l
< l f Enclorura to NRC-89-0044 Page 91-Scsfety Evaluation No: 88-0151, 88-0152, 88 0156 Procedure, Test or Experiment No.: SOE R1102-88-03, R1102_88-04, ,
91102-88-05 h =ary-I These sequence of events were written to verify the operation of the ) Division I ar.d II 480 volt ESS busses to meet the requirements of.the- -{ Technical Specifications which require that for a Loss of Offsite Power (LOSP) the loads on the bus will shed to allow correct load sequencing back to the EDGs. Safety Evaluation No: 88-0153 Procedure, Test or Experiment No.: RERP Plan Summary: The RERP Plan was revised to comply with Fermi Management Directives. The changes included, but were not limited to, the consolidation of 23 procedures into one plan, the reformatting of the plan to account for the consolidation, the revision and upgrading of several diagrams for clarity, and the incorporation of human factors principles. .All changes made were determined to be administrative in nature.
]
1 Safety Eva.luation No: 88-0155 Procedure, Test or Experiment No.: 29.000.01, 29 000.02, 29.000.03
.h w y: f I
The Emergency Operating Procedurec (EOP) were revised to reflect changes made in the BWROG Emergency Procedure Guidelines (Revision 4) and to include human factors consideration. These procedural revisions did not involve any physical changes to the facility nor any significant changes to normal operating procedures. l l I
. . i
Enclorure to
.NRC-89-004H Page 92 )
Safety Evaluation No:. 88-0163 l Procedure, Test or Experiment No.: STUT.06B.017 m-macy: The safety evaluation was written to evaluate the Level 1 test ; acceptance criteria for displacement sensor D203 located on the RCIC l steam supply line and force sensor F004A located on the Main Steam . line outside containment. Analysis of the results by Sargent & Lundy determined that the measured piping response by the two transducers was acceptable. , Safety Evaluation No: 88-0169 Procedure, Test or Experiment No.: 24.000.05 mamary: Safety Evaluation was performed to evaluate change to procedure 24.000.05, Revision 20 (Monthly Continuity Light and Channel Check Ltest) which provides an alternate means for verifying continuity of the~ Standby Liquid Control-(SLC) squib valves. Safety Evaluation No.: 88-0186 UFSAR Text Change Figure Change Procedure, Test or Experiment No.: DER 88-1125, 88-1349 Sammary: At present the radwaste system is not operating in accordance with the UFSAR. This is because 1) the evaporators are not in use, 2) the asphalt-extrudec system is not in service, and 3) precoat filters are presently used in place of the etched-disc / oil coalescer trains. An extensive UFSAR update is to be performed. It should be noted that the UFSAR already specifically addresses the fact that the subject subsystems may be taken out of service for periods of time, and the overall system was designed to handle such situations (permanent bypasses and interconnections exist,, as well as permanent piping and valving to the Onsite Storage Facility (OSSF) for vendor processing). It also clearly addresses the operation of the precoat filters, and of vendor solidification in the OSSF. Taking any of these subsystems out-of-service on a more permanent basis would not cause an unreviewed safety question.
^
Enclorura to NRC-89-0044 Page 93 1 Safety Evaluation No.: 88-0187 UFSAR Text Change l Procedure, Test or Experiment No.: DER 88-0164 h==ary: j The wording of the UFSAR is to be changed to permit channel calibration during periods of extended shutdown by using a shaker j table to generate precise acceleration magnitudes at specific test l frequencies-or by impacting the system and indirectly exciting the l sensor using a calibration' impact hammer at defined reference i locations and comparing the output of the channels to initial . ( benchmark data. Both methods of verifying the calibration of the j Loose Parts Monitoring System (LPMS) produce equivalent results. I i Safety' Evaluation No.: 88-0188 UFSAR Text Change Procedure, Test or Experiment No.: CECO for component N62N018 Summary: The offgas hydrogen concentration alarm setpoint.specified in the UFSAR will be revised from 3 9% to -1.5%. This change will bring the UFSAR into agreement with the actual plant design setpoint. The design alarm setpoint change was made in April, 1983 by DCN 9216, Rev. O. The change from 3 9% to 1.5% was made to provide additional time for operator response without increasing nuisance alarms. The design change is reflected in plant alarm response procedure 6D15 Since this change was made in the conservative direction, it would not pose an unreviewed safety question. Safety Evaluation No.: 88-0192 UFSAR Text Chrnge Procedure, Test or Experiment No.: DER 1184 humary: The Technical Specifications and the UFSAR require the effective multiplication factor, k.eff, of fuel in storage to be 0.94 or less. The direct calculation of this number requires special procedures, a special programs, and details of the storage rack geometry. This is not part of the fuel vendor's normal scope of work. Instead, the vendors and utilities are adopting an equivalent but indirect method to calculate k-eff of fuel in storage which uses standard methods and geometry. The UFSAR is mute on the qualification of fuel for storage thereby implying the direct method will be used. Based on GE's l l I
. -J
1
' Enc 1ccurs to NRC-89-0044 Page 94 recommendation, the UFSAR will be revised to document that the indirect method is the basis being used to qualify fuel for storage.
No change is required to the Technical Specifications. No unreviewed safety question exists. Safety Evaluation No.: 88-0197 UFSAR Text Change Procedure, Test or Experiment No.: N/A Rumancy: _ J Figure 8.2_6 is a Fermi I drawing, 6E721F-3, that describes the 120KV bus and circuit breakers. This is not a Base Configuration Design Document (BCDD) for Fermi 2. Figure 8.2 6 is referred to in two places in the UFSAR text and in one'UFSAR Table. Deletion of this figure is an editorial modification and does not create an actual change or modification to the Fermi 2 plant or procedures. Safety Evaluation No.: 88-0199 UFSAR Text Change Procedure, Test or Experiment No.: NE-5.6 Snamary: The internal reporting requirements for the Leakage Reduction Program was changed from making the inspection and test results available to plant operators to issuing a management summary report on effectiveness within 90 days of the conclusion of each reactor refueling outage. This permits planc-management to focus attention on the content of the report as appropriate. The change will be reflected in Revision 2 of the UFSAR. This change does not represent any unreviewed safety questions. l 1 f
Enslosura'to-NRC-89-0044-Page 95- ; I t Safety Evaluation No.: 88-0205 UFSAR Text Change Procedure, Test or Experiment No.: DER 88-1053 Summary: 1 Clarification of the Control Air system will be provided in Revision 2 of the UFSAR. These updates include the following: . 7.6.1.17.4.1 - Control air systen equipment. design initiating circuits - revised to: include NIAS isolation and compressor auto start- ! on loss of offsite power. Also, compressor auto start on Level 2 LOCA signal. 9.3.1.2 - Process auxiliaries compressed air systen. description updated identification of systems using NIAS. The operation of the station air compressors, control air compressors and NIAS systen , isolation features were also clarified. 15.6.1.1 Loss of instrument air, starting condition and assumptions- 4 were revised to indicate. that more than 1 isolation valve is open. I 15.16.1.2 - Loss of instrument air event' description - clarified l events if the second station air compressor is not adequate to ! maintain system pressure. 15.16.2.1 - Loss of instrument air - EFFECTS - revised scenario of I effects on loss of station air supply to the NIAS' system and IAS ! system. 15.16.2.2 - Loss of instrument air - ANALYSIS - clarified that the i noninterruptible air system is capable of supplying equipment i requiring instrument air to perform its safety function. Since these changes were made for clarification only, they pose no unreviewed safety questions. Safety Evalnation No.: 88-0206 Procedure,. Test or Experiment No.: DER'88-1063 Summary: The basement wall between the Turbine Building and the Auxiliary Building along column line H between columns 12 and 13, elevation 564'.0" up to 583'-6", contains 32 penetration seals. The penetration schedule drawing did not indicate any of these as " water-tight". t
- k
l Enclosure to NRC-89-0044 Page 96 After extensive evaluation it was concluded that the "as-found" conditions of the penetration seals would not cause a safety challenge to the plant safe shutdown systems and components under a postulated site flooding condition. Safety Evaluation No: 88 0208 UFSAR Text Change Procedure, Test or Experiment No.: STUT.06A.003 Summary: The safety evaluation evaluates the elimination of the steady state vibration testing for the RHR Hoad Spray piping inside containment during normal Head Spray operation. In earlier testing, it was discovered that it was not necessary to use Head Spray for its intended function or any other safety functions, so administrative controls were provided to prevent its use. Safety Evaluation No: 88 0209 UFSAR Text Change Procedure, Test or Experiment No.: STUT.06B.033 Sn==ncy: The safety evaluation reviews the modification of the Startup Test program which deferred the Turbine Stop and Control Valve Closure Transient vibration test for the main steam piping outside containment from TC6 until an inadvertent turbine trip from appropriate reactor power level. Earlier test results had been overridden by outside
" noise" by a failed sensor. This testing was deferred to avoid unnecessarily subjecting the plant to a transient event.
.5, f' Enclo2ura.to L 'NRC-89-0044 L ' Page ; 97 -- _
_l l si , j j
. Safety Evaluation No.: 88-0210, 86-0167 UFSAR Text Change Implementation Document No.: N/A System No.: T5700
' Title of Change: Increase in Technical Specification Primary Containment average temperature. q Suminary:
3 Increase Technical Specification Primary Containment average temperature from 135"F to'145 F. Safety Evaluation Summary: The change.does not involve a physical modification to the plant or.a L change-in operating practices. The chan e does involve a change in the limiting conditions for operation which has been evaluated against environmental qualification requirements, drywell concrete ~ design requirements, piping, hanger and support requirements, safety-related..
-equipment in the drywell.and the bounding safety; analysis accident (LOCA).- 'By' increasing the limit for drywell average air temperature during normal operatioiis,. there is no impact on the. dynamic loads' or' the containment' response during a LOCA. The analysis indicates that the peak ~drywell pressure, peak between pressure and drywell pressurization rate actually decreases with the increase in drywell temperature.
Safety Evaluation No.: 88-0216 UFSAR Text Change Figure Change Procedure, Test or Experiment No.: G.E. NEDC-31515 h==a ry : Expand power flow map to allow operation up to 102% rod line capped at 100% power. 4 A safety evaluation was-performed to determine if operation to the ' 102% rod line subject to feedwater temperature and core inlet subcooling constraints created any unreviewed safety questions. All core-wide transients described in the Fermi 2 UFSAR Chapter 15 were examined for operation in both the Maximum Extended Load Line Limit Region (MELLLR) and the normal operating power / flow map. GE evaluations were performed at two power / flow operating points (102% power /75% flow and 102% power /100% flow. s
, -)
Enclosure to NRC-89-0044 Page 98 l l After extensive review of this issue, Detroit Edison concluded that operation to the 102% rod line with the specified restrictions on feedwater heating and core inlet subcooling does not result in an unreviewed safety question. Safety Evaluation No.: 88-0227 UFSAR Text Change l Procedure, Test or Experiment No.: N/A Smma7y: The primary content of these UFSAR changes is in an update in the descriptions of Dry Active Waste (DAW) operations (collecting, sorting, compacting, storing), in particular, the fact that we are now using a small portion of the OSSF (Onsite Storage Facility) for DAW short-term st.orage and sorting. Also, some minor changes are made to words or phases in order to clarify or make operations less restrictive. Thirdly, a few minor word changes are made to reflect full-time vendor operations in the OSSF, versus asphalt system. These changes in DAW operations have no, direct interfaces with other plant systems or nearby pieces of equipment which could cause malfunction of safety-related equipment. Safety Evaluation No.: 88-0229 Procedure, Test or Experiment No.: Technical Specification Improvement Program Summary: The present statement in the UFSAR, i.e., "The Figure 7.2 2 shows all the rod block functions" appears to be an editorial error that had gone unnoticed. This figure does not show, nor is there any other figure in the UFSAR that shows all rod block functions. Instead, all of the rod block functions are described in detailed prose in various UFSAR sections. In order to remove any interpretive misnomers that can be created, editorial changes have been made to the existing statements to more clearly reflect the design of the rod block trip channels. The plant design, hardware or the transient and accident analysis are not impacted. On the contrary, the new wordings reflect more clearly the existing plant design and accident analysis bases.
IL I EncloIura to
,4 KRC_89_0044 Page 99
)
4 Safety Evaluation No.: 88 0231 UFSAR Text Change Procedure, Test or Experiment No.: DER 88-033 Summary: Safety evaluation specifies the intent'of' performing breaker operating and protective relay testing. Present wording in the UFSAR indicate that breakers and relays will be tested initially,Eone yearL thereafter, and then coincident with reactor. shutdown. The proposed { changes state the CAIO test were initially completed, then ) Pre operational or Acceptance tests and protective relay tests prior
'to fuel load; subsequent tests are performed in accordance.with the preventive maintenance and surveillance programs. . Additionally, some $
minor wording changes are made and a partial list'of test instructions j is deleted. The description of.the CAIO and Pre-operational test-instructions was made past tense. -Since these changes are explanatory and editorial in nature, they pose no unreviewed safety concerns. Safety Evaluation No.: 88-0243 UFSAR Text Change Figure Change f Procedure, Test or Experiment No.:'RERP Plan mamary: The UFSAR is being revised to reflect the current status of the RERP Plan, Rev. 0 (July, 1988). All changes are being made to reflect current company names, update data location and retrieval methods, provide consistency with definitions in the~RERP Plan,. changes in location of personnel from the NOC to other areas of the site, and changes in the seating configurations of the TSC and EOF to be consistent with the RERP Organization. Because these changes are for clarification and consistency with the RERP Plan, it~can be concluded that no unreviewed safety questions exist. Safety Evaluation No.: 89-0044 UFSAR Text Change Figure Change Procedure, Test or Experiment No.: N/A m==a ry : The proposed' changes to UFSAR Sections 13 1 and 13 5 are a result of major organization changes in the Detroit Edison Corporate structure and in the Fermi 2 plant organization. These proposed changes do not result in an unreviewed safety question because they are I
l Enclorura to
'NRC-89-0044 Page 100 organizational changes that do not affect the operating license, license certification requirements minimum qualification requirements.
for Fermi 2 plant personnel as deflued in Reg. Ouide 1.8 and Section 4 of. ANSI-N18.1 (1971), or Technical Specifications. Safety Evaluation No.: 89-0045 UFSAR Text Change Procedure, Test or Experiment No.: -QA Audit 88-0208 Summary: QA Audit 88-0208, Observation 9, stated that UFSAR Section 5.2.8.7 , l referenced the wrong code addenda; Winter, 1981 instead of Winter, 1980. Both Sections 5.2.8.1 and 5.2.8.7 contained a 40 month review interval which was not applicable to Fermi 2. A change was made to
'the addenda year and the. statement concerning the 40 month interval was deleted. -The changes'are strictly editorial in nature and provide consistency with other references in the affected sections. j l
I
- i i
i 1
- !}}