NG-98-1767, Cyclic Rept of Facility Changes,Tests & Experiments,Fire Plan & Commitment Changes,For Period from 970301-980930. with
| ML20195H575 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 09/30/1998 |
| From: | Franz J IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT |
| To: | Collins S NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 NG-98-1767, NUDOCS 9811240041 | |
| Download: ML20195H575 (208) | |
Text
b ALLI ANT g UTluTIES its cuiine,inc.
Duane Arnold Energy Center IES Utilities 32n orte noaa Palo. I A 52324-9785 Ofhce: 319.8513611 November 19,1998 3 FuL ;,851 , ;)8l
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NG-98-1767 Mr. Samuel Collins, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station 0-PI-17 l Washington, DC 20555-0001 I
Subject:
Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49 Cyclic Report of Facility Changes, Tests and Experiments, Fire Plan ;
Changes and Commitment Changes l File: A-118e
Dear Mr. Collins:
In accordaace with the requirements of 10 CFR Section 50.59(b), and NUREG-0737 (Item II.K.3.3), please find enclosed the subject report covering the period from March 1, '
1997 through September 30,1998. A summary of changes to the Duane Amold Energy Center Fire Plan during the same time period is included, as well as a summary of various commitment changes. There are no new commitments made in this letter.
Should you have any questions regarding this matter, please contact this office.
Sincerely, , 1 John F. Franz l Vice President, Nuclear Attachment b, N\
9811240041 90093031 PDR ADOCK O R
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Mr. Samuel Collins NG-98-1767 '
November 19,1998
- Page 2 cc: D. Barta E. Protsch R. Laufer (NRC-NRR)
James Caldwell(Region III)
NRC Resident Office D. Wilson DOCU l
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Attachment to NG-98-1767 )
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November 19,1998 Cyclic Report of Facility Changes, Tests and Experiments, Fire Plan Changes and Commitment Changes Table of Contents Section' A - Plant Design Changes Pages I through 102 Section B - Procedure / Miscellaneous Changes Pages 103 through 198 Section C - Experiments Page 199 Section D - Fire Plan Changes Page 200 through 202 Section E - Commitment Changes Pages 203 through 205 i
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Section A - Plant Design Changes This section contains brief descriptions of plant design changes completed during the l
period of March 1,1997 through September 30,1998, and summaries of the safety evaluations for those changes, pursuant to the requirements of 10 CFR Section 50.59(b).
All changes were reviewed against 10 CFR 50.59 by the Duane Arnold Energy Center (DAEC) Operations Committee. None of the changes involved unreviewed safety questions.
The basis for inclusion of an Engineering Change Package (ECP), or Plant Modification Package (PMP) in this report is operational release of the associated modification at the DAEC during the period of March 1,1997 through September 30,1998. The basis for inclusion of an Engineered Maintenance Action (EMA) is completion of all the changes described in the Safety Evaluation, during the period of March 1,1997 through September 30,1998. Portions of some of the modifications listed were partially closed or partially operationally released in previous years.
SE 95-10 (Revision 1) PMP 100 - Increased Ilydrogen Injection 1
Description and Basis for Chance '
The IIydrogen Water Chemistry (HWC) System was originally designed I to supply the necessary quantity of feedwater hydrogen to reduce the electrochemical corrosion potential (ECP) in the stainless steel Reactor l Recirculation System piping. This would arrest crack growth and preclude I new cracks from initiating due to intergranular stress corrosion cracking (IGSCC). An ECP value less than -230 mV SHE (standard hydrogen electrode) will arrest cracking in austenitic stainless steels. A Special Test was performed in 1989 to determine the optimum hydrogen injection rate to protect reactor vessel internals. It was determined that a feedwater hydrogen injection rate of 25 scfm would achieve an ECP of-230 mV SHE at the lower core plate. Injection at this rate will protect vessel intemal components in the downcomer annulus and bottom head regions.
The original intent of this modification was to increase the HWC hydrogen injection rate to 25 scfm to protect vessel internals. Based on the implementation of Noble Metal Chemical Addition (NMCA) during RF014, protection of reactor vessel internals is now maintained with a nominal hydrogen injection rate of 6 scfm. Subsequently, this modification was changed to retum the system, to an operating range of 0-10 scfm hydrogen injection (0-5 scfm oxygen).
From 4/8/96 to 10/10/96 IIWC was operated at a hydrogen injection rate of 15 scfm. Prior to this, the injection rate was approximately 9 scfm.
MSL Radiation Monitor setpoints were increased to accommodate this higher rate; new setpoints were calculated and specified by an approved I
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calculation. A study was conducted to evaluate the radiological effects of this injection rate, and determine compliance with 10 CFR 20 and 40 CFR 190. The study concluded that the annual dose to an individual in the Badging Center (most limiting for 10 CFR 20) would be less than the limit of 100 mrem /yr. The annual dose at the site boundary was calculated to be 21.9 mrem, which is less than the limit of 25 mrem prescribed by 40 CFR 190.
This modification included the following:
. Hydrogen System Modifications - Excess flow check valve internals were replaced to increase their isolation setpoints.
. Offgas Oxygen Supply System Modifications - The one-halfinch piping from the liquid oxygen tank to the flow control station (approximately 350 feet) was replaced with one inch pipe to reduce flow restriction through the system. The outlet piping from the oxygen flow control station was re-routed to a new injection point downstream of the Offgas Jet Compressor on an existing vent line. This modification completed the oxygen piping reroute within the Turbine Building and the tie-ir at the flow control station. The associated flow I indicating transmitters were replaced with new dual range units. The excess flow check valve internals were replaced to increase its isolation setpoint.
Safety Evaluation Summary None of the systems directly related to the implementation of this modification are identified as the causes of accidents which have been evaluated in the Updated Final Safety Analysis Report (UFSAR) and Nuclear Safety Operational Analysis (NSOA). Therefore, this activity did not increase the probability ofoccurrence of an accident evaluated previously in the SAR. Changes to the Ilydrogen and Oxygen Storage and Transfer Systems were performed in accordance with accepted codes, standards, and practices. Industrial safety practices and procedures were followed in the design, construction and operation of the hydrogen and oxygen systems.
The IIWC System and the systems it ties into are not relied upon to mitigate the consequences of any accidents evaluated in the UFSAR or NSOA. The consequences of an accident evaluated previously in the SAR are not increased. The equipment modified, as well as interfacing systems, do not have any safety significance, therefore, the probability of occurrence of a malfunction of equipment important to safety is not 2
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increased. None of the equipment directly affected by this modification is l relied upon to mitigate the consequences of any malfunctioning safety related components. The modifications to the HWC System present no new potential for accidents of a different type. Although the excess flow check valve setpoints have been increased, the system still provides isolation for pipe breaks. The electrical high flow trips provide additional redundancy. None of the changes involved safety significant equipment; therefore, no new failure modes are possible. The Ilydrogen Water Chemistry System is not a Technical Specification required system. None of the changes affected Technical Specifications, Technical Specification Bases, or established margins of safety.
SE 97-06 Condensate Pump Effluent Sampling Temporary Modification Description and Basis For Chance The purpose of this Temporary Modification was to install temporary connections between the Condensate Pumps' sample lines and the ,
Corrosion Product Sampler. This was needed to help isolate the source of I copper which was detected in the Condensate System.
The Corrosion Product Sampler was added to the Turbine Building Sample Station to track and trend particulate such as iron in the Feedwater System. This Temporary Modification disconnected sample line connections and tied-in the effluent of the Condensate Pumps' sample points to the existing tubing at these locations. Sample points for condensate to demineralizer sampling was unaffected, only the flow to the l Corrosion Product Sampler was interrupted. The temporary connection j was accomplished using tubing with adequate temperature and pressure '
rating for this service.
Safety Evaluation Summary This modification did not increase the probability of occurrence of an accident evaluated previously in the SAR since no accidents related to the Turbine Building Sample Station have been evaluated in the SAR. The probability of occurrence of a malfunction of equipment important to
- safety evaluated previously in the SAR was not increased since this change did not affect any equipment which has an important to safety
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function. Likewise, this activity did not increase the consequences of a !
malfunction of equipment important to safety evaluated previously in the SAR, and it did not create the possibility of an accident of a different type l than any evaluated previously in the SAR. This change did not affect any l equipment which has an important to safety function. There are no bases for any Technical Specifications associated with the affected sample lines.
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l SE 97-07 Temporary Modification To Drywell Cooling '
Description and Basis For Chance '
This safety evaluation was prepared to support the draining of a loop of Drywell cooling and the backflush of a Drywell cooler. To accomplish this, flow was interrupted to a loop of Drywell cooling which was drained, I then service air was introduced into the system, and a drain path was !
established to the well water return header. The changes made to the Well !
Water System and Drywell Cooling System were made to support maintenance on the system. The systems are not safety related. The well water piping inside the Drywell is seismic and is considered a closed loop system inside containment. The system drain activities were done at locations external to the containment isolation valves so the closed loop requirements were unaffected by the temporary modification. The backflush of the Drywell cooler briefly created a well water leak inside containment. This piping was considered a closed loop because the backwash was a manual operation with a person present for the entire time under administrative control, and the backwash took less than ten minutes.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The Drywell Cooling System and Well Water Systems are not accident initiator systems. The Well Water and Drywell Cooling Systems are not required to mitigate the consequences of an accident. They do remove a significant amount of heat from the containment, therefore, this activity was performed with the reactor subcritical to minimize the challenge to the containment.
This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. Isolating and draining a loop of Drywell cooling occurred at a time when the Drywell heat load was within the capability of one cooling loop.
Drywell temperature was maintained below the 135 degree F Technical Specification Limit. Therefore, equipment in the Drywell was not adversely affected and the operation of the equipment was not affected.
The Well Water System is not safety related, and its failure cannot cause an accident of a different type than any previously evaluated. The total loss of the Drywell Cooling System is a design feature of the plant, and this temporary r. odification did not cause anything that severe. The margin of safety as defined in the basis for any Technical Specification was not affected, because no safety systems were affected.
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SE 97-08 ECP 1585 - Fire Protection Modification l Description and Basis For Channe i This modification removed intervening combustibles between redundant Divisional raceways within Fire zone 16F in the Pumphouse by rerouting, I through conduit, cables that were in an open tray between the raceways, and portions of the open tray were removed. A new sprinkler system was installed below the 761' floor to provide suppression to the 747'-6" l elevation. This modification also removed Allison Convoi fire detection panels, and the portion of another panel associated with the 3 A Taermo-lag sensor wire. These panels provided detection for the tray protected with Thermo-lag material and were no longer requirs Safety Evaluation Summary The operation of the Fire Protection System does not affect any of the inputs considered in the accidents analyzed in the UFSAR or the NSOA.
The changes do not alter the interface between the Fire Protection System and the plant and cannot cause an accident or increase the likelihood of an accident. The probability of occurrence of the accidents discussed in the UFSAR and NSOA is based on initial conditions and assumptions which do not depend on the end use of or interactions with the Fire Protection Systems.
The activities involved with this modification meet the design, material, and construction standards as delineated in the NFPA codes. Based on the UFSAR this modification did not impact the operating performance of any of the other sprinkler areas. Therefore, this modification did not result in a condition which increases the probability of occurrence of an accident previously evaluated in the UFSAR.
There was no increase in the radiological consequences of any previously analyzed UFSAR accident. This modification did not change, degrade or prevent actions de cribed or assumed in an accident discussed in the UFSAR. This modification did not alter any assumptions previously made in evaluating the radio'sen! consequences of an accident, nor did it play a direct role in mitigatmg the radological consequences of an accident described in the UFSAR.
The Sprinkler System can affect the following Systems, Structures, or Components (SSCs) important to safety in the Pumphouse 747' basement elevation by inadvertent system initiation:
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The two River Water Supply (RWS) Discharge isolation valves and associated control devices I The River Water Radwaste Dilution Line isolation valve The Instrument Air Supply Root isolation Solenoid Valve and associated energizing circuitry If any of the above devices were to fait due to an initiation of the proposed Sprinkler System, the following would occur:
The RWS Discharge isolation valves would fail in the full open position. This action would place the RWS in an operating alignment to support the safe operating mode of the RWS System.
The River Water Radwaste Dilution Line isolation valve would fail in the closed position. This action would support the safe operating mode of Radwaste Dilution by preventing radwaste from discharging to the dilution structure.
All other safety related cables have been purchased as environmentally qualified. Thus cable failure due to wetting by sprinkler initiation is not credible. Additionally, a pressure decrease in the installed Sprinkler System will alarm in the control room permitting operator action within a short period of time. The sprinkler piping installed above any of the divisional raceways in the Pumphouse 747' basement elevation is supported seismically using standard seismic supports. This precludes the failure of any safety related equipment due to the structural failure of the Sprinkler System during or after a seismic event. The Fire Pump sizing and Fire Control System water supply are still adequate because the newly installed Sprinkler System is hydraulically bounded by the largest existing Sprinkler System plus 1000 GPM requirement. Additionally, the new system was installed using the proper material, construction standards and acceptance testing. New penetrations meet the requirements specified in the UFSAR and will not degrade the ability of the walls to function as fire barriers. These fire penetrations comply with Section Ill.G of Appendix R to 10 CFR 50, BTP APCSB 9.5-1, Appendix A, and ANI requirements.
New accidents of a different type than those of the UFSAR have not been created by this modification. This modification does not affect any equipment or systems important to safety in any way not previously evaluated. The addition of a Sprinkler System in Fire Zone 16F meets the same design and installation codes, and requirements as the existing systems and components that are part of the Fire Protection System except that the NFPA codes are the most recent editions. The Sprinkler System performs as evaluated in the UFSAR and will not affect the integrity of the 6
existing Fire Protection Systems. Therefore, no new failure modes are introduced and the probability of a malfunction of a different type has not been created. The requirements for Fire Protection Systems at the DAEC are not addressed in the Technical Specifications. The installation of the new Sprinkler System in Fire Zone 16F and the other activities performed by this modification replace the Thermo-lag barrier between the redundant safe shutdown trains. The activities involved complied with the requirements of the UFSAR as well as the DAEC Fire Plan, therefore, the margin of safety is not reduced.
SE 97-10 ECP 1584 - Condensate Demineralizer System Cam Replacement Descrintion and Basis For Chance The Condensate Demineralizer cams required frequent adjustment to provide proper timing sequences. This resulted in increased operating cost to maintenance and radwaste. An operator work around existed because of the numerous malfunctions in the " Auto" backwash and precoat cycles.
These malfunctions occurred routinely and operators were required to monitor the operation, and usually manually operate the backwash and precoat cycles. Subsequently, the cam unit on each filter unit and one main resin programmer were replaced with Programmable Logic Controllers (PLCs). The PLC units provide microswitch and timing sequence functions similar to the mechanical cam. Installation of PLC units and the review of the timing sequence for the backwash /precoat cycles resulted in several additional minor changes in the Condensate Demineralizer System. These changes consisted of the following:
. The three position influent Valve handswitch for each demineralizer was replaced with a two position switch which does not have an
" Auto" position, so the valve no longer opens automatically.
. The Bypass valve handswitch for each demineralizer was replaced with a three position switch. Previously, Operations could never close the bypass valve when the microswitch was closed on the cam unit.
Now, with the handswitch in " Auto", the new permissive from the PLC will control the valve. When the handswitch is taken to " Closed" or "Open", the PLC permissive is removed from the logic, allowing the control valve to function as the Operator desires.
. A time cday relay installed for each demineralizer, to keep the Main Drain Valve open to allow additional drainage of water to radwaste, was removed since the new PLC software facilitates opening of the Main Drain Valve without the time delay relay.
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e The Fill Relay contact for each demineralizer was removed. The fill relay is now energized by the PLC units. The PLC time sequence manipulates the Vent /Overfiow Isolation Valve for each of the demineralizers.
. Pressure switches for each demineralizer which stopped the demineralizer tank fill cycle were removed. The new PLCs control the fill relay via a timing sequence.
Safety Evaluation Summary The Condensate Demineralizer cam units replacement with PLCs will not prevent the Condensate Demineralizer System from performing its design function. The cam units replacement in the Condensate Demineralizer System cannot initiate any of the accidents described in the UFSAR or the NSOA. The Condensate Demineralizer System is not safety related.
Because this modification did not affect any system performance in a manner that could lead to an accident, the probability of an accident previously evaluated in the SAR is not increased. The Condensate Demineralizer System is not required for the mitigation of any accidents defined in the NSOA and UFSAR. Replacement of the cam units with PLCs did not prevent the Condensate Demineralizer System from performing its design basis function as defined in the SAR. Therefore, this ECP did not prevent or degrade any essential safety function assumed by the NSOA to mitigate the consequences of design basis accidents. This activity did not increase the consequences of an accident evaluated in the SAR. The Condensate Demineralizer System has not been defined as being important to safety. Single failure proof criteria does not apply to the Condensate Demineralizer System. The failure modes of the PLC have been evaluated for impact on the Condensate Demineralizer System.
The modification to the filter demineralizer backwash and precoat circuitry meet the original design specification installation criteria.
Installation of PLCs did not impact the Condensate Demineralizer System from performing its design function. System reliability has been enhanced wi h the installation of PLC units. Since the modified circuitry still meets the original design basis and specification requirements, the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No failure could be postulated by this
- modification that could create an accident of a different type. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The failure modes of the PLC have been evaluated for impact on the Condensate Demineralizer System. This 8
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modification did not introduce any new failure modes into the filter demineralizer circuitry. This change did not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR. This modification did not change or affect any setpoints or surveillance requirements. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification. i SE 97-11 Engineered Maintenance Action (EMA) For 'D' Well Sodium Ilypochlorite Injection Upgrade Description and Basis For Chance l j
Since its installation, the Sodium Hypochlorite Injection System at the 'D' Well has had a rather abundant repair history. Leaks had been the dominating problem, some resulting from degraded / broken parts and others from ineffective sealing of piping ji ints. Early piping joint leaks were primarily the result of faulty assembly techniques which were eventually corrected. More recent problems have often stemmed from chronic chemical degradation of the polypropylene 4-function valves and from the unavoidable flexure of threaded connections due to equipment i
vibration and stresses from redundant train repair work. Several leaks ;
amounted to substantial hypochlorite spillage. Always, the repair work i has included clean-up and/or handling of hypochlorite, an innate hazard to ;
chemical technicians and maintenance workers. Evolution of the Well l Water chlorination has resulted in an increase to the typical rate of )
hypochlorite injection needed to where the delivery capacity of the two metering pumps was only marginally adequate. !
The following changes have been made to the 'D' Well Sodium l Hypochlorite Injection System: l The system electrical logic now deenergizes the hypochlorite pump if the
'D' Well Water pressure switch senses a pressure drop below its set point. ;
The heart of the new system is a Pulsar Pulsafeeder metering pump. One I of the more significant simplifications of the system is the elimination of the day tanks. The metering pump takes continuous suction directly from the hypochlorite bulk storage tank. The metering pump and its suction 1 lines are spill protected by an epoxy coated catch-basin under the pump !
and an overflow tray that empties out through the hypochlorite house wall i
into the spill basin for the storage tank. The under-pump catch-basin is l drainable into a bucket by a ncrmally closed valve. I Level indication for the storage tank is provided by an indicator / switch with Hastelloy C diaphragm seals. The differential pressure gauge is weatherized by a silicone liquid fill and is connected by capillary tube to l the diaphragm seals. The gauge reads in inches of Sodium Hypochlorite l 9
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I and technicians can utilize existing conversion charts to convert inches of level height to gallons of tank volume. Dial mounted settings for high and low leve; shut off the metering pump on fall at low level and will turn on a warning light on rise at high level. The waming light is mounted so as to be visible te both the truck driver responsible for the tank filling and to the attending cht mistry technician. An injection quill penetrates the 'D' Well Pump discharge pipe between the discharge check valve and the floor slab of the 'D' Well Pump House.
Safety Evaluation Summary The refurbishment of the Sodium Hypochlorite Injection System into the
'D' Well does not affect any of the inputs considered in the accidents analyzed in SAR or the NSOA. This change does not alter the interface between the Well Water System and the plant and cannot cause an accident or increase the likelihood of an accident. The probability of an occurrence of the accidents discussed in the SAR and NSOA is based on initial conditions and assumptions which do not depend on the end use of or interactions with the Well Water System. Therefore, this activity did not result in a condition which increases the probability of occurrence of an accident previously evaluated in the SAR. There is no increase in the radiological consequences of any previously analyzed SAR accident. The changes made by this activity do not change, degrade or prevent actions described or assumed in an accident discussed in the SAR. This change does not alter any assumptions previously made in evaluating the radiological consequences of an accident, nor does it play a role in mitigating the radiological consequences of an accident described in the SAR. The modifications to the system made by this activity have no impact on systems, structures, or components important to safety. This activity increased the reliability of the Sodium Hypochlorite Injection system. Although the installation of a larger hypochlorite metering pump increases the injection capacity and thereby the potential chlorine cmntration in the well water, a chlorine concentration resulting from a maximum output of the pump with normal well water flow rates will have no immediate effect on the function of the well water or any other system.
Chlorine monitoring equipment was not adversely affected. The power consumption of the altered Sodium Hypochlorite Injection System and storage tank monitoring has been fully evaluated and is within parameters which have no impact on 'D' Well Pump operation and control. The Well Water System has no safety significance in the SAR. The SAR has evaluated the potential for equipment damage due to a cooling water leak in containment. The existence of chlorine in the water is within the bounds of this evaluation. This activity made no modifications to the Sodium Hypochlorite injection System into the Well Water System. The Well Water System has no safety significance in the SAR. There are no credible scenarios where the failure of this system in a non-conservative ic
l direction could increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No physical or electrical separation criteria are affected by this alteration. The equipment ,
used to chlorinate Well Water was improved, but the net etTect on the Well j Water System was minimal. This alteration will not create an accident of a different type than any evaluate d previously in the SAR. Chlorination is believed to be an aid to the efficiency of the Drywell coolers and other components. This alteration improves the reliability of the well water I chlorination. Therefore, the risk to plant ::hutdown due to inadequacy of l Drywell cooler capacity may be reduced by this alteration. This change does not reduce the margin of safety as defined in the basis for any Technical Specification. There are no Technical Specifications associated with Sodium Hypochlorite Injection into the Well Water System.
SE 97-12 EMA For Radwaste System Control Panel Logic Change Description and Basis For Chance This modification deleted the automatic opening function of the Radwaste Centrifuge Bypass to Waste Sludge Tank Valve. Two previous modifications made automatic operation of this valve both unnecessary and undesirable. The NSOA and Design Basis Documents (DBDs) were reviewed. No safety significance was identified.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR and it did not increase the consequences of an accident evaluated previously in the SAR. There are no accidents related to the Solid Radwaste System which have been evaluated in the SAR. This change did not adversely affect any equipment which has an important to safety function. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The important to safety function of the Solid Radwaste System is to retain radioactive materials. For this reason the pressure boundary function of this system is Quality Level 2. The changes made were to logic only and therefore did not affect the pressure boundary function.
This change does not introduce any new accidents, therefore it did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and it did not create the possibility of a malfunction of equipment important to safay of a different type than any evaluated previously in the SAR. There are no bases for any Technical Specifications associated with the affected radwaste valves or logic,
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therefore this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-21 EMA For Offgas Recombiner Well Water Sample Tubing Replacement Description and Basis for Chance The instrument installed in the Well Water System to measure total l residual chlorine at the Offgas Recombiner Sample Station was detecting l: inaccurate amounts of total residual chlorine due to copper tubing sample connections interacting with chlorine. This Engineered Maintenance Action (EMA) removed the copper tubing and replaced it with stainless steel tubing. This change also provided an additional connection for iron I and copper sampling.
l Safety Evaluation Summarv l- The work involved with this change was Quality level 4, non-safety !
related, and non-seismic. The Well Water System is not safety related, f and does not mitigate any transients or accident in any special events. The l NSOA does not address the Well Water System, and there are no l Technical Specifications associated with the Well Water System. No
- l. accidents related to the Offgas Recombiner room sample instrument have L been evaluated in the SAR. This change does not affect any equipment 1 l which has a function important to safety. There are no accidents related to l' the Offgas Recombiner Room Sample Station which have been evaluated l in the SAR. There are no bases for any Technical Specifications
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SE 97-22 EMA For Relocation Of Circulating Water Sample Point l
l Description and Basis For Chance
, This modification provided water to the Circulating Water System turbidimeter by relocating the sample point from a low spot on the Circulating Water blowdown line to a high point on the Circulating Water L line and removing an unneeded needle valve. The turbidimeter samples ;
ll the Circulating Water System blowdown water to provide control for the ,
Chemical Injection System pumps. This modification climinated the tendency of silt to settle out which could have produced an unrepresentative sample and frequently caused the line and the needle valve to plug. Throttling characteristics of the needle valve were not
- required since the pressure in the Circulating Water System is less than 35 psi. j 1
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Safety Evaluation Summarv i There are no accidents related to the Pumphouse Sample Station which have been evaluated in the SAR, therefore this change did not increase the probability of occurrence of an accident evaluated previously in the SAR and it did not increase the consequences of an accident evaluated j
previously in the SAR. This change did not affect any equipment which has an important to safety function and therefore, it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR and it did not increase the l
consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The possibility of an accident of a different type than any evaluated previously in the SAR was not created. Since this I
change did not affect any equipment which has an important to safety function it did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. There are no bases for any Technical Specifications associated with the affected sample lines, therefore the margin of safety as defined in the basis for any Technical Specification is not reduced.
l SE 97-32 (Revision 1) Temporary Modification To Bypass High/ Low Offgas !
Residual Oxygen Concentration Trip For Ilydrogen Water Chemistry (IlWC)
Descriotion and Basis For Chance This temporary modification prevented the HWC System from isolating while allowing the replacement of a relay in the HWC panel as well as the cleaning and calibration of the Offgas Residual Oxygen Analyzers. l Replacement of a relay for the Offgas Analyzer Train Select Switch was l needed to help resolve past problems when switching from one train to the other. It is believed that Offgas Residual Oxygen spiking and isolations ;
have occurred as a result of the relay. Offgas Residual Oxygen Analyzers were cleaned and calibrated to eliminate spikes caused by moisture buildup on the elements. During the time that this temporaly modification was installed Offgas Residual Oxygen Concentration and Offgas Hydrogen Concentration were monitored.
Safety Evaluation Summary The installation of this modification did not affect the probability of occurrence of an accident previously evaluated in the SAR. This change did not alter the initial conditions assumed for any of the accidents analyzed, nor did it alter operator or automatic actions of any safety systems that are relied upon for accident mitigation. HWC does not 13
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interface with any safety related systems or equipment that are used to I mitigate an accident. The charcoal adsorbers typically operate at about 80 degrees F and are designed to limit the temperature of the charcoal to well below the charcoal ignition temperature, precluding overheating or fire and consequent escape of radioactive materials. The hydrogen concentration of the gases from the air ejector is maintained below the flammable limit by providing adequate steam flow for dilution at all times.
This steam flow rate is monitored and alarmed. The preheaters are heated by steam rather than electricity to eliminate the presence of potential ignition sources and to limit the temperature of the gases in the event of cessation of gas flow. The recombiner temperatures are monitored and i alarmed to indicate any deterioration of performance. A hydrogen l analyzer downstream of the recombiners provides an additional check. I The Offgas System is designed to be explosion resistant in the unlikely event a combustible mixture exists. This modification did not change any oxygen concentrations in the OfTgas System. In the unlikely event, that elevated oxygen concentration would occur, the Offgas System is designed 1 so that any quantities of gaseous radwastes inadvertently released will ,
result in radiation levels within the annual exposure limits of 10 CFR 20. '
Therefore, the consequences of an accident evaluated previously in the SAR is not increased. Neither HWC nor the Offgas Oxygen Analyzers interface with equipment important to safety, therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR is not increased. The HWC System and Offgas residual oxygen analyzers do not afTect or communicate with equipment important to safety. In the unlikely event the System is required to be shutdown, capability exists to manually shutdown the System. Therefore, this change does not increase the consequences of a malfunction of equipment important to safety. New failure modes are not created by the use of this modification. This change prevents automatic isolation / trip of the HWC System in the event of High/ Low Offgas Residual Oxygen concentration condition. However, monitoring of hydrogen concentration was available in addition to alarm capability to the operators in the control I
room. The bypassing of the trip did not create the possibility of an accident of a different type than previously evaluated in the SAR. The i
bypassing of the trip did not have any impact on equiprnent that was required to shutdown the plant. Therefore, this change did not create the possibility of a malfunction of equipment important to safety of a different type. HWC is not in Technical Specifications and is not safety related.
The Offgas oxygen analyzers are also not in the Technical Specifications and therefore, there is no reduction in the margin of safety as defined in the Technical Specifications.
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L SE 97-34 Temporary Modification To Install Recorder To Monitor Reactor Recirculation Pump Runback l
l Description and Basis For Chance A Gould four channel recorder was installed in a Control Room panel to '
monitor the input, output,45% run back signal, and 120 VAC power for the 'B' Reactor Recirculation (Recirc) Pump Motor / Generator (M/G) Set speed controller. The recorder facilitated the troubleshooting process for the 'B' Recirc Pump speed change which occurred on January 23,1996, i Prior to installing the recorder the scoop tube for the speed controller was locked, precluding any unwanted M/G Set speed change which could have !
resulted in a plant transient. This was performud in accordance with existing plant Operating Instructions. Once the recorder was installed, the scoop tube was unlocked and the monitoring process was started. The high impedance of the recorder used did not affect the operation of the M/G Set. With the exception of the 120 VAC power supply connection, ring lugs were utilized to provide long term reliable connections to the recorder. The M/G Set speed controller function was not affected by the installation of the recorder.
Safety Evaluation Summary There are no accidents related to the Recire M/G set speed controller. All credible failures of the M/G Set speed controller are bounded by accident and transient evaluations contained in the SAR. This change did not increase the probability of occurrence of an accident evaluated previously in the SAR, and it did not increase the consequences of an accident evaluated previously in the SAR. This change did not create the possibility of en accident of a different type than any evaluated previously in the SAR. This change did not affect any equipment which has an important to safety function, therefore this change did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This change did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The bases for Technical Specincations associated with the Recire System are not affected by this change. The margin of safety is denned for this system by other bounding events which are contained in the SAR. Therefore, this change did not reduce the margin of safety as denned in the basis for any Technical Specincation.
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l SE 97-35 Temporary Modification To Connect Well Water System And General Service Water (GSW) System 1
Description and Basis For Chance '
A mechanicaljumper was installed between the Well Water and General Service Water Systems to supply cooling loads in the Auxiliary Boiler room from Well Water as opposed to GSW. The GSW piping was repeatedly clogging with sediment. Well Water provides a cleaner supply of cooling water and prevents fowling of the cooling water lines and heat exchangers. The Well Water System operates at a lower temperature and a slightly higher pressure than the GSW System. Therefore, the coolers involved had a higher effectiveness with Well Water than with GSW. The line classes of both systems are the same. Check valves were installed to l prevent backflow from GSW to Well Water. In addition a valve on the l
GSW System was closed to prevent backflow from the Well Water System to the GSW System. I Safety Evaluation Summary This change did not increase the probability of occurrence of an accident I evaluated previously in the SAR and it did not increase the consequences !
of an accident evaluated previously in the SAR nor did it create the possibility of an accident of a different type than any evaluated previously in the SAR because, there are no accidents related to Well Water or GSW which have been evaluated in the SAR. This change did not increase the l
probability of occurrence of a malfunction of equipment important to !
safety evaluated previously in the SAR and it did not increase the i consequences of a malfunction of equipment important to safety evaluated i previously in the SAR, and this change did not create the possibility of a I malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This change does not affect any ,
equipment which has an important to safety function. The margin of I safety as defined in the basis for any Technical Specification is not reduced. There are no bases for any Technical Specifications associated with GSW or Well Water. The added cooling load to the Well Water System is extremely small. These loads are at the end of the loop of both systems and therefore do not affect any other cooling loads.
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SE 97-37 Temporary Modification To Install Vent Line From Steam Jet Air Ejector Condensate Return Pumps To Condensate Return Tank l.
Description and Basis For Chance The purpose of this temporary modification was to install vent lines from l the discharge of the boosier stages on the Steam Jet Air Ejector
!' Condensate Return Pumps back to the source, the Condensate Return
! Tank. This vent path evacuates entrained air from the interstage area of i
the pumps. Per the vendor, vent lines were needed to assure satisfactory i operation in applications for vacuum service. These lines eliminate air j entrainment and ensure proper pumping action of the pumps. Installation of these vent lines with this temporary modification have brought the arrangement into conformance with vendor recommendations. The materials used have been approved by the vendor and are adequate. The condensate return tank and its asseciated components are not addressed in the NSOA or any Design Basis Documents.
Safety Evaluation Summary The equipment involved in this temporary modification is not safety l significant and is not associated with any accidents evaluated in the SAR.
Therefore, the probability of occurrence of an accident evaluated previously in the SAR is not increased, and the consequences of an accident evaluated previously in the SAR are not increased. This change l did not increase the probability of occurrence of a malfunction of l equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to l safety evaluated previously in the SAR. This change did not create the l possibility of an accident of a different type than any evaluated previously l in the SAR, and it did not create the possibility of a malfunction of i
equipment important to safety of a different type than any evaluated i previously in the SAR. The equipment involved in this temporary modification was not safety significant and was not associated with any l safety significant equipment; therefore, this temporary modification did l not reduce the margin of safety as defined in the basis for any Technical l Specification.
l SE 97-38 Temporary Modification Connecting Service Air System And Main i
Plant Secondary Heat Loop Descrintion and Basis For Chance Pressurizing the Main Plant Secondary Heat Loop with Service Air is necessary in order to blow down all of the fluid from the Main Plant Intake Coils. This activity assures that the maximum amount of glycol is I 17 L
j removed from the system prior to opening the loop to the Well Water System for summer operation. Removing the fluid from the coils is necessary when placing the system in the summer mode, and during times when the coils need to be dry for maintenance activities. The Seconda'v IIeat Loop is removed from service and is isolated from the Well Water System and the Demineralized Water System during the time this 1
temporary modification is in place.
Safety Evaluation Summary Installing an air hose to the air station and connecting it to the Secondary lieat Loop for the purpose of blowing down the IIeat Loop coils and piping does not result in or initiate an accident. This activity is not directly associated with any system, component, or structure whose failure or malfunction has the capability to cause an accident as described in the SAR, This modification had no impact on the operating envelope assumed in the accident analyses, therefore, the initial conditions at the onset of any accident evaluated previously remain valid. The number of unacceptable results are not increased by performing this activity. Cross-connecting the Service Air System with the Secondary lleat Loop is a manual evolution (manual valves and connections), automatic actions / responses remain unchenged. In addition, this activity is not associated with fuel safety limits. or radiological barriers. This
- modification was evaluated in order to ensure that equipment which is important to safety would not be directly, or indirectly affected. The only safety equipment in the vicinity which has a potential to be impacted by l this activity are the components which contribute to Secondary Containment integrity. The service air hose serves as a Secondary Containment boundary during the time that it is connected and valves are open. This modification did not increase the probability of a malfunction with the Secondary Containment System. The air hose was of sufficient design and manufacture to provide pressure boundary integrity against l Service Air System pressure. This was adequate assurance that the hose would provide integrity against Secondary Containment design pressure.
Furthermore, the valves were not opened daring times that the hose was not connected. Administrative control of the valves ensured that the Secondary Containment was not opened through the Secondary IIcat Loop
- piping. Installation of this temporary modification did not increase or decrease the pressure of the Secondary Containment to unacceptible levels, nor did it cause the Service Air System or the Secondary Heat Loop l System to have an adverse impact on Secondary Containment components l which could result in a loss of Secondary Containment ftmetion. The likelihood of a different accident was not increased. A Service A ir imse failure, or an opening in Secondary Containment through the Secondary lleat Loop piping would not result in a loss of Secondary Containment l
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l integrity. Rupture of, damage to, or a loss of function with the Secondary IIeat Loop and/or associated compoaents would have no impact on equipment important to safety. The Secondary Heat Loop System is not necessary for safe plant operation. The possibility of a mHfunction was not increased due to this temporary modification. This me.iification did I not impact the ability of the system to achieve the margins of safety described in the T chnical SpecMeations.
l SE 97-40 Reactor Building Supply Fan Dampers Temporary Modification '
Description and Basis For Chance This temporary modification prevented the Reactor Builaing supply fan inlet dampers from cycling by removing the auto operation feature and replacing it with a manual control. Design requirements called for both supply and exhaust fan inlet dampers e operate independently to maintain Reactor Building exhaust trunk pressure at -0.5" water pressure. During certain fan configurations the two control systems would " fight" each other to maintain proper exhaust trunk pressure. This resulted in high trunk pressures and low supply fan flows due to damper oscillations.
Safety Evaluation Summary Replacing the automatic pneumatic controller with a mant al pneumatic controller did not result in, or increase the p(ssibility of, or initiate an accident as described in the S AR. The positii a of the suppiy fan inlet uampers, whether positioned automatically or manually, has no inptet on safety related equipment. The systems affected by this activity were th -
Reactor Buildim Ventilation Supply System and the Plant Instrument Air System. These systems do not rerve as nuclear system process barriers.
rea;tivity control devices, fuel barrier cooling systems, or could otherwise experience a failure which could lead to the release of radioactive material from one or more barriers. The consequence of each accident evaluated previously is sufficiently extreme so as to envelope the consequence of any potential accident resulting from this temporary modification. This change had no impact on the operating envelope a emed in the accident analyses. Therefore, the initial conditions at the onset of any accident evaluated previously remain valid. This provides assurar.ce that the consequences of an accident are not increased due to a deviation of the initiai conditions. This modification did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The dampers were modulated to control the amont of fresh air supplied w the Reactor Building. The components and systems tssociated with this activity were not considered important to nuclear safety as it relates to the UFSAR accident analysis The 19
components associated with this temporary modification are all non-safety related. The dampers which were swapped from auto to manual control in this temporary modification are not dampers which contribute to Secondary Containment integrity. The manual controller positions the same dampers that the automatic controller did. No new controlled loads have been added to the control loop. The controller, and the dampers are connected to the plant Instrument Air Systemjust as before. Operations personnel could adjust airflow as necessary to optimize space temperature.
Equipment important to safety (i.e. secondary containment isolation dampers, and logic) remains as far removed from the Supply Fan Volume Control System as before. This activity did not increase the consequences of a malfunction of equipment important to :;afety evaluated previously in the SAR. The only equipment important to safety which had the potential to be impacted by this activity were the components which contribute to Secondary Containment integrity. With regard to Secondary Containment, the satisfaction of" safety design bases for accidents" required that the containment retain its integrity for certain accident situations. The UFSAR safety analysis assumes that the safety objective of the Secondary Containment System is maintained. The safety objective is to limit the release to the environs of radioactive materials so that offsite doses from a postulated design bases accident will be below the guideline values of 10 CFR 100. Secondary containment integrity is maintained as long as internal pressure remains below the maximum allowable value of 7" of water. This modification did not increase or decrease tl.e pressure of the Secondary Contairanent to unacceptable levels. There was nu impact on Secondary Containment components which could result in a loss of Secondary Containment function. The number of unacceptable results is not increased by performing this activity. Capability of the Secondary Contaimnent System to perform its function in accordance with General Design Criteria is maintained. This activity is .ot associated with fuel safety limits, or radiological barriers. This change does not create the possibility of an accident of a different type than any evaluated previously in the SAR. NSOA events were reviewed and fotmd to be suf& ' uly complete such that they bound any accident that could be caused y this modification.
Although unlikely, the types of events that the acih ity could have created include insufficient, or excessive ventilation supply to the Reactor Building, neither of which constitute, an accident or increase the possibility of an accident. The Reactor Building Exhaust Fan Damper Control System remained functional during normal plant operation. The system could automatically respond to a manual adjustment of the supply fan dampers if needed. The Secondary Containment could perform its safety function regardless of mispositioned supply fan dampers. Reactor Buildig gaeous effluent is monitored contir,uously regardless of Reactor 20
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Building pressure. The logic which trips off the supply fans during high l
pressure in the Reactor Building vent shaft remained functional. The Secondary Containment Isolation System functioned as designed. The Specific Design Safety Standards associated with the Reactor Building l Ventilation System are satisfied with the temporary modification in place.
l Since the Specific Design Safety Standards are met, the possibility of an accident of a different type is not increased, and the possibility of a l malfunction of equipment important to safety was not increased due to this
! temporary modification. The possibility of causing a malfunction with the l
Secondary Containment System, components, or structures is not increased because this change did not present a new or unusual challenge to the Secondary Containment System, components, or structures. A malfunction of equipment / components associated with the item under evaluation does not affect equipment important to safety regardless of whether previously evaluated. The safety margins defined in Technical
[ Specification Bases remain valid. The Secondary Containment isolation System is equally effective regardless of the position of the supply fan inlet dampers. The fans and dampers are located outside of the Secondary
!- Containment boundary, and do not contribute to the integrity of the boundary.
SE 97-41 EMA For Design Change Of'A' Standby Diesel Generator (SHDG)
Room Hoist Hook Plate j I Descrintion and Basis For Chance The hook plate for the two hoists mounted on the monorail in the 'A' ;
i . SBDG room were replaced to meet requirements during a seismic event. '
r The main purpose for this change was based on the enhancement needed for the Monorail System to fully comply with the effects of seismic 1 classification. The concern was the hoist breaking loose from the secured position and allowing it to travel down the monorail and impacting on the safety related Diesel Generator.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The upgrading of this secondary support did not alter the SAR or previous evaluations by engineering. This secondary support did not impact the ability of the primary supporting member to provide its structural integrity. The hoist hook plate was
, evaluated to increase the margin of safety and thereby decreases the
! consequences of an accident. This activity did not increase the probability l of occurrence of a malfunction of equipment important to safety evaluated
, previously in the SAR. This change did not increase the possibility of a 21 i
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plant transiem. The main monorail framing structure was qualified for seismic loading. The detailed evaluation for the hook plate installation was upgraded by enhancing the design to withstand seismic Design Basis Earthquake (DBE) loading without exceeding design stresses or interacting with adjacent structures or components. This change did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR and it did not create the possibility of an accident of a different type than any evaluated prevkusly in the SAR. This activity did not create the possibility of a raalfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This activity precludes tbc possibility of a malfunction to the Diesel Generators as a result of the hook plate being redesigned to increase the margin of safety. This activity is not ;
specifically addressed in the Technical Specification and does not reduce g the margin of safety as defined in the basis.
SE 97-42 EMA For I)esign Change Of'B' Standby Diesel Generator (SBDG)
Room Hoist Hook Plate Description and Basis For Chance The hook plate for the two hoists mounted on the monorail in the 'B' SBDG room were replaced to meet requirements during a seismic event.
The main purpose for this change was based on the enhancenent needed for the Monorail System to fully comply with the effects of seismic classification. The concern was the hoist breaking loose from the secured position and allowing it to travel down the monorail and impacting on the safety related Diesel Generator.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The upgrading of this secondary support did not alter the SAR or previous evaluations by engineering. This secondary support did not impact the ability of the primary supporting member to provide its structural integrity. The hoist hook plate was evaluated to increase the margin of safety and thereby decreases the consequences of an accident. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This change did not increase the possibility of a plant transient. The main monorail framing structure was qualified for seismic loading. The detailed evaluation for the hook plate installation was upgraded by enhancing the design to withstand seismic Design Basis Earthquake (DBE) loading without exceeding design stresses or interacting with adjacent structures or components. This change did not y
increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR and it did not create the possibility of an accident of a different type than any evaluated previously in the .
SAR. This activity did not create the possibility of a malftmetion of-equipment important to safety of a different type than any evaluated previously in the SAR. This activity precludes the possibility of a malfunction to the Diesel Generators as a result of the hook plate being redesigned to increase the margin of safety. This activity is not I specifically addressed in the Technical Specification and does not reduce the margin of safety as defined in the basis.
SE 97-46 EMA For Removal Of Outage Temporary Power Supply Transformer Descrintion and Basis For Chance The purpose of this modification was to change the load of breaker i B3215 from " Outage Temporary Power Supply Transformer 1 Y32" to a test equipment receptacle. Transformer lY32 was removed from service and an electrical box with a receptacle was installed. This test equipment receptacle supplies power to the test set used during inverter calibrations in switchgear room 1 A3. No other equipment power source is located in I A3.
Safety Evaluation Summary I The removal of 1Y32 and the installation of a test equipment receptacle did not impact the ability of 1B32 to perform its intended function. The load on IB32 has been reduced by approximately 25 amps. This change did not affect the operation of the system and will not increase the probability of an accident previously evaluated in the SAR, and this change did not affect the operation of the system and does not increase the consequences of an accident previously evaluated in the SAR. Also the new breaker provides better coordination with the 1B32 load center breaker hence reducing the possibility of a fault induced transient affecting 1B32. This modification did not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR. With 1B3215 on and the test set inservice, inverter / regulating transformer calibrations require an additional 7 amps of IB32. The additional 7 amps, conservatively increases the total 1B32 load to well within the Motor Control Center design rating. The additional load on 1B32 will have an insignificant effect on the diesel accident scenario 4
loading, due to the magnitude and the infrequent use of the test set receptacle (used twice a cycle). This modification did not increase the consequences of a malfunction of equipment important to safety that has 23
previously been evaluated in the SAR, and it did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created.
The system functions as it did previously. This change did not reduce the margin of safety as defined in the basis for any Technical Specification.
' SE 97-47 EMA For Replacement Of 'B' Residual Heat Removal Heat I i
Exchanger (RHRHX) Relief Valve Description and Basis For Chance The purpose of this change was to replace the 'B' RIIRHX shell side pressure relief valve with a newer, more reliable model and to abandon the steam service valve which was no longer required since the steam condensing mode of the system has been disabled. The scope of this l activity included adding flange connections to the existing relief valve l
piping to facilitate maintenance, replacing the 'B' RHRHX Shell Side '
Pressure Relief Valve, and installing a gag screw on the 'B' RHRHX Inlet Pressure Relief Valve.
Safety Evaluation Summary This activity does not increase the probability of occurrence of an accident evaluated previously in the SAR. The change of thermal relief valve manufacturer, method of attachment, as well as the abandonment of the steam safety valve which was no longer required, does not increase the probability of an accident. The design specification of the replacement valve has been reconciled to the original code of construction for relief valves which meets the design specification requirements of the originally supplied valve. The flange connection is allowed by the piping class described in a specification. The slight reduction in the orifice area of the replacement valve did not adversely aEcct perfonoce and the setpoint remained the same. The abandonment of the steani condensing made of the RHRHXs has been previously analyzed in conjut.ntion with another modification. This activity did not increase the conseqaences of an accident evaluated previously in the SAR. The discharge line of the two relief valves is connected to the Suppression Pool through a penetration which is an extension of containment. The replacement valve was hydrostatically tested to 100 psig on the outlet side. The steam service valve is equipped with a bellows seal which isolates the outlet from the top works of this valve which has been hydrostatically tested to 150 psig.
These tests demonstrated the top works of the valve with the packed lifting lever configuration and the gagged steam service valve are capable of withstanding the maximum pressure in the Torus. The discharge piping 24
from the heat exchanger is protected with a pressure relief valve in addition to the replacement valve. This change did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. While the RHRHX is in service, a valve is open to allow discharge.
With this valve open the shell side of the heat exchanger communicates with the RHRHX Discharge Header Pressure Relief Valve. This valve has a setpoint of 425 psig and discharges to open Radwaste. This arrangement negates the effects of backpressure on the tailpipe of the 'B' RHRHX Shell Side Pressure Relief Valve due to pressurization of the Torus during an accident. This modification did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the ;
SAR. The maximum RHR pump deadhead pressure is approximately 250 '
psig, which is much lower than the heat exchanger design prc ure of 450 !
psig. This modification did not challenge the installed equipment, and it did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The J
installation of a newer model heat exchanger sheh side relief valve
{
improved the reliability of the system. The installation of flange !
connections allows the new valve to be removed for testing and service. !
The gagging of the larger steam service relief valve prevents it from l
inadvertently lifting and sending excessive flow to the suppression pool.
This activity did not have an effect on the system's operation, setpoints, capacity, or any of the operating modes described in the Operating License and Technical Specification, therefore this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-49 (Revision 1) ECP 1590 - Residual Heat Removal System and Core Spray System Minimum Flow Instruments Modification Description and Basis For Chance This modification reconfigured the sensing lines for switches used to sense pressure across several flow orifices associated with the Residual Heat Removal (RHR) and Core Spray Systems. The configuration of tubing on these instruments required that the instruments be drained and calibrated with air. The instruments were then refilled and vented to establish a complete wet leg for proper indication. This procedure was difficult to implement due to the lack of adequate venting capability which could introduce inaccuracies into the measurement of differential pressure. In order to provide calibration without draining and subsequent venting of the system, the tubing was reconfigured to allow calibration of the switches while filled with water, a " Wet Calibration". Since the 25
instruments are no longer drained during calibration, separate collection chambers were also required to be added to collect the solids.
Safety Evaluation Summary This activity does not increase the probability of occurrence of an accident previously evaluated in the SAR. The RHR and the Core Spray Systems are not adversely affected by this modification. This modification provides adequate venting of the System when required and eliminates the need for periodic draining and venting. Isolation valves, tubing and collection chambers added by this activity are passive in nature; no new active components were introduced by this modification. New passive components added are seismically mounted and meet all design , material, environmental suitability and construction acceptance criteria. Because this activity does not affect the overall system performance in a manner that could lead to an accident, the probability of occurrence of an accident previously evaluated in the SAR is not increased, and this activity does not increase the consequences of an accident evaluated previously in the SAR.
The function of providing minimum flow affects equipment important to safety, however this function is not adversely affected by this modification. The specifications and pressure rating of the components used in this modification confonn to the requirements of the UFSAR and are consistent with the existing components. Therefore, there is no degradation of the pressure boundary. Since the reconfigured instruments meet the original system bases, electrical separation,10 CFR 50 Appendix R, seismic and environmental requirements, the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased. The consequence of malfunction of equipment important to safety previously evaluated in the SAR is not increased because the ability of RHR pump minimum flow valves and Core Spray minimum flow bypass valves is not compromised by this modification.
Since no new active components are added by this modification and all passive components are qualified and seismically installed for the applications, no new failure modes are introduced in the system. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not create the possibility of an accident of a different type than any evaluated previously in the SAR. This activity can not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR. Since no instruments were moved from their existing location, these modifications did not adversely affect any setpoint calculations associated with these instruments. This modification did not adversely affect the bases as described in the Technical Specifications. The combination of Operable subsystems to assure the availability of the minimum cooling system as specified in the Limiting 26
Conditions for Operation (LCO) in the specifications were not adversely affected by this activity. Since the conditions (operability, number of components, out-of-service times, inspection and test frequencies) that must be met for functional requirements of the Low Pressure Coolant injection (LPCI) mode of RIIR, and the Core Spray System were not affected, the margin of safety as defined in the basis for any Technical Specification was not reduced by this activity.
SE 97-50 EMA For Drywell Airlock Test Valve Replacement Description and Basis For Chance This change replaced two Drywell Airlock Local Leak Rate Test, LLRT, I Test Connection Isolation valves, which were damaged and could not be operated. The size of the valves was reduced from I inch to 3/4 inch. The type of valves, pressure rating, and class of valves were not changed. The 1 inch upstream pipe nipples to the valves were replaced with 1 inch to 3/4 inch reducers and 3/4 inch nipples. The 1 inch downstream pipe nipples and caps were replaced tvith 3/4 inch fittings. The welds to the couplings connecting the nipple to the airlock were remover cleaned and the reducer welded to the coupling. This activity was perfoad with the reactor at 400 psig with Primary Containment required. l l
Safety Evaluation Summary This activity did not inc rease the probability of occurrence of an accident evaluated previously in the SAR. The valves perform a passive safety i function. This activity did not increase the consequences of an accident evaluated previously in the SAR. The change in valve size did not affect ;
the ability of the valve 1o establish Primary Contairunent isolation or pressure boundary integrity. The consequences with the valve open are unchanged. During the valve replacement, the inner door to the Drywell airlock was closed. The leakage past the inner door was within established limits. This change did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to ufety evaluated previously in the SAR. The consequences may actuly decrease as the possible leakage path with the valve open has decreased. The possibility of an accident of a different type than any evaluated previously in the SAR is not increased, and this activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This change did not reduce the margin of safety as defined in the i basis for any Technical Specification. The change in valve size does not affect the margin of safety as the safety function can be performed by 27
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l either size valve. During the replacement, the appropriate LCO was entered and actions required by Technical Specifications were taken. This change occurred within the established Technical Specification basis.
l l SE 97-52 ECP 1591 - Well Water To General Service Water (GSW) Cross-tie l For Auxiliary Boiler IIcat Loads I
! Description and Basis For Chance This modification installed a permanent cross-tie between the Well Water j and GSW Systems. This was done to supply cooling loads in the Auxiliary Boiler room from Well Water as opposed to GSW. In the past,
! the GSW piping repeatedly became clogged with sediment. Well Water l provides a cleaner supply of cooling water and prevents fowling of the l
cooling water lines and heat exchangers.
Safety Evaluation Summary This change does not increase the probability of occurrence of an accident evaluated previously in the SAR. There are no accidents related to Well Water or GSW which have been evaluated in the SAR. This change does not increase the consequences of an accident evaluated previously in the SAR and it does not increase the probability of occurrence of a l malfunction of equipment important to safety evaluated previously in the SAR. This change does not affect any equipment which has an important to safety function, and therefore does not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This change did not create possibility of an accident of a different j type than any evaluated previously in the SAR, and it did not create the
- possibility of a malfunction of equipment important to safety of a different l type than any evaluated previously in the SAR. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification. There are no bases for any Technical Specifications associated with GSW or Well Water. The added cooling load to the Well Water System is extremely small. These loads are at the end of the loop for both systems and therefore do not affect any other cooling loads.
! SE 97-53 Temporary Modification Replacing Radwaste Floor Drain Filter Effluent Conductivity Element Description and Basis For Chance A new type and model conductivity element for the Radwaste Floor Drain Filter Efiluent Conductivity Element was installed temporarily to evaluate data over a three month period to detennine if this new type element i
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would be permanently installed. The new type element was controlled with a flow control valve, back pressure valve, and flow indicator.
The original element was installed to trend and track the effluent conductivity from the Floor Drain Filter to the Floor Drain Demineralizer in the Liquid Radwaste System. It never ftmetioned properly because it was installed at the heel of an elbow in the main process line, and it was not designed to operate at the flow rates at that location. The element installed under this temporary modification was provided with valving and flow indication, along with re-piping to remove it from the main process flow. The new element was compatible with the conductivity transmitter.
Safety Evaluation Summary This change did not affect the system description as presented in the SAR.
Changing the type of conductivity element did not increase the probability of an accident evaluated previously in the SAR. There are no accidents related to the conductivity elements for this system which have been evaluated in the SAR. This activity did not increase the consequences of an accident evaluated previously in the SAR. This change did not affect any equipment which has an important to safety function. The Radwaste System performs no safety functions and operates independent of safety related systems. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This temporary modification was installed in accordance with existing design requirements so the possibility of an accident of a different type than any evaluated previously in the SAR were not created, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. There is no basis for any Technical Specification requirements associated with the affected conductivity elements, therefore this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-55 Temporary Modification For Low Level Radwaste (LLRW) Office Area Cooling Unit Description and Basis For Chance The office area in the LLRW Building was being used as a Metrology Laboratory. This required that the space be provided with fine temperature and humidity control. This temporary modification connected an unused humidity controller to the cooling unit start logic for the area.
A high humidity condition in the return air duct causes the cooling unit to run. On a hi-hi humidity condition the second stage cooling starts. This 29
l results in a drop in the Metrology Laboratory room temperature. This causes the temperature controls to turn the heater on in the unit to maintain the room temperature.
Safety Evaluation Summary This change did not increase the probability of occurrence of an accident i evaluated previously in the SAR. Changes to the controls to the LLRW building office area HVAC System cannot cause an accident because they are not physically near any safety related, important to safety or power generation systems. The consequences of an accident evaluated l previously in the SAR are not increased since the LLRW HVAC is not required to mitigate the consequences of an accident. This change does l not increase the probability of occurrence of a malfunction of equipment l important to safety evaluated previously in the SAR because the LLRW building HVAC does not support any equipment important to safety.
Therefore, changes to the controls cannot affect equipment important to safety. The consequences of a malfunction of equipment important to safety evaluated previously in the SAR are not increased. This change did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and it did not create the possibility of a malfunction of equipment important to safety of a different type than any ,
evaluated previously in the SAR. The LLRW building HVAC cannot )
create an accident of a different type than any evaluated previously in the l
SAR. The HVAC can fail completely and not affect the reactor or any power generation systems. This change did not reduce the margin of safety as defined in the basis for any Technical Specification. The LLRW Office area HVAC is not a system described in the Technical Specifications or Technical Specifications bases.
l SE 97-57 EMA For Replacement of Condensate Demineralizers' Flow Indicating Controllers l
Descrintion and Basis For Chance The Condensate Demineralizers' Outlet Flow Indicating Controllers were l obsolete and a maintenance problem. The Fischer-Porter pneumatic controllers, with remote set point input, were replaced with a pneumatic Moore Loading Station Model.
l l Safety Evaluation Summary i
! This activity did not increase the probability of occurrence of an accident previously evaluated in the SAR. The condensate demineralizer flow controller replacement does not prevent the Condensate Demineralizer 1
30
l l System from performing its design function. The Condensate Demineralizer System cannot initiate any of the accidents described in the UFSAR or the NSOA. The Condensate Demineralizer System is not safety related. Because this modincation did not affect any system performance in a manner that could lead to an accident, the probability of an accident previously evaluated in the SAR is not increased. This change did not increase the consequences of an accident evaluated previously in '
the SAR. The replacement of the flow indicating controllers reduces the probability of a failure of the Condensate Demineralizers, therefore the reliability of the Feedwater System is increased. Since the modified control logic allows the flow to be balanced and a mimmum system differential maintained, the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased. The replacement of the flow indicating controllers did not increase the consequences of a loss of feedwater event previously evaluated in the SAR. Therefore, this activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. Replacement of the flow indicating controllers did not impact the Condensate Demineralizer System from performing its design function. No failure can be postulated by this modification that '
could create an accident of a different type. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. This change did not create the possibility of a malftmetion of equipment important to safety of a different type than any previously evaluated in the SAR. Technical Specifications identify surveillance requirements and chemistry limits for the Reactor Coolant System. The monitored parameters include conductivity, chloride concentration and pH. This modification did not che ige or affect any setpoints or surveillance requirements, therefore. this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-59 EMA For Installation Of Three-Phase Receptacle In Essential Switchgear Room Description and Basis For Chance This change involved installing a permanent 50 amp,480 volt, three phase power receptacle in the essential switchgear room, l A4. The supply breaker for this receptacle is part of a non-essential bus. The receptacle is used to support the use of testing equipment in the room.
l Safety Evaluation Summary The non-essential buses are not included in the NSOA analysis. This change did not impact the discussion of the <.lectrical systems as described 31
l l
l in the Design Basis Documents. Technical Specifications ident'Ber u1e electrical systems limitations and surveillance requirement to assure an adequate supply of electrical power for operation of those systems for i
safety. This change did not impact these requirements or the ability to maintain them. Adding a power receptacle supplied by a non-essential bus in the I A4 switchgear room does not affect the operation of the system l and will not increase the probability of occurrence of an accident previously evaluated in the SAR. The non-essential busses do not support the response of the plant to an accident. This change does not affect the I operation of the system and will not increase the consequences of an accident previously evaluated in the SAR. The probability of occurrence i
l of a malfunction of equipment important to safety evaluated previously in j the SAR is not increased. The non-essential bus is not essential to the support of plant safety systems. This modification will not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR and it will not increase the consequences of a malfunction of equipment important to safety that has previously been evaluated in the SAR. This change did not create the possibly of an accident of a different type than any evaluated previously in the SAR. The Power System Analysis (PSA) has evaluated the addition of this load and found it to be acceptable. The possibility of a malfunction of equipment important to safety of a difTerent type than any evaluated previously in the SAR was not created. The electrical system functions as it previously did. This change did not reduce the margin of safety as defined in the basis for any Technical Specification.
1 SE 97-64 EMA For Primary Plant Computer Upgrade Description and Basis For Chance The Primary Plant Computer (PPC) host processor was upgraded with a model by the same manufacturer. The new processor is a functionally equivalent upgrade with similar architecture, but with improved processing and data handling capabilities. The interfaces necessary for the new computer to communicate with the remainder of the system was installed and tested. Several peripheral devices were also changed out.
The reason for this modification was primarily to achieve improved performance from the plant's Main Computer System. The new system provides faster execution times by means of an improved processor and more efficient memory use and allocation. This results in a more reliable l system capable of handling the increasing data delivery and processing i
demands being placed on the PPC System.
32
Safety Evaluation Summary As indicated in the NSOA basis analysis, the Process Computer System only provides safety actions via operator observation of the core neutron flux distribution for the plarmed operational events, heatup, power operation, and achieving shutdown. No safety actions by the Process Computer are prescribed for any of the abnormal operational transients, l accidents, or other events described in the NSOA or UFSAR. All of the accidents evaluated in the UFSAR assume an initial set of plant conditions. Some of these events require knowledge and control of the core neutron flux distribution by the operators which can be provided by the Process Computer. This change does not affect the ability of the operators to obtain the necessary information to maintain and control the core neutron flux distribution. Because this change does not change the functionality of the Process Computer, the probability of occurrence of an accident previously evaluated in the SAR is not increased. This activity did not increase the consequences of an accident evaluated previously in the SAR. The only system affected by this change was the Process Computer System which is not considered by the UFSAR to be safety related equipment. Regardless of this, the new processor is an upgraded version which meets and, in some cases, exceeds the original design specifications of the previous processor. This new processor is considered to be more reliable and less susceptible to failure than the previous unit.
No adverse effect on the computer system resulted from this modification. l This activity did not increase the probability of occurrence of a !
malfunction of equipment important to safety previously evaluated in the SAR. Although the likelihood of a malfunction is reduced by the increased reliability of the new processor, the consequences of failure remain the same. These consequences do not, however, impact any safety related system as defined by the UFSAR. This ac'ivity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not create the possibility of an accident of a difTerent type than any evaluated previously in the SAR.
This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The Process Computer System is not considered in the Technical Specifications, therefore, the changes made did not reduce the margin to safety as defined in the basis for any Technical Specification.
l i
33
SE 97-70 EMA For 'A' Reactor Feedwater Pump Auxiliary Lubricating Oil
, Control Logic Modification 4
Description and Basis For Chance This change modified the control logic for the 'A' Reactor Feedwater L ;
Pump / Motor Skid Oil System. This change allows the Opera: ors to shut the plant down in a controlled manner in the event the Skid Oil System would lose pressure. Previously, the logic would have resulted in a Feedwater Pump trip and potential plant SCRAM in the event the Skid Oil System would lose pressure due to the failure of the Shaft Driven Oil Pump. This modification required a change to the Condensate /Feedwater System P&lD found in the UFSAR. The change to this drawing consisted i of changing an equipment identification number for a relay and adding a note showing that the relay has an alarm function.
- The following modifications were made to the control logic for the ,
Feedwater Pump / Motor Skid Oil System:
. A time delay relay was removed, allowing the Auxiliary Lube Oil Pump to auto start based on system oil pressure.
A seal-in contact was installed around the Auxiliary Lube Oil Pump Auto Start Pressure Switch to preunt the Auxiliary Lube Oil Pump ,
from cycling on pressure.
- A relay was installed to activate an annunciator window to indicate that Auxiliary Lube Oil Pump has started. This alarm window is activated any time the Auxiliary Lube Oil Pump is running.
Safety Evaluation Summary The change to the control logic for the Reactor Feedwater Pump Auxiliary Lube Oil Pump increased the reliability of the Feedwater Control System. >
The logic change allows the Auxiliary Lube Oil Pump to maintain lube oil pressure in the event of a failure of the Shaft Driven Lube Oil Pump. This logic change did not increase the probability of an accident as described in SAR. The change to the control logic for the Reactor Feedwater Pump Auxiliary Lube Oil Pump made the loss of the Feedwater System as described in SAR less likely. This activity did not increase the consequences of an accident previously evaluated in the SAR and it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This change made the Feedwater Control System more reliable. Upon loss of oil pressure, 34
l 4
this change allows the Auxiliary Lube Oil Pump to try to maintain system oil pressure. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This activity did not create the possibly of an accident of a different type than any evaluated previously in the SAR. The logic change to this syste n increased the reliability of this system by allowing the Auxiliary Lube Oil Pump to supply lube oil in the event the Shaft Driven Pump fails.
Thi:, activity did not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR. The Feedwater Control System is not discussed in the Tecluiical Specifications. Therefore, this activity did not reduce the margin of safety as defined in the basis for the Technical Specifications.
SE 97-71 EMA For Isolation and Drain Valves Added To Auxiliary Boiler Description and Basis For Chance The Condensate Return Unit for the Auxiliary Heating System Boiler, receives drainage from the various steam traps for the Auxiliary Heating System Boiler and returns the collected condensate to Deaerator Heat Exchanger. During maintenance activities on the Auxiliary IIeating System Boiler, there are no drain valves to drain the condensate return piping from the Condensate Return Unit during repair or replacement of the Condensate Retum Unit pump or discharge check valve. Draining this line requires removal of a check valve bonnet and draining approximately 50 gallons of boiler water from the Deaerator Heat Exchanger through the bonnet of the check valve. This change provides an isolation valve and a drain valve to facilitate future maintenance on the Condensate Return Unit for the Auxiliary Heating System Boiler. This activity was limited to the installation of an isolation valve and a drain valve on the one inch condensate return piping from the Condensate Return Unit.
Safety Evaluation Summary The Auxiliary Heating System Boiler is not a contributor or initiator of any transients, accidents, or special events evaluated in the UFSAR .
There are no permanent connections from the Plant Heating Boiler System to any safety related equipment. Therefore, this activity did not increase the probability of occurrence of an accident. The Auxiliary Heating System Boiler is not relied upon to mitigate any transients, accidents, or special events. Therefore, this activity did not increase the consequences of any accident. This activity will not implement any changes in the method of operation of the plant. Therefore, it had no effect on the probability of equipment malfunction. Therefore, this activity did not increase the consequences of a malfunction of equipment important to I
35
_..m _. _.___..__-_ _ _ _. _ ._.. _ _ _ _ . _ . _ . .
l l
l l safety, and it did not create the possibility of an accident of a different - i
[ _ type. This activity will not create any permanent connections to safety related equipment / equipment important to safety, or implement any changes in the method of plant operation. Therefore, it did not create the 3
possibility of a malfunction of equipment important to safety. There are
[ no specific Technical Specification requirements for the Auxiliary lleating L System Boiler. This system is not a contributor or initiator of any of the l accidents evaluated in the UFSAR. There are no permanent connections i
to any equipment important to safety, and no permanent connections are created by this activity. The Auxiliary Heating System Boiler does not impose a negative impact on any Technical Specifications or safe operation of the plant. Therefore, no margin of safety is reduced.
SE 97-75 Temporary Modification To Install Blank Flange For Condensate Phase Separator Tanks '
l l Description and Basis For Change. !
i This activity involved installation of two blank flanges, one in the vent line for each of the Condensate Phase Separator Tanks and a second in the
- overflow line for each of the tanks. These tanks overflow to the Radwaste l Building Equipment Sumps. Opening the tanks provides for an increased l opening from the Radwaste Building (outside Secondary Containment). A <
! blank flange was installed in the overflow line at the tanks to isolate the overflow line from the Reactor Building airspace, and blank flanges were j installed in the vents to isolate each tank from the overflow in the other
- l. tank. This assured that neither of the tanks was connected to Reactor l Building ventilation (i.e. Secondary Containment), during maintenance on
[ the tanks.
Safety Evaluation Summarv i
This activity did riot increase the probability of occurrence of an accident
, evaluated previously in the SAR. Installation of the blank flanges in the L vent and overflow lines for the purpose ofisolating the Condensate Phase l Separator Tanks from the Reactor Building ventilation did not result in, or l initiate any accidents described in the SAR. Installation of these blank L flanges per the temporary modification ensured secondary containment integrity. These blanks relocated the boundary from the tank walls to the blank flanges. The blanks were made from a Quality Level 1 steel plate which is the same thickness as the walls of the tank. Additionally the
! blanks were bolted in the same configuration as the original flanges, therefore, the blank flanges were as good as/or better than the tank walls as
, a boundary. During the temporary modification, the Standby Gas
} Treatment System could still maintain Secondary Containment.
a 36 i
_..-__.m _ _ . _ _ . _ _ _ _ _ . _ _ _ _ _ . _ _ . - _ . . . _ _ _ . _ _.
- l. ,
l-r Therefore, the blank flanges did not result in or initiate any accident as !
, ' described above in the SAR This activity did not increase the ,
i consequences of an accident evaluated previously in the SAR. The l consequence of each accident evaluated previously is sufficiently extreme I L so as to envelope the consequence of any potential accident resulting from :
j this activity. Therefore, the initial conditions at the onset of any accident i
- l. evaluated previously remain valid, and the probability of occurrence of a !
l malfunction of equipment important to safety evaluated previously in the SAR is not increased. The probability of a malfunction of equipment I l important to safety previously evaluated in the SAR was not increased.
! The blanks were installed in the same location as the existing flanges.
g.
Therefore, the possibility ofleakage is the same as the original installation.
All other components (toe Reactor Building structure, dampers, ductwork, fans, filters and control hgic) of the Secondary Containment remain '
unaffected by the activity under this evaluation. This activity did not >
i increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. Establishing secondary l containment has an Unacceptable Result associated with radioactive ;
! material release. The consequences of a malfunction with the Safety Action is not increased because the activity under evaluation is not ,
expected to have an impact on unacceptable results. A malfunction which results in a loss of Secondary Containment function is not assumed in the safety analysis. The only equipment important to safety which has the potential to be impacted by this activity are the components which contribute to Secondary Containment integrity. With regard to Secondary .
Containment, the satisfaction of safety design bases for accidents requires :
that the containment retain its integrity for certain accident situations. The
! UFSAR safety analysis assumes that the safety objective of the Secondary Containment System is maintained. The safety objective is to limit the i release to the environs of radioactive materials so that offsite doses from a postulated design-br.ses accident will be below the guideline values of 10 l L CFR 100. Secondary Containment integrity is maintained as long as l internal pressure remains below the maximum allowable value of 7 inches of water. Secondary Containment integrity was maintained during the l temporary modification of the blank flanges. The temporary modification installation and calculation shows that the blanks can withstand greater '
than 7 inches of water differential pressure. Installation of this temporary modification did not decrease the differential pressure of the Secondary i Containment System, and therefore Secondary Containment remained l operational with the blank flanges installed. Therefore, this activity did not increase the consequence of a malfunction of equipment important to l
! safety evaluated previously in the SAR. None of the events related to this
. - activity initiate, or contribute to the release of radioactive material.
Therefore, this activity did not create the possibility of an accident of a different type than previously evaluated in SAR, and the blanks did not 37
_. - . - _~_.I
create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the S AR. The effect of such a small opening on the Secondary Containment did not impact the ability of the system to achieve the margins of safety described in the Technical Speci6 cations.
1 SE 97-78 ECP 1593 - IIPCI Keep-fill Modificatmn <
l Descrintion and Basis For Chance ,
1 Both the IIPCI and RCIC inject lines are to be kept full of water to the last block valve to prevent water hammer during initial startup of the system.
While either system is in the nonnal suction lineup to the Condensate Storage Tanks (CSTs) and the water level is greater than 8'-0", the l available static head will keep the IIPCI and RCIC injection lines filled I and pressurized. This level is administratively controlled. If either of the CST level instruments are declared inoperable or if the level in the CSTs were to fall below 8'-0", IIPCI and RCIC are required to be manually l aligned in a "off-normal suction lineup" (ONSL) to the Suppression Pool. l When in the ONSL, both the liPCI and RCIC injection lines are subject to voiding via draining back to the Suppression Pool. Any degree of discharge line voiding in these systems would cause both IIPCI and RCIC to be considered inoperable.
This modification provided a permanently installed available valve lineup for a pressure source to the HPCI injection line. The Fuel Pool Cooling and Cleanup Condensate Return line was identified as the most promising permanent pressure source for a IIPCI keep-fill line. This section of piping is contiguous with the common CST's return line which is kept pressurized and full of water by running at least one CRD Pump at all times. The continually full water column provides a static head differential of about 8 psi at the last block valve (based simply on elevation difference).
This modification installed a tee between two existing vent isolation valves, on the HPCI Test Line upstream of the liPCI CST Test Return Line Isolation Valve. The tee connects to the Fuel Pool Cleanup Return Line To CST's Drain Connection valve through two new piston check valves. A 1" pipe nipple and threaded cap downstream of the vent isolation valves was also replaced.
With this new piping configuration, a pressure source (Condensate return line water column) can be manually lined up to maintain a positive pressure at the last block valve while in the ONSL to the Suppression Pool. While this flow path is open, the two new piston check valves 38
- - . - . -...- - _ - --. _--- - . - . _ - -.- ~.- -.- -
L L j mitigate any leakage from the Torus to the outside environs (beyond Secondary Containment) while providing redundant pressure boundary I isolation capability. This design provides for the preservation of vent and p drain capability.
i l Safety Evaluation Summarv l I L The llPCI System provides safety functions of core and primary containment cooling under abnormal operational transients ofloss of Feedwater and offsite power. IIPCI also provides this safety function ,
under accident conditions ofloss of coolant accident (pipe break) inside l and outside primary contaimnent. The design, fabrication, and testing of this modification was performed in accordance with applicable codes and i standards. Material requirements met the same code as original l construction. Seismic Category I requirements continue to be met. These i requirements meet or exceed the original codes and standards for the plant j {
design. Furthermore, this modification affects only piping outboard of the '
Primary Containment Boundary. Thus, this change did not increase the l likelihood of a loss of coolant accident, and it did not increase the t
probability of occurrence of an accident evaluated previously in the SAR. I l The intent of this modification was simply to insure the filled condition of the HPCI Inject Line regardless of which suction lineup is in place. The safety function of HPCI is thus preserved as designed. This modification provided a redundant isolable, seismically adequate pressure boundary within the existing boundary valves and thus the designed capability to mitigate effluent leakage was maintained. The consequences of postulated accidents are bounded by analysis provided for the previously existing configuration. This modification did not increase the consequences of an accident evaluated previously in the SAR, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased. This modification did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR.
L The new configuration differs from the previously existing configuration only in that an alternate source of static head is now available during ONSL. When in the standby readiness mode and normal suction lineup, the HPCI System is normally pressurized by the static head due to normal CST inventory. While in ONSL with the modification in place, the static head is provided by the CST return line and can be guaranteed as long as
- one CRD pump is running. Furthermore, a minimum of two normally
, closed valves precludes any unanticipated drainage from the Fuel Pool Cooling and C' leanup System. This modification did not create the 39 l
l .- __ _
possibility of an accident of a different type than any evaluated previously in the SAR.
The modified configuration of the HPCI Test line v nt with the addition of the keep-fill piping is consistent with the existing plant design basis. Its intended functioa of providing an isolable Class 11 pre:sure boundary and a full flow test line is not impacted. As proven by analysis, tie new configuration is no more susceptible to failure than the previoesly existing i piping. The new check valves were procured to the same quality standards ;
as the previously existing vent valves and are equally reliable. De l installation of redundant check valves precludes unanticipated diversion of HPCI Injection in the event of a failure of a single check valve. The keep-fill lineup does not provide any unanticipated drainage path from the Fuel l Pool Cooling and Cleanup System since it is protected by two normally ;
closed valves. The potential for failure of these valves has not increased !
due to this modification. This modification did not create the possibility of a malfunction of equipment important to safety of a different type than j any evaluated in the SAR. l l
The HPCI System maintains adequate coolant inventory and provides core
{
cooling with the objective of preventing excessive clad temperatures. This l is an important barrier in the prevention of uncontrolled release of fission I products. This modification had no impact on the HPCI System's safety function of maintaining adequate coolant inventory for core woling and the system will continue to perform its intended function based on the specified low level scram and initiation setpoints. This modification did not change any level setpoints for CST / Torus suction swap. This modification augments the plant's ability to inaintain HPCI operability during ONSL. This modification did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-79 EMA For Removal Of Service Air Filter From Fuel Pool Cooling And Cleanup System Descrintion and Basis For Chance Maintenance records indicate this filter had not been serviced since it was installed until the filter housing developed a leak in May of 1994. The source of water which caused the filter housing to fail was due to the filter location and the piping configuration. If the Fuel Pool Cooling and Cleanup Compressed Air Line Check Valve, which has no record of maintenance since it was installed, was not properly seated, water could have been forced into the filter which is the low point of the system. The Service Air System supply has several filters and dryers upstream of this filter, making it unnecessary. Additionally, the Service Air quality is 40
periodically monitored. The results of sampling has consistently shown a l~ particulate level no greater than 3 microns. This modification removed the filter and housing and replaced it with a 1" section of pipe.
i l' Safety Evaluation Summarv ,
1
! This activity did not increase the probability of occurrence of an accident enluated previously in the SAR. The elimination of this filter from the air supply line to the Fuel Pool Filter Demineralizers did not increase the j probability of an accident evaluated in the SAR. The elimination of this filter did not impact operation of the Spent Fuel Pool Cooling and Cleanup System or the Compressed Air System. This activity did not increase the l consequences of an accident evaluated previously in the SAR. The l
elimination of this filter had no effect on the cleanliness of the air delivered to this system. Service Air quality is monitored and maintained I at appropriate cleanliness levels. This modification was made using ;
l approved plant procedures and materials, further assuring no adverse plant l effect. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The i elimination of this filter did not adversely impact the system operation and may improve the air quality by removing a possible source of contaminants. The line is seismic class 11 and has no nuclear seismic requirements. The reduction in weight from the removal of the filter will not adversely affect the line mounting. The service air connection to the l Fuel Pool Filter Demineralizer is used for backwashing the system and is not required for the performance of any accident function. This change i will not create the possibility of an accident of any type, and it did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This activity did not have any effect on the operation of the associated equipment. The filter and the Service Air System are not considered important to safety.
This activity did not reduce the margin of safety as defined in the basis for any Technical Specification. This activity had no effect on the system's operation, setpoints, capacity, or any of the operating modes described in the Operating License and Technical Specification.
SE 97-84 EMA For 'B' Reactor Feedwater Pump Auxiliary Lube Oil System Logic Modification Description and Basis For Chance
! This change modified the control logic for the 'B' Feedwater Pump
- Auxiliary Lube Oil System. This change allows the Operators to shut the y
! 41 l'
l I
plant down in a controlled manner in the event the Oil System would lose pressure. Previously, the logic would result in a Feedwater Pump trip and potential plant SCRAM in the event the Oil System would lose pressure due to the failure of the Shaft Driven Pump. A time delay relay was removed to allow the Auxiliary Lube Oil Pump to auto start based on system oil pressure. A seal-in contact was installed around a pressure switch which prevents the Auxiliary Lube Oil Pump fro:a cycling on pressure. A relay was installed to activate an annunciator window to indicate that the Auxiliary Lube Oil Pump has started. This alarm window will be activated any time the Auxiliary Lube Oil pump is running. This change also modified the P&lD for the Condensate and Feedwater System.
l The drawing change consisted of changing a relay equipment identification number and adding a note indicating the relay has an alarm function.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident previously evaluated in the SAR. The change to the control logic for the Auxiliary Lube Oil Pump increases the reliability of the Feedwater Control System. The logic change will allow the Auxiliary Lube Oil
, Pump to maintain lube oil pressure in the event of a failure of the shaft driven lube oil pump. This logic change did not increase the probability of an accident as described in the SAR. This activity did not increase the consequences of an accident previously evaluated in the SAR, and it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The consequences of a malfunction of equipment important to safety evaluated previously in the SAR has not increased. The change to the control logic of the Auxiliary Lube Oil Pump made the system more reliable by allowing the Auxiliary Lube Oil Pump to supply lube oil in the event the Shaft Driven Pump fails.
This activity did not create the possibly of an accident of a different type than any evaluated previously in the SAR and it did not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR. The Feedwater Control System is not discussed in the Technical Specifications. Therefore, this activity did not rede.ce the margin of safety as dermed in the basis for the Technical Specifications.
42
l l
l SE 97-86 ECP 1595 - Modification To Line Between Main Steam Isolation Valves
! Description and Basis For Chance On September 9,1997 a leak was identified on the weld between a reducing insert and the piping on the 'A' Main Steam Line (M3L) branch connection immediately upstream of the outboard Main Steam Isolation Valve (MSIV). A suspected cause for the leak was fatigue failure of the weldjoint due to high cycle fatigue from normal system operation. The previous configuration included three valves located approximately three feet from the centerline of the Main Steam Line and oriented j l
perpendicular to the centerline (a fourth valve was in the vertical leg). !
This configuration was cantilevered from the MSL and likely produced the cyclic loading necessary to result in the failure of the weld.
This modification changed the piping configuration to eliminate the unnecessary valves in the configuration, and the remaining components i are now in a vertical branch from the MSL thereby minimizing the effect from the cantilevered configuration. This modification was installed on all four main steam lines.
Safety Evaluation Summary The probability of occurrence of an accident previously evaluated is not increased since this modification was evaluated and it conforms to existing design requirements. In fact, the probability of an accident has been decreased because the piping configuration has been improved to be less susceptible to fatigue failure. The consequences of an accident are unchanged as the design uses the same size components so the possible break size is not increased and the release path is not changed. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The modified design uses equipment of the same size, and of similar function to the previously existing equipment. The configuration was seismically analyzed and meets acceptance standards. In fact, the probability of a malfunction has been decreased because the piping configuration has been improved to be less susceptible to fatigue failure. The modified configuration has only a passive safety function. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and the possibility of an accident of a different type than any evaluated previously in the SAR has not increased. The possibility of a malfunction of equipment important to i safety of a different type than any evaluated previously in the SAR was
( not created. The margin of safety as defined in the basis for any Technical 43
i l
Specification has not been reduced. In fact, the margin of safety has been increased since the improved piping configuration makes the piping less susceptible to fatigue failure. The margin of safety for the Primary Containment is not reduced.
SE 97-89 EMA For Installation Of Circulating Water Corrosion Monitor System Description and Basis For Chance This change installed a new Corrosion Monitor System for the Circulating i Water System in place of the decommissioned General Service Water Sampler System. The sample point is the'same one used previously for monitoring the General Service Water System and Circulating Water. The discharge path is the same path that was used previously by the General Service Water Sampler System. The flow rate required by the previously used GSW Sampler System was one to two gpm, while the new Corrosion Monitor System requires three to five gpm. The new system discharges to the same floor drain that the General Service Water Sampler System utilized. A previously decommissioned drain isolation solenoid valve will now close upon low header pressure to preclude header drain down upon a General Service Water pump trip.
Safety Evaluation Summary None of the accidents previously evaluated are affected by this activity.
The systems involved were the General Service Water System and the Circulating Water System, which are not safety related systems. The Corrosion Monitor System was installed in the same location as the original General Service Water Sampling System and discharges to the
, same location. The piping to the Corrosion Monitor System meets or l- exceeds the design temperature and pressure rating required by the system.
- i. There was no impact on the overall performance of either the General j Service Water System or Circulating Water System. The flow required by l the new Corrosion Monitor System is three to five gpm and is slightly higher than the original sample system flow rate of one to two gpm. The
~
increase in flow is within the capacity of the General Service Water pumps and the floor drain piping. The installation of the corrosion monitor on the discharge header of the General Service Water System did not impact the operation of any safety related equipment. Therefore, this activity did not increase the probability of the occurrence of an accident previously
! evaluated in the SAR, and the consequences of an accident evaluated previously in the SAR are not increased. The General Service Water System provides cooling to non-safety related equipment. The Circulating Water System provides cooling to the main condenser. These two systems ,
44
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l I
are important to power production, however, they do not provide a safety function and they do not provide cooling to any equipment which provides a safety function. The probability of occurrence of a previously evaluated ;
malfunction to equipment important to safety is not increased, and the consequences of a previously evaluated malfunction to equipment important to safety is not increased. There is no credible possibility of an accident or malfunction of a different type occurring due to this activity.
There are no Technical Specifications for either the General Service Water or Circulating Water System. There are no components that require I L General Service Water or Circulating Water to maintain a margin of safety as defmed in the basis for any Technical Specification. This activity did not reduce the margin of safety as dermed in the basis for any Technical Specification.
SE 97-95 EMA To Remove KAMAN 12 Trip Function From Various Fans i Description and Basis For Chance The KAMAN 12 trip function was removed from the following fans l
associated with the Low Level Radwaste Processing and Storage Facility (LLRPSF) ventilation system:
. LLRPSF Future Expansion Area HVAC Exhaust Fan
- LLRPSF Future Expansion Area Vent Supply Fan
- LLRPSF HVAC Equipment Room Exhaust Fan l
- LLRPSF HVAC Equipment Room Vent Supply Fan
!
- LLRPSF Processing Area HVAC Exhaust Fan l
- LLRPSF Processing Area IIVAC Recirculation Fan
!
- LLRPSF Processing Area Vent Supply Fan e LLRPSF Storage Access Area Exhaust Fan l
- LLRPSF Storage Area Vent Supply Fan A previous change added the trip function in the LLRPSF Ventilation l System so that upon receiving ' Gas Rad Hi,' ' Gas Rad Alert' or ' Gas Rad Fail' signal the HVAC system for the LLRPSF would isolate preventing radiation release to the environment. This trip was provided as an j' engineering design conservatism and not due to any regulatory requirement. The LLRPSF ventilation system isolation increased the potential for system freezing during winter months. Also this trip became i an issue during the hot summer months when the LLRPSF ventilation system isolated for any length of time which made conditions in the LLRPSF unpleasant. Although the operability of the KAMAN system monitors was improving, spurious alarms and equipment failures did 45
~. ~. - - . . - - -- _ . - - - - .. - . - . .. ~ . . _ _.. - . . - - . . - - . . _ . .
n e
4
, occur. Having to be concerned about LLRPSF ventilation while recovering from these trips unnecessarily complicated the effort. Since
}'
elevated levels of radiation are not expected in the LLRPSF ventilation system, no accident range monitor is installed. Therefore, the trip function l practically did not serve any useful function and was removed from the LLRPSF ventilation system. The capability of the LLRPSF ventilation
, system to isolate on detection of a fire was not compromised.
f Safety Evaluation Summary None of the accidents discussed in the UFSAR attribute LLRPSF 4
ventilation system or the LLRPSF radiation monitor. Except for some termination lugs and wire, no new components are added by this activity.
Because this activity did not affect the overall system performance in a ;
L manner that could lead to an accident, the probability of occurrence of an
' i accident previously evaluated in *he SAR was not increased. Functions of f j the radiation monitor to provide annunciation in the control room and i
information to the Plant Process Computer (PPC) before and after the j implementation of this change remains unaffected. Though the system description in the UFSAR was revised, none of the bases described earlier was adversely affected by this activity. Radioactive effluent control required by 10 CFR 50.36a,40 CFR Part 190,10 CFR 20.1302, and 10
- CFR 50 Appendix I, is maintained. Because none of the functions were
{
- altered by this activity, the LLRPSF ventilation and the radiation
- monitoring systems still operate the same. This activity did not increase t the consequences of an accident evaluated previously in the SAR. The modified circuits involve wiring changes to the LLRPSF Ventilation l System fans to remove radiation monitor interlock for isolation on 4
elevated radiation level or its failure. The fans and the radiation monitor 1 '
associated with this change are identified as Quality Level 4. Neither the j LLRPSF Ventilation System nor the radiation monitor perform any safety
- function. The functions of providing annunciation and data to the plant process computer were not adversely affected by this activity in any way. ,
Since the modified circuits for the LLRPSF Ventilation System meet the i original system bases, electrical separation, Appendix R, seismic and environmental requirements, the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased. Since no new active components were added by i this modification, no new failure modes were introduced in the system.
Therefore, the affected circuits and detector remain operable in all design ,
basis event environments. This activity did not increase the consequences ;
of a malfunction of equipment important to safety evaluated previously in the SAR. The new passive components; lugs and wires, added by this ,
modificetion by themselves can not cause any accident. Thus, no failure can be postulated by these changes to create an accident of a different t
46
l type. This activity did not create the possibility of an accident of a
- j. different type than any evaluated previously in the SAR. Since no new
! active components are added by this modification, no new failure modes I
are introduced in the systeni. This activity did not create the possibility of -
L a malfunction of equipment important to safety of a different type than any l previously evaluated in the SAR. The LLRPSF ventilation system and L LLRPSF radiation detectors are not in the DAEC Technical Specifications. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-96 EMA For Reactor Recirculation Pumps' Motor Watt Transducers Replacement Description and Basis For Chance
! The output of the transducers was not representative of actual
! Recirculation Pumps' Motor power over the entire operating range of the -
Recirculation Motors. The watt transducers degraded to the point that they were not accurate at any frequency. The Recirculation watt computer point indications were accurate e the normal operating frequencies
! because the computer points' range was increased by a factor of 1.4 (40%)
to compensate for the transducers' inaccuracies. The Recirculation Pumps' Motor watt indications at the Control Room panel did not agree l with the computer points nor were they representative of actual l Recirculation Pump Motor watts. The watt transducers were replaced with .
i variable frequency type transducers, which are accurate over the entire
! operating range of the Recirculation Pumps' Motors.
l-The replacement watt transducers are not self-powered, therefore a separate 120 VAC power supply was utilized in the new transducers installation. The replacement watt transducers utilize variable frequency current transducers. The current transformers are now located inside the junction box on the side of the Recirculation Pumps' Motor-Generator sets. The new watt transducers are located on the side of the Recirculation Pumps' Motor-Generator sets.
Safety Evaluation Summary This modification was an improvement in technology and design. This modification met the design, material, and construction standards of the previously installed watt transducers. The currently installed watt transducers meet and or exceed all design and material requirements of the previously installed watt transducers. Because this activity did not affect
l the overall system performance in a manner that could lead to an accident,
]
t the probability of occurrence of an accident previously evaluated in the p
47
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SAR is not increased. None of the accidents previously evaluated were
~a ffected by this activity. The addition of the watt transducers' load to
. Instrument AC had no adverse effect on Power System Configuration and Analyses (PSCA). Starting currents, and total continuous loading of eqtipment were within the ratings of the existing configuration. These chat ges could not increase the consequences of an accident evaluated predously in the SAR. The structures, systems and components afTected g by this change are associated with the Recirculation and the Instrument 1-
' AC Systems. The portion of the systems drected by this modification is not safe;y related. Since the replacement equipment does not affect any original system bases, electrical separation,10 CFR 50 Appendix R, i
seismic and environmental requirements, the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased, and the consequences of a malfustion of equipment important to safety evaluated previously in the SAR is not increased. This
- activity did not create the possibility of an accident of a different type than L any evalmted previously in the SAR. There is no credible possibility of an accident of a different type occurring due to the replacement of the i Recirculation Pumps' Motor watt transducers. There are no new failure modes introduced that could cause an accident not already described in the SAR. This activity did not create the possibility of a malfunction of equipment important to safety of a different type other than any previously evaluated in the SAR. There are no Technical Specifications for the Recirculation Pumps' Motor watt transducers, nor are there any Technical Specifications affected by this activity. This activity did not reduce the margin of safety as defined in the basis of any Technical Specification.
SE 97-99 Temporary Modification To Measure Well Water Flow To Control i
Building Chillers Descriotion and Basis For Chance This modification installed a differential pressure gage at the Bell and Gossett circuit sensor near the Cooling Water Bypass Valve to measure i Well Water flow w the Control Building Chillers. This enabled the flow in this branch to be adjusted using the Bypass Line Throttle valve as needed to provide proper flow to the Chillers. The Well Water System
- P&ID required flow through the Main Plant Intake Coil Bypass line to be
- . _ adjusted to 310 gpm. This temporary modification allowed flow to be adjusted as needed for optimal Chiller operation.
Safety Evaluation Summary The installation of the differential pressure indicator to monitor flow, and adjusting the well water flow rate to the Control Building Chillers 48
improved the performance of the Control Building Chillers. This
- modification did not affect system performance in a manner that could lead to an accident, therefore the probability of an accident previously
' evaluated in the SAR was not increased. The Well Water System is not required for the mitigation of any accidents defined in the NSOA and UFSAR. This activity did not increase the consequences of an accident evaluated previously in the SAR. There was no impact on the Emergency Service Water System's capability to provide post accident cooling for the Chillers. The Chillers continued to fulfill their design function. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. This change did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No failure can be postulated by this change that would create an accident of a different type.
r The possibility of an accident of a difTerent type than any evaluated previously in the SAR is not created by this activity, and this change did not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR. The Control ,
Building Chillers were still operable per Technical Specifications with the '
temporary modification installed. The margin of safety as defined in the basis for any Technical Specification was not reduced.
SE 97-100 Generator Rewind Project Temporary Modification l
i Description and Basis For Chang The purpose of this modification was to provide temporary power for the Generator Rewind Project during Refueling Outage 15. This modification disconnected a load center bus, which normally provides power to the Instrument Air Compressors, from its normal power source in order to connect a temporary high voltage cable to a disconnect switch. This temporary cable supplied 4160 VAC to a temporary 4160 VAC - 480 VAC transformer located outside the Turbine Building near the Main Transformers. Power was then routed via temporary cabling to the Turbine Operating Deck. The loads on the load center bus remained energized by a cross-tie breaker.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The loss ofinstrument air is not rui initiator of any of the design accidents in the plant. This modification did not affect any of the inputs considered in the accident analysis of the UFSAR or the NSOA. This change did not alter the interface between plant electrical systems. 'inc vtivity did not increase the consequences of 49
_~ , _ , ._. _ _ _ . _ . _ . _ _ _ . _ _ _ . .. - _
an accident evaluated previously in the SAR. Instrument Air from the Compressor Building is not relied on for recovery of an accident previously evaluated in the USFAR. There were no increases in the radiological consequences of any previously analyzed UFSAR accident.
The changes made by this modification did not change, degrade or prevent actions described in the UFSAR. The probability ofoccurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. This change had no impact on systems, structures or components important to safety. Providing temporary power from the Instrument Air Compressor's power supply does not degrade the system's ability to perform its design function. The Instrument Air System has no safety significance in the UFSAR. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. Both the Instrument Air System and the 4160 VAC line that supplies power to the Instrument Air Building are Quality Level IV. A total failure of the Compressed Air System could not affect the ,peration of safety related equipment. The possibility of an accident of a different type than any evaluated previously in the SAR was not created. The temporary power cenfiguration eliminated a redundant power source to the Air Compressor Building. The cross-tie breaker maintained all the compressors operable. The Air System and its power supply are not safety related and could not create an accident of a different type. This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The margin of safety as defined in the basis for any Technical Specification was not reduced. There are no Technical Specifications associated with Instrument Air or the 4160 VAC offsite power feed to the Compressor Building.
SE 97-103 EMAs For Replacement Of Condensate Demineralizer Precoat Return Valves Description and Basis For Chance Maintenance records indicated the Condensate Demineralizer precoat return butterfly valves required excessive service and did not provide adequate isolation to prevent the precoat/ backwash recirculation header from pressurizing when not in service. This leakage could have potentially damaged this line which had a 150 psig design rating. The line was protected with a thermal relief valve, which was damaged due to constant opening. The relief valve was replaced during Refueling Outage 14 with a similar valve, after the butterfly valves were serviced. The butterfly valves continued to show signs ofleakage and have been replaced with full port ball valves and new actuators.
50
Safety Evaluation Summary The replacement of the existing butterfly valves and actuators did not increase the probability of an accident evaluated in the SAR. This Condensate Cleanup System is considered as part of the Steam and Power Conversion System. All of the Cleanup System components affected by this change were located outside of the Drywell and did not contribute to any accident evaluated in the SAR. This activity did not increase the consequences of an accident evaluated previously in the SAR. The Condensate Demineralizers are not considered in any accident analysis.
The replacement of these valves and operators with items which meet the design requirements did not have any etTect on any of the evaluated accidents. This activity did not increase the probability of occurrence of a ;
malfunction of equipment important to safety evaluated previously in the SAR. The Condensate Demineralizer System is not considered important '
to safety. The precoat retr.rn lines are seismic class 11 and have no nuclear seismic requirements. This activity did not increase the consequences of a malfunction of equipmer.t important to safety evaluated previously in the I
SAR. The activity associated with this EMA did not challenge the installed equipment. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR.
This change did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This activity was performed using approved plant procedures and I materials that met existing design standards. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification. This activity had no effect on the system's operation, setpoints, capacity, or any of the operating modes described in the Operating License and Technical Specifications.
SE 97-108 ECP 1597 - Installation Of Permanent Cross-tie Between Well Water and Reactor Building Closed Cooling Water (RBCCW) IIcat i Exchangers Description and Basis For Chance The DAEC maintained temporary cross-tie capability between Well Water i supply and the normal General Service Water (GSW) supply to the ' A' and 'C' RBCCW lleat Exchangers. The cross-tie utilized PVC temporary piping stored in situ and assembled each Refueling Outage to augment RBCCW capability during extended GSW out-of-service (OOS) windows.
The plant had to isolate and drain the tube sides of both the heat exchangers to connect the cross-tie piping, an evolution that cost critical refuel outage resources. Additionally, there were several leaks that were remedially repaired in the PVC temporary piping sections. A joint failure 51
_ . . . ..-.m __ ,_._ _ . _ . - _ . _ . _ _ _ _ _ _ _ _ . _ _ _ _
j in the unqualified plastic system while in use could have caused a major j disruption of any refuel outage schedule.
This modification replaced the temporary PVC piping that was hanging above the RBCCW Heat Exchangers with permanently connected welded carbon steel piping, The permanent connections were made to the 'A' and l 'C' heat exchangers with pipe routing identical to the temporary PVC piping. The supports were fully qualified for permanent installation with minor modifications. Removable spools ( similar to those in the temporary configuration ) as well as several new vent and drain points
. were provided as additional design features to allow for the periodic cleaning of the tube-side waterboxes on the 'A' and 'C' heat exchangers.
New butterfly isolation valves were installed on the inlets and outlets of the waterboxes as boundary valves. Similar butterfly valves were used to i
replace the PVC valving. The new valves provide for maximum l operational flexibility.
l Safety Evaluation Summary L
l This change did not decrease the effectiveness of any of the safety systems in the plant and could not cause an accident or increase the likelihood of an accident. The probability of an occurrence of the accidents discussed in L the UFSAR and NSOA is based on initial conditions and assumptions ,
i which do not depend on the end use ofor interactions with the GSW or l RBCCW systems. The UFSAR discusses the consequences of Well Water l leaks in the Drywell Cooling Loop but this moditication had no impact on that evaluation. Therefore, this activity did not result in a condition which j increased the probability of occurrence of an accident previously evaluated in the UFSAR. Well Water, GSW or RBCCW are not relied upon for recovery from an accident previously evaluated in the UFSAR. Since the l Well Water to GSW cross-tie was designed as an extension of the Well Water System, this would also be true ofit. There was no increase in the radiological consequences of any previously analyzed UFSAR accident.
[ This activity did not alter any assumptions previously made in evaluating the radiological consequences of an accident, nor did it play a role in mitigating the radiological consequences of an accident described in the UFSAR. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The modifications to the systems made by this activity had no impact on systems, structures or components important to safety, nor did they degrade the systems' abilities to perform their design function. The
, three systems affected are all non-safety related. A pennanently available i
! capability to cross-tie between Well Water and the GSW heat sink for the RBCCW did not degrade the existing performance of any of the three
L i i l l
l l
L normally isolated by capable valves. The primary use of such a capability would be during GSW OOS maintenance windows during refuel outages.
Although the cross-tie could also be made available to augment the GSW heat sink for RBCCW during on-line scenarios, the normal Operating Instructions for Well Water would not permit this without further l evaluation. The diversion of Well Water to the RBCCW heat exchangers during GSW OOS windows within refuel outages minimizes the impact of such a diversion on the plant. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The permanent cross-tie between Well Water and l the GSW heat sink for RBCCW augments the capabilities of the three ,
systems. Since there we.: no safety related components either below or in I the immediate vicinity of the new piping runs, a Seismic Category II i installation was adequate. This activity did not create the possibility of an ;
accident of a different type than any evaluated previously in the SAR.
l The new piping meets the specification for the contiguous Well Water i System which is the more severe of the two systems (Well Water and i GSW). The heat sink capability of RBCCW is augmented by this modification. Inadvertent leakage between the Well Water System and RBCCW would be negligible. Since Well Water is the higher pressure j system, any leakage would be from the higher quality system to the lower quality system. This activity did not create the possibility of a t malfunction of equipment important to safety of a different type than any l
evaluated previously in the SAR. There are no Technical Specifications L
associated with the systems affected, therefore this activity did not reduce the margin of safety as defined in the basis for any Technical ;
Specification. l i
SE 97-110 EMA For Replacement Of Moisture Sepe= tor Reheater First Stage I Reheat Steam Drain Valve i
Descriotion and Basis For Chance This valve was a Velan globe valve which experienced problems with !
vibrating offits seat during plant operation. The valve was replaced with an Anchor Darling double disc gate valve. The double disc gate valve
]
improved the isolation leak tightness and therefore improved the system l
perfonnance. This activity also required that the stem nut be replaced due ;
to a change in the stem diameter. The valve is not located over any I seismic related pipes or equipment. The gate valve is also taller than the ;
globe valve but it does not interfere with any other equipment. i h
s
! 53 i
Safety Evaluation Summary This activity did not increase the probability ofoccurrence of an accident evaluated previously in the SAR. None of the accidents evaluated in the SAR are affected by this valve. The probability of occurrence of an accident is unchanged. This activity did not increase the consequences of an accident evaluated previously in the SAR, and it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This valve does not impact the ability of a structure, system, or component important to safety to perform its intended function; therefore, this change had no adverse impact on a safety function, and the consequences of a malfunction of equipment important to safety were not increased. The possibility of an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created.
There are no Technical Specification requirements for this valve. This valve is not a contributor to or initiator of any of the accidents evaluated in the UFSAR. This change had no impact on any Technical Specifications, or safe operation of the plant. Therefore, no margin of safety was reduced.
SE 97-116 ECP 1602 - Control Building Chillers' Fast Unload Modifications Description and Basis For Chance The Control Building Chillers did not have an automatic system to limit compressor load based on compressor discharge pressure. This resulted in Chiller trips when the load exceeded the capacity of the service water supplied to the condenser. The machines were equipped with a fast unload circuit that actuated to unload the machine on low suction pressure to prevent the machine from tripping on a freeze protection safety switch. To improve the system availability, a pressure switch was installed in the refrigeration system of each Chiller to actuate the fast unload logic prior to the compressor high discharge pressure trip. This action also alarms in the Control Room.
Safety Evaluation Summary The Control Building ventilation system is not an accident initiator.
Modifying the controls on the Chillers did not have any afTect on accident initiating systems. The new switches added by this activity were procured safety related, mounted seismically and meet current design, material, environment suitability and construction acceptance criteria. None of the accidents evaluated in the SAR were affected by this activity. This modification did not increase the probability of an accident evaluated previously in the SAR. The Control Building Ventilation is a support l 54
system. This modification improved the availability of the Chillers. In the event of an imbalance between the condenser water available and the load on the machine, the machine will run at a reduced capacity instead of tripping and providing no cooling. This activity did not have any affect on the radiological consequences of any accident analysis described in the SAR. This modification did not increase the consequences of an accident evaluated in the SAR. The seismically installed switches are highly reliable and inadvertent operation of their contacts is highly unlikely.
Since the changes meet original system bases, electrical separation, physical separation. Appendix R, seismic and environmental requirements, this modification did not increase the probability of a malfunction of equipment important to safety evaluated in the SAR. The consequences of a Chiller malfunction was not increased. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The new pressure switches added by this modification by themselves can not cause any accident. Thus, no failure can be postulated by these changes to create an accident of a different type. The possibility of a malfunction of equipment important to safety of a different type that previously evaluated in the SAR was not created.
The availability of the Control Building Chillers was enhanced by eliminating the unnecessary tripping on high discharge pressure.
Therefore, this activity did not reduce the margin of safety.
SE 97-118 Temporary Modification To Repair Leak In Drain Line To Main Condenser Description and Basis For Chance A leak was identified on an elbow on the High Pressure Coolant Injection System (HPCI)/ Reactor Core Isolation Cooling System (RCIC)/Offgas Drain line to the Main Condenser. This leak was on the first elbow upstream of the branch connection of the drain line to the main steam drain line. The leak was through the body of the elbow and located where steam impingement on the inside of the elbow would likely occur. The leak was located approximately six feet from the condenser. The leak l
needed to be kept small enough that condenser vacuum was able to be '
maintained to keep the plant operating. There was no isolation betwen !
the leak and the condensen The elbow could not be replaced unless n ;
isolation was installed F
<en the leak and the condenser or unless the plant was shutdown.1 he hmporary Modification required for repair !
involved welding a branch connection to the leaking elbow. This configuration also included a valve on the connection to depressurize and control the leakage during the welding process.
l I
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l l l The branch piping also functioned as a leakage barrier for the Main Steam Isolation Valve -Leakage Treatment System (MSIV-LTS). The function of this system is to direct MSIV leakage post-accident to the condenser for treatment. The system function is safety related as it functions to mitigate
- the radiological consequences of an accident that could result in potential offsite exposures comparable to the dose reference values specified in 10 CFR Part 100. The branch piping acted as a boundary to ensure post accident releases were within allowable levels.
Safety Evaluation Summarv i-The NSOA review identified several accidents, transients, or special events that could have been affected by the modification. These included LOCA inside containment, LOCA outside containment, and Control Rod drop accident. These accidents rely on the piping for pressure boundary integrity. The probability of occurrence of an accident previously evaluated was not increased since the modification was evaluated as seismically adequate and it reestablished the pressure boundary to design requirements. The consequences of an accident were unchanged. The -
modified design used equipment that met acceptance standards. The modified configuration had only a passive safety function. The consequences of a malfunction of equipment were unchanged. The l consequences of a malfunction were limited by the piping size which was unchanged. The possibility of an accident of a different type was not created. The possibility of a malfunction of equipment of a different type was not created. The margin of safety as defined in the basis for any Technical Specification was not reduced and the margin of safety for the Primary Containment was not reduced. The radiological calculations used to demonstrate compliance with 10 CFR 100 and 10 CFR 50 limits were unchanged with this modification.
SE 97-119 EMA For Main Generator Stator Rewind Description and Basis For Chance The rewind of the Main Generator stator during Refueling Outage 15 was in response to problems associated with stator water leaks and various tests that identified stator bar degradation over the last several years.
Documentation of the stator bar water leak problem in a generic sense can be found in GE TIL 1098. General Electric has a high degree of l confidence that the new stator bar design is impervious to the corrosion and leakage problems associated with the previous manufacturing method.
7 DAEC requested new stator bars to be manufactured and installed by GE
( to replace the degraded bars with the provision that the machine capability rating of the rewound machine not be less than the previously installed 56 l
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-v - - . . - ,. -
machine. This modification also involved the addition of a core monitor hydrogen gas analyzer panel.
The Main Generator stator rewinding occurred during Refueling Outage
- 15. The scope of this activity included generator disassembly, generator field removal, generator wedge removal, and stator winding removal. The core iron was inspected and cleaned in preparation for the new stator bars.
New stator bars were installed, secured with side ripple springs, rewedged, and connected together by brazing bottle clip style fittings that combine the mechanical cooling and electrical conductor functions of the bars.
Extensive generator testing was performed to verify stator bar hydraulic integrity and electrical insulation quality. The final processes involved baking the conformable materials used to secure windings to a cured state and bump testing the end winding support system. The majority of this work took place on the Turbine Building operating deck with some support activity and work on auxiliary cooling systems taking place on the floor directly below the generator in the Turbine Building. Some minor changes were made to the stator water cooling system to make flow adjustments per the requirements of the new stator bars.
The main generator supplies AC power to the plants auxiliary power transformer and to the offsite power grid load. The new stator winding is more efficient than the previous winding, having benefited from technological advances in material and design over the years. However, these improvements did not change the operational capability of the machine significantly. Various machine electrical resistance and reactance values are now slightly different with the new stator bar design. The minor differences associated with the reactance values did not result in any significant change to the operating characteristics of the machine.
The new stator bars are shaped exactly like the original stator bars thus guaranteeing proper fit in the DAEC generator core slots. Improved insulation materials used on the bars and between the strands in the bars allow a mixture of solid and hollow strands, versus all hollow strands, to be used resulting in a more efficient bar requiring less stator water cooling.
Factory testing of the bars for leakage and high voltage verified the new bars exceed original standards.
The core monitor hydrogen gas analyzer equipment monito'rs the hydrogen gas passing through the machine for the presence of pyrolosate material l which flakes off the stator core when high temperature conditions are t
reached. Monitoring equipment taps into the hydrogen sample lines that run through the Generator Hydrogen and Cooling Water Control Panel. A trouble alarm function was added to this panel.
57 i
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Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. Replacing the stator winding on the main generator did not cause the system to operate in a different manner.
The main generator is not a nuclear safety related piece of equipment nor is it considered important to safety. The change to the generator stator winding did not decrease the effectiveness of any safety system in the plant and it did not increase the likelihood of an accident. The main generator has the same capability rating and essentially the same capability curve as before. New generator resistance and reactance parameters resulted from a more efficient stator bar design and improved computer based design calculation methods used to model the new bar design. All changes to the bar design and the bar placement within the generator core slot and end winding configuration were within design limits. Support systems, such as stator water cooling, used to cool the main generator during operation were modified to reduce flow slightly in order to accommodate the new more efficient stator bar winding. Stator water cooling alarm and runback setpoints were adjusted to accommodate for the modified flow requirements. No cooling system piping is operated at higher than rated pressures as a result of this modification. The probability of occurrence of an accident previously evaluated in the SAR was not increased. Replacing the main generator stator winding and the addition of the core monitor equipment did not have any radiological consequences on the plant or affect any UFSAR accident evaluations. The changes made by these activities did not change, degrade, or prevent actions described or assumed in an accident analysis discussed in the UFSAR. No radiological boundaries changed as a result of these modifications, therefore, these activities did not increase the consequences of an accident evaluated previously in the SAR. The main generator and its support systems are all non-safety related and are not considered equipment important to safety. Existing protective relay settings were reviewed and determined to be adequate for the rewound generator. The modifications did not have any impact on systems, structures or components important to safety, nor did they degrade any safety system's ability to perform its design function. Therefore, no increase in probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR occurred. Existing generator protective relay settings were reviewed and found to be adequate for operation with the modified generator resistance and reactance parameters.
The stator water cooling alarm and runback setpoints were adjusted to accommodate the modified flow requirements resulting from the mixed hollow and solid strand arrangement of the new stator bars. An annunciator has been added to the Generator Hydrogen and Cooling Water Control Panel for the purpose of waming operators when the generator 58 ,
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i core temperature is too high. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The changes made to the main generator stator {
winding did not create the possibility of an accident of a different type I than any evaluated previously in the SAR. The possibility of a malfunction of equipment of a ditTerent type than any evaluated previously in the SAR was not created. There are no Technical Specifications associated with the main generator, thus no margin of safety was reduced by this activity.
SE 98-03 EMA For Installation OfIsolation Valves in The Electrohydraulic Control System Description and Basis For Chance The reason for this activity was to improve the ability to calibrate the Turbine Control Valve pressure switches by installing isolation valves and calibration valves in the Electrohydraulic Control (EIIC) instrument lines to the Control Valves Emergency Trip Low Oil Supply Pressure Switches.
These valves allow the calibration of the pressure switches without securing and draining the EHC system. This reduces the amount of time and coordination required to perform the calibration and testing of these pressure switches. This change installed a quarter turn ball valve between the Fast Acting Solenoid Valves on the Turbine Control Valves and the Reactor Protection System (RPS) pressure switches. Another valve was also installed at the pressure switch to allow performing the calibration of the pressure switch after it is isolated. The main concern with installing an isolation valve in this instrument line for calibration was the possible effect on the response time of the instrument. To ensure that the response time was not adversely afTected by this EMA, changes to the tubing configuration were kept to a minimum and no new flow restrictions were put into the system that could increase the response time. When the quarter turn ball valve is open, it provides a flow path that is larger in area than the existing tubing. This ensures that no flow restrictions were added to the tubing. The tubing was altered to allow the installation of the new valves. The calibration / test valves are considered to be part of the tubing volume because the valves are normally closed and the inlet area adds a small amount of volume to the tubing configuration. To reduce the impact on the response time, the tubing modifications were kept to a minimum.
To ensure the effect on the response time was minimal and to ensure the response time remained within the required 30 milliseconds, the l surveillance test was completed after the modification.
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Safety Evaluation Summary All of the accidents and events previously evaluated in the UFSAR were reviewed with respect to this activity. The installation of the isolation valves for the pressure switches that sense the decrease in EIIC pressure as part of the turbine control valve fast closure did not increase the probability of the occurrence of an accident previously evaluated. This activity only afTected the pressure switch operation required in the event of turbine control valve fast closure. There is no increase in the probability of the occurrence of an accident caused by this activity. A single failure of one of the instrument lines to the pressure switches will not cause an accident. This activity did not increase the consequences of an accident previously evaluated. The major parameter of concern with this activity is the response time of the pressure switches associated with the decrease in EliC pressure. This signal must be provided to both the RPS logic and the Recirculation Pump Trip (RPT) End-of-Cycle (EOC) logic within 30 milliseconds. The orifice size of the valve is greater than the tubing between the fast acting solenoid valve and the pressure switch. This ensures that there is no flow restriction in the line that will result in the response time increasing. To allow for the installation of the isolation valve, the tubing was lengthened. The length of the tubing was as short as possible. This ensured that the response time was not adversely affected I by the increase in tubing length. After the modification, the response time was tested to ensure the 30 millisecond time response was met. The installation of an isolation valve does not prevent the pressure switches from meeting the single failure criteria for either RPS or RPT-EOC.
Therefore the consequences of an accident were not increased, and the installation of the isolation valve did not increase the occurrence of the probability of a malfunction of equipment important to safety. A review of the failure modes of the pressure switches revealed no credible failure mode in which the installation of the isolation valves could make the consequences worse for any accident described in the UFSAR. No credible failure regarding the increasing of the response time could be identified that would be worse than closing the isolation valve completely.
Therefore, the closure of an isolation valve in the EllC line between the fast acting solenoid valve and its associated pressure switch will not prevent a scram or a recirculation pump trip. The installation of the isolation valve did not increase the consequences of a malfunction of equipment important to safety, and the possibility of an accident of a
! different type than any evaluated previously in the SAR was not created.
There is no possibility for the installation of the isolation valves to create a malfunction of any equipment important to safety other than any previously evaluated in the SAR. The margin of safety as defined in Technical Specifications was not reduced.
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SE 98-06 EMAs To Change Setpoint Of Residual Heat Removal System Rooms' Ambient Temperature Switches Description and Basis For Chance The purpose of this activity was to change the high temperature alarm setpoint for the Residual Heat Removal (RHR) Rooms' ambient temperature switches. This change lowered the alarm setpoint to within the design limitations of the equipment within the RHR rooms.
Previously the alarm setpoint was 200 degrees Fahrenheit, while the equipment in the room is generally rated at 140 degrees. The alarm setpoint was lowered to 130 degrees. Additionally, the lowered alarm l
setpoints were incorporated into the Emergency Operating Procedures 1 (EOPs).
I Safety Evaluation Summary '
This activity did not increase the probability of occurrence of an accident I evaluated previously in the SAR. The temperature detection and alarm system does not affect the probability of the initiating accident (pipe break). Lowering the alarm setpoint did not affect the probability of a I LOCA in either corner room while the plant is entering or is in the shutdown cooling mode. Additionally, revising the EOPs to incorporate the lower setpoints did not have any impact on accident initiators because EOP actions are not accident initiators. This activity did not increase the l consequences of an accident evaluated previously in the SAR. This change lowered the detection threshold for a break in a RHR line. 1 Changing the alarm setpoint did not increase the radiological consequences of any accident previously evaluated. Additionally, revising the EOPs to incorporate the lower setpoints did not have any adverse i impact on the consequences of an accident because the EOP actions are designed to decrease the consequences of an accident and the lower setpoints do not prevent the necessary EOP actions. In fact, the lower I setpoints cause the EOP actions to be performed more conservatively.
This activity did not increase the probability of occurrence ef a malfunction of equipment important to safety evaluated peviously in the SAR. The setpoint change is within the design range of the instrument, no l hardware alterations were required. This modification did not increase the I likelihood of a malfunction of any component (safety or non-safety).
Since there is no trip / isolation function associated with these instruments, there was no increase in the probability of an inadvertent (premature) SSC l isolation associated with this change. The incorporation of the lower alarm setpoints in the EOPs reduces the probability of occurrence of a malfunction of the equipment contained in the subject rooms because the 1
temperature switches do not provide any automatic actions and the lower 61
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setpoints alert operators earlier to potentially harmful room temperatures.
This allows for earlier corrective actions to be taken to mitigate the effects on the equipment. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This modification did not increase the consequences of any safety related malfunction. The incorporation of the lower alarm setpoints in the EOPs did not increase the consequences of a malfunction ofequipment because the lower alarm setpoints do not directly affect any equipment.
Any equipment which the EOPs make use of to mitigate the consequences of an accident remained available. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. This change allowed for earlier corrective actions to be taken to mitigate the effects on the equipment. Based on a review of Technical Specifications and the DAEC UFSAR, there was no impact on margin of safety as defined for the lower alarm setpoints or the operator actions contained in the EOPs.
SE 98-07 EMA For Installation OfInsulating Board and Cooling Air Duct For Peak Recording Accelerograph Descriotion and Basis For Chance The Peak Recording Accelerograph (PRA)is a monitor that measures the magnitude of a seismic event on the Reactor Pressure Vessel (RPV). This monitor is designed to operate in a temperature range of-20 F to 300 F.
The monitor's temperature during plant operation was 500 F because the monitor is located directly on the steel stabilizer bracket on the East side of the RPV and it was poorly insulated. This high temperature could damage the magnetic recording tape inside the monitor. This change provided better insulation and a cooling air duct for the monitor, which reduces the operating temperature to less than 300"F. This activity slightly altered the air flow pattern inside the Drywell.
This activity was limited to surrounding the PRA with a box fabricated from insulation board and routing a 6 inch diameter air duct t o keep a constant supply of cool air to the box. The 6 inch air duct connects to the existing 14 inch ventilation riser that delivers approximately 1000 cfm to the area above the well seal plate. Approximately 100 cfm was diverted to cool the seismic monitor. The new ductwork was installed in accordance to the specification for ventilating ductwork. The chemical content of the j insulation board was evaluated based on information provided by the
- vendor. It was concluded that the halogen, sulfur, and nitrate content of the board was not high enough to cause any damage to other equipment either below or at the elevation of the box. There is not enough moisture at the elevation of the box to promote teaching of these chemicals from the 62
board. There are no requirements for the insulation to be Seismic Category 1, however the design of the box is small and rigid, and it was mounted to a Seismic Category I support.
Safety Evaluation Summary 1
The Drywell IIVAC system is not a safety system, and cannot fail in such .
a way to initiate an accident or transient previously evaluated in the SAR. l This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. This activity did not increase the probability of occurrence of a malfunction of equipment important to i safety evaluated previously in the SAR. This change did not impact the l ability of a structure, system, or component important to safety to perform its intended function. The overall affect on average Drywell temperature ,
and circulation is insignificant. The new duct is designed and installed in accordance with the design specification for ventilation ducts in the Drywell; it is seismically supported, and it does not affect the structural integrity of the main IIVAC duct. The chemical content of the insulation board is within impurity limits for the Drywell and does not pose a threat l to any stainless steel components below or at the 816' elevation which are l important to safety. The amount ofinsulation board is very small, and ifit falls, it would not pose a significant threat to safety related equipment.
Should the insulation board break and fall to the Suppression Pool, it would have a negligible impact on the debris loading bases for the ECCS suction strainers. Reducing the operating temperature of the seismic monitor improves its reliability, and helps ensure that the monitor is available to record a potential seismic event. Therefore, this change had no adverse impact on a safety function. The Drywell ventilation system does not have any design function to mitigate the consequences of an accident or transient. Therefore, the consequences of the accidents evaluated previously in the SAR were not increased and the consequences of a malfunction of equipment important to safety evaluated previously in the S AR were not increased. No plausible accidents or transients of a different type have been identified. This change had no adverse impact on existing safety equipment, and it did not create a possibility of a malfunction of equipment important to safety of a different type than that evaluated previously in the SAR. This change did not affbet the Drywell heat loading and it did not impact the average Drywell iemperature. The ability to measure and control Drywell temperature was not impacted.
This change posed no impact to any Technical Specifications, LCO's, or safe operation of the plant. Therefore, no margin of safety was reduced.
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SE 98-08 ECP 1605 - RIIR Fuel Pool Cooling Suction Piping Vent Valves Descrintion and Basis For Chance This ECP added vent valves, piping and supports to support adequate venting of the R11R Fuel Pool Cooling suction piping. The piping to which the vent line and valves were added is a line which connects the suction of the RIIR Pumps (via the common Shutdown Cooling suction line) to the Spent Fuel Pool Skimmer Surge Tanks. The RHR Fuel Pool suction piping is separated from the Spent Fuel Pool Skimmer Surge Tanks by a normally locked closed valve. The piping to which the vent line and valves are added is only used when it is desired to supplement the Spent Fuel Pool Cooling System with the RHR System although the line is connected to, and cannot be isolated from, the common Shutdown Cooling suction line. The line is connected downstream of the containment / reactor coolant pressure boundary isolation valves. The piping to which the vent line and valves are added is not considered to be a line ";hich penetrates Primary Containment although it is considered as e part of the reactor coolant pressure boundary when the RHR System is in the Shutdown Cooling Mode of operation.
Safety Evaluation Summary The probability of occurrence of an accident previously evaluated in the SAR is not increased. The loss of RIIR Shutdown Cooling Mode is not an accident previously analyzed in the SAR because the loss of Shutdown Cooling is not considered to be an accident. Reactor coolant pressure boundary items are postulated as rupturing and are evaluated in the SAR.
However, since the piping and valves installed meet all required design acceptance criteria and are of a similar size and design as other vent lines installed on the Shutdown Cooling piping, the probability of occurrence of a pipe rupture is not increased. This activity did not increase the consequences of an accident evaluated previously in the SAR. RilR Shutdown Cooling remains unafTected by this change. The RHR Shutdown Cooling Mode is not required to mitigate the consequences of an accident previously analyzed in the SAR and the affected piping is downstream of containment / reactor coolant pressure boundary isolation valves (which do limit the consequences of an accident). Thus, there is no affect on dose consequences. The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR is not increased. There is no adverse affect on Shutdown Cooling l operation. The installed piping and valves meet the design, pressure, temperature and seismic requiremet ts. Additionally, the piping and valves installed for this change were tested tuch that they were verified to ndequately perform their required func' ions. This activity did not increase 64 1
the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The possibility for an accident of a ditTerent type than any evaluated previously in the SAR was not created.
The loss of Shutdown Cocling has been evaluated per the NSOA and, since there are several other vent lines on the Shutdown Cooling suction piping, a break in the line added by this ECP was bounded by the evaluation of a break in one of the pre-existing lines. This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. There is no margin of safety defined in Technical Specifications and the NSOA that could be potentially affected by this modification and therefore the margin of safety was not reduced.
SE 98-12 Temporary Modification For Moisture Separator Reheaters Drain and Dump Valves Pneumatic Jumper Description and Basis For Chance This Temporary Modification improved the plant thermal performance by allowing the "B Side" Moisture Separator Reheater (MSR) First Stage Reheater Drain Tank level control to be accomplished by the normal drain valve, rather than the dump valve. By doing this, the water drained from the Drain Tank could be cascaded through the Feedwater lleaters, recovering up to 3 MW of energy.
This Temporary Modification disconnected the failed Drain Level Transmitter from the Drain Valve Controller, installed temporary fittings and valves on the output of the Dump Valve Level Transmitter, connected the Dump Valve Transmitter to both the Dump Valve Controller and the Drain Valve Controller, and temporarily recalibrated both Controllers to allow them to work together. The Drain Valve Controller controlled during normal circumstances, including load changes, and the Dump Valve Controller was available in case it was needed. During the installation of the Temporary Modification, there were periods when automatic control of the Drain Tank level was disabled. This Temporary Modification installation sequence instructions provided explicit instructions for manual control of the tank Drain and Dump valves in order to maintain acceptable level in the tank.
Safety Evaluation Summarv l
l This Temporary Modification did not increase the pmbability of occurrence of any accident previously evaluated in the SAR. Feedwater heating was supplied via extraction steam to the various heaters. In order to maximize heater efficiency, the drains were cascaded from heater to I 65 l
l l heater and from the MSR drain tanks to the high pressure heaters. This Temporary Modification maximized efficiency by reestablishing the cascaded drain system. Normal extraction steam flow paths do not pass through the MSR first stage reheater drain tanks, therefore there was no possibility ofinitiating a loss of feedwater heating transient. This Temporary Modification did not change the results of any accident previously evaluated in the SAR. This activity did not result in an increase in radiation dose consequences for any accident previously evaluated in the SAR. This activity did not increase the probability of i
occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The cascaded drain concept is utilized to maximize thermal efficiency. Manipulation of the pneumatic valve control signal did not have an affect on system pressure boundary. This activity did not increase the consequences of a malftmetion of equipment important to safety, and it did not create the possibility of an accident of a different type that any previously evaluated in the SAR. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. Reestablishing Drain valve control for the first stage reheater drain tank utilizing the level j transmitter for the Dump valve did not introduce the possibility of a malfunction which could have an affect on systems important to safety.
This activity did not affect any margin of safety as defined in the basis for any Technical Specifications. This modification did not reduce the margins of safety as defined in the basis for any Technical Specifications.
SE 98-15 Temporary Modification To 'B' Standby Filter Unit Exhaust Duct
, Descrintion and Basis For Chance l Replacement of the Control Building 'B' Standby Filter Unit (SFU) charcoal required partial disassembly of the SFU train and removal of the l l SFU Exhaust Isolation Valve and some ducting between the valve and the
- l. SFU Exhaust Fan. Removing the 6 inch ducting at this location, could have opened a path from the open duct through the SFU Exhaust Fan to the Control Building roof, which would have equated to a 6 inch opening l in the Control Building envelope. This opening could have challenged the l ability of the 'A' Control Building SFU unit to maintain a positive Control Building pressure. Therefore, a blank flange (duct cap) was temporarily installed to maintain the Control Building envelope. Prior to opening the l ducting, during installation of the duct cap, a glovebag was installed to l maintain the control building envelope. The glovebag was also used during removal of the duct cap.
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Safety Evaluation Summary The Control Building ventilation system is not an accident initiator.
Temporarily modifying the ducting of one SFU did not have any effect on accident initiating systems. This modification did not increase the probability of an accident evaluated in the SAR. The Control Building Ventilation is a support system. This temporary modification maintained a seal on the opening of ducting where a portion of the ductwork must be removed to access the 'B' SFU, while the 'B' SFU was in a planned LCO for charcoal replacement. During the installation of the duct cap, a glovebag was installed to maintain the Control Building envelope and therefore, operation of the ' A' SFU was not adversely affected. The 'A' SFU was fully operational and the 'B' SFU was under the restrictions of the LCO allowed by Technical Specifications. The 'A' and 'B' SFUs are redundant systems. This modification did not increase the consequences of an accident evaluated in the SAR, and it did not increase the probability of a malfunction of equipment important to safety evaluated in the SAR.
This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The consequences of a SFU malfunction was not increased. The temporary modification was removed prior to exiting the applicable LCO. The remaining ductwork is seismically acceptable. The temporary modification to the SFU ducting maintained the reliability of the system since it allowed the ' A' SFU to perforrn its function. This change did not create the possibility of an accident of a differen: type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. This change did not reduce the margin of safety as dermed in the basis for any Technical Specification. During installation and removal of the temporary modification and with the temporary modification installed, the 'A' SFU was still operable per Technical Specifications. The 'B' SFU was already inoperable and remained within the confines of the LCO until after the temporary modification was removed.
1 SE 98-16 EMA For Replaecment Of Handswitch On Remote Shutdown Panel Description and Basis For Chance This change replaced the handswitch for the 'B' RHR Heat Exchanger Inlet Valve on the Remote Shutdown Panel (RSP). Per the Residual Heat j Removal (RHR) System Operating Instruction, for 'B' side RHR, when in the Shutdown Cooling Mode, it is necessary to throttle both the 'B' RHR Heat Exchanger Inlet Valve and the 'B' RHR Heat Exchanger Bypass Valve to control cooldown rate. This change allows the Operator to 67
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throttle the 'B' RHR Heat Exchanger Inlet Valve from the RSP. The i
maintained contact switch was replaced with a spring-return-to-normal I switch. This change allows better control of the heatup/cooldown rate- ,
from the RSP when the RHR System is in the Shutdown Cooling Mode.-
Safety Evaluation Summary The probability of occurrence of an accident previously evaluated in the SAR was not increased. RHR Shutdown Cooling was enhanced by allowing the 'B' RHR Heat Exchanger inlet Valve to be throttled from the RSP. This in turn enhanced the functionality of the RSP to allow control of the plant from an alternate location. The loss of RHR Shutdown i Cooling Mode is not an accident previously analyzed in the SAR because the loss of Shutdown Cooling is not considered to be an accident. The consequences of an accident evaluated previously in the SAR was not ,
increased. There was no affect on dose consequences. The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR was not increased. RHR Shutdown Cooling is a non-safety related function. The consequences of a malfunction of equipment important to safety previously evaluated in the SAR were not increased because there was no adverse affect on Shutdown Cooling operation. The possibility for an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The new switch meets '
all required design requirements and controls the 'B' RHR Heat Exchanger Inlet Valve within its design basis. The margin of safety was not affected because the new switch meets all design requirements which provides that there is no adverse affect on Shutdown Cooling operation. 2 Based on a review of Technical Specifications / Bases and the UFSAR, no margin of safety was reduced.
SE 98-17 Temporary Shielding On A/B Recirculation Pumps Suction and Discharge Piping Descrintion and Basis For Chance Temporary lead and/or water shielding was provided around the 'A' and
'B' Recirculation suction and discharge vertical piping in the Drywell. -
This enabled work near the Main Steam Isolation Valves and Feedwater check valves and related activities during Refuel Outage 15 to reduce personnel exposure for As Low As Reasonably Achievable (ALARA) requirements. The shielding was installed and removed during RF015 cold shutdown conditions. The lead blankets were installed in the form of 68 i
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L a concentric cylinder of about 40 inches inside diameter around the 22 inch vertical piping. .
Safety Evaluation Summary The subject lead / water shielding did not create an accident because it was -
installed and removed when the plant was in Cold Shutdown conditians.
lt did not compromise the seismic integrity of any SSC even under the {
DBA safe shutdown earthquake (SSE) conditions. This activity did not increase the probability of occurrence of an accident evaluated previously
- in the SAR. Because the stresses in the structure due to the lead / water shielding were within the code allowable limits, and the shielding in no ,
way degraded any SSC, the temporary shielding did not alter any assumptions for evaluating the 10 CFR 20 and 10 CFR 100 guidelines l described in the UFSAR. This activity did not increase the consequences of an accident evaluated previously in the SAR. The lead / water shielding j did not come in contact with any equipment, piping or instrument. A mechanistic pipe break was not introduced. The structure supporting the ;
j shielding was evaluated and found to be capable of withstanding the loads l under extreme conditions based on the allowable stresses specified in the
- j. applicable codes. This activity did not increase the probability of
- occurrence of a malfunction of equipment important to safety evaluated .
previously in the SAR, and it did not increase the consequences of a l L malfunction of equipment important to safety evaluated previously in the l
SAR. The temporary shielding was in place during the period when the l' plant was already in Cold Shutdown conditions. The shielding did not impact maintaining the plant in safe shutdown condition, the integrity of the reactor coolant pressure boundary, and the capability to prevent or ,
mitigate the consequences of an accident which could result in potential offsite exposures in excess of those allowed by 10 CFR 100. This activity l did not increase or affect the acceptance criteria of 10 CFR 20 and 10 CFR i : 100. No new failure modes were created as a result of the temporary l shielding. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different ,
type than any evaluated previously in the SAR was not created. The ?
method ofinstallation of the lead / water shielding was reviewed per applicable codes and margins of safety stipulated therein, for the ability of the fastening mechanisms to prevent the shields from falling or shifting l under static or extreme dynamic conditions, and found acceptable. The l l_ method also incorporated two chains and two hooks for each l blanket / shield for redundancy. This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
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SE 98-18 ECP 1606 - Wetlift Equipment lastallation l
DescriMicn and Basis For Chance The disassembly and reassembly of the reactor during refueling outages results in radiation exposure to personnel. The equipment added by this ECP significantly reduced this exposure by allowing the radioactive reactor internals to be handled underwater, thus reducing personnel exposure. This change added three new strongbacks, a Pole liandling
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System, and new Main Steam Line Plugs. The Pole Handling System '
made a change to a safety related system, component or structure by welding the Pole Storage Rack and attaching the Pole Handling System to the Refueling Bridge superstructure. Two new strongbacks for remotely handling the Refueling Shield (Cattle Chute) and Fuel Transfer Slot Shield ;
Blocks were added to the equipment used to perfomi refueling activities j during refueling outages. The addition of the new tooling required that the !
Refueling Procedures (RFPs), used to perform the activities on the Refuel l F!oor, be changed to address the new processes. The Master Document l List was updated to include the drawings and technical information for the i new tooling. The UFSAR was updated to reflect the new handling {
process. Some drawings of the older equipment were removed from the UFSAR. Some operational procedures were changed to address the flooding and draining of the Steam Lines during the new handling process. I Safety Evaluation Summarv l
This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The addition of the Pole Handling i System was evaluated as to its affect on the seismic and structural design of the Refueling Bridge. The structural and seismic qualifications were unaffected by this change. The Pole System can be used over irradiated l
fuel; however, the total weight of the poles and the handling system is less {
than that analyzed in the Design Basis Accident for a dropped fuel l assembly and is bounded by that analysis. The poles are typically used in the reactor with the Steam Separator installed, decreasing the likelihood of fuel damage should the Pole Handling System fail. They could, however, l be used with the separator removed, but the failure of the Pole Handling System is bounded by the DBA Refueling Accident. The Pole Handling System did not increase the likelihood of the DBA Refueling Accident in that the Pole Handling System is not designed for fuel handling. The Steam Line Plugs are bounded by previous analysis done for the use of Steam Line Plugs. The new design does not increase the probability of an accident evaluated previously in the SAR. The Steam Line Plugs are similar in fit and function to previously used Main Steam Line Plugs.
They are Quality Level One safety related devices capable of withstanding
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l L Main Steam Isolation Valve Local Leak Rate Test pressures. The Dryer / Separator Strongback replaces the original Dryer Separator Sling. It also is similar in fit and function and is designed to NUREG-0554 and NUREG-0612 criteria. It is designed to work above the water or L
submerged in water. The new design does not increase the probability of occurrence of an accident evaluated previously in the SAR. The Refueling L Shield Strongback and the Fuel Transfer Canal Block lifting devices are
!' new tools. They are designed to the same criteria as the Dryer / Separator
. Strongback, thus the new design does not increase the probability of occurrence of an accident evaluated previously in the SAR. The use of the strongbacks does not change the handling processes in that slings and shackles were previcusly used to handle the loads, now remotely operated strongbacks are used. Load paths were not change, only the lifting devices have changed. The addition of the Pole Handling System to the Refueling t {
Bridge affects two design basis issues; seismic qualification of the Bridge i and the potential for a load dropped on irradiated fuel. This modification i was analyzed to show that the additional weight and configuration does . I not affect the seismic design or structural integrity of the Refueling L
Bridge. The Pole Handling System total weight is bounded by the weight i and impact assumed in the DBA Refueling Accident, therefore, this 1 activity did not increase the dose consequences of an accident evaluated previously in the SAR. The Lifting Devices and Strongbacks are bounded ;
by previous analyses in that the process or procedures do not change with 3
the use of the Strongbacks. The same load paths are employed, only the ;
l rigging was improved. The loads being carried by the new devices have not changed, therefore, this activity does not increase the consequences of i l an accident evaluated previously in the SAR. The Steam Line Plugs are safety related, Class 1' devices. They are not considered a heavy load and the dropping of a Steam Line Plug on irradiated fuel is bounded by the DBA Refueling Accident. This analysis shows that this activity did not increase the consequences of an accident evaluated previously in the SAR.
The addition of the Pole Handling System did not affect the Refueling L Bridge in such a way that the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR would increase. The structural change to the Refueling Bridge does not affect its L ability to safely handle fuel or other loads over the reactor or Spent Fuel l Pool. The Main Steam Line Plugs and the lifting devices are used in the L reactor vessel and cavity and are handled by the Reactor Building Crane l and/or the Refueling Bridge. Their use does not increase the probability of occurrence of a malfunction of equipment important to safety evaluated
- previously in the SAR. The Mair 3 team Line Plugs are designed to fit into the DAEC Steam Lines. The Steam Lines' safety function is not L affected. The lifting devices are designed to be attached to the Reactor
[ Building Crane and do not increase the probability of a malfunction of the
, crane. The Pole Handling System was analyzed for its affect on the 71 f,
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i seismic and structural design of the Refueling Bridge. Neither were affected. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated paviously in the SAR. The loads lifted by the Reactor Building Crane did not change. The rigging used to lift the loads did not increase the consequences of a malPanction of the Reactor Building Crane. The Main Steam Line Plugs have been analyzed to be used in the reactor, using the Refueling Bridge and Reactor Building Crane, and do not increase the consequences of a malftmetion of equipment important to safety evaluated previously in the SAR. The Pole Handling System is no different than any other tool used for work in or around the reactor vessel. The failure of the Pole Handling System does not create the possibility of an accident of a different type than any evaluated previously in the SAR. The DBA Refueling Accident bounds any possible accident that may involve the Pole Handling System. The use of Steam Line Plugs and lifting devices for reactor internals have been previously analyzed for their possible failure modes. The new devices did not introduce any new technique or process that would create the possibility of an accident of a different type than any evaluated previously in the SAR. The failure modes of the Refueling Bridge and Reactor Building Crane are not changed. Addition of the Pole Handling System to the Refueling Bridge did not result in a malfunction of the bridge of a different type already analyzed. Failure of the Pole Handling System or the Lifting devices did not result in or create a malfunction of the safety related equipment used on the Refuel Floor. The new devices did not introduce any new technique or process that would create the possibility of an accident of a different type than any evaluated previously in the SAR.
This activity did not reduce the margin of safety as dermed in the basis for any Technical Specification. The limits and margins of safety were not reduced. Dropping of reactor internals during refueling activities is considered incredible and the new equipment does not change that margin of safety. The Poles and the Main Ster.m Line Plugs weigh less than the analyzed fuel assembly. Their impact on fuel would not result in a reduction in the margin of safety already described in the DBA Refueling accident. The use of the new equipment does not reduce the margin of safety for the Reactor Building Crane or Refueling Bridge. The addition of the Pole Handling System has been analyzed to show that the seismic and safety classification of the Refueling Bridge is not affected by the addition of the Pole Handling System. The Reactor Building Crane remains single failure proof.
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SE 98-23 Temporary Shielding For Recirculation Pump Stub Tubes Description and Basis For Chance 1
A lead blanket was wrapped along each of the six, one inch diameter inspection stub tubes (three stub tubes per pump) on the Reactor i
Recirculation System pumps. This shielding enabled general area dose reduction during Refuel Outage 15 for ALARA requirements.
i Safety Evaluation Summary The subject lead shielding did not create an accident because it was installed and removed when the plant was in Cold Shutdown conditions.
It did not compromise the seismic integrity of any SSC even under the DBA safe shutdown earthquake (SSE) conditions. Nuclear System Leakage Control was not impacted because the reactor was not pressurized, and adequate vessel inventory injection remained available.
This activity did not increase the probability of occurrence of an accident l evaluated previously in the SAR. Because the stresses in the stub tube due to the lead shielding were found to be within the code allowable limits, and the shielding in no way degraded any SSC, the temporary shielding l did not alter any assumptions for evaluating the 10 CFR 20 and 10 CFR 100 guidelines described in the UFSAR, and there was no increase in dose consequences. This activity did not increase the consequences of an L accident evaluated previously in the SAR. The lead shielding did not degrade the equipment (stub tube) in any way. The stub tube was found to be capable of withstanding the loads under extreme conditions. Hence, a mechanistic pipe break was not introduced. This activity did not increase
! the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR and it did not increase the consequences of a malfunction of equipment important to safety evaluated
! previously in the SAR. No new failure modes were created as a result of the subject temporary shielding. Lines one inch or less in diameter are not considered as potential sources of pipe movement or jet spray for pipes inside containment. This activity did not create the possibility of an l accident of a different type than any evaluated previously in the SAR. The SSCs associated with the temporary lead shielding perform as before and this activity did not lead to any failure mode of a different type than the types evaluated in the SAR. Therefore, this activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This activity did not prevent the plant from being kept in safe shutdown conditions, and no SSC was degraded. No margins were affected by the temporary shielding.
This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
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SE 98-24 Temporary Shielding On Recirculation Pump Casing and Piping l
Description and Basis For Chance Temporary lead shielding was placed on the 'A' and 'B' Recirculation System suction and discharge piping, including the pump casing in the Drywell. The shielding was intended for ALARA purposes for working l near. motor operated valves and Drywell coolers, for carrying out in-service inspections, for general area dose reduction, and other refuel outage related activities in the Drywell. The shielding included wrapping of the Recirculation System pipes, elbows and valves by lead blankets, and placing the lead blankets on the Recirculation Pump casing and securing the blankets using chains and hooks.
Safety Evaluation Summary The subject lead shielding did not create an accident because it was installed and removed when the plant was in Cold Shutdown conditions and the stresses caused by the shielding was within the code allowable limits. It did not compromise the seismic integrity of any SSC even under the DBA safe shutdown earthquake (SSE) conditions. This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The shielding did not alter any assumptions for evaluating the 10 CFR 20 and 10 CFR 100 guidelines described in the UFSAR. This activity did not increase the consequences of an accident evaluated previously in the SAR, it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new failure modes were created as a result of the temporary shielding. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The method ofinstallation of the lead shielding was reviewed and found acceptable per applicable codes, and margins of safety were stipulated for the ability of the fastening mechanisms to prevent the shields from falling or shifting under static or extreme dynamic conditions. No margins were affected. This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
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SE 98-27 EMA To Change Low Level Radwaste Breaker Setting i
Description and Basis For Chance l
i I
Discrepancies were identified during a review of the Low Level Radwaste l (LLRW) breaker settings. Review of the upstream power supplies and breaker settings suggested that the motor control center (MCC) breaker did i not have the correct settings for protection of the downstream equipment and supply cabling. The protective settings were evaluated for all load center breakers used in the LLRW electrical distribution system. Several ;
LLRW load center breakers were identified as having incorrect setpoints or overcurrent setpoints that did not offer optimum settings for coordination of the electrical distribution system.
This activity readjusted the overcurrent settings of the LLRW load center l breaker 1BR826 to the values required within the National Electrical Code, NFPA section 70 to offer practical safeguarding of this electrical system and protection from overcurrent within the rating limitations of the l electrical components being supplied by this load center breaker. Breaker 1BR826 receives its power from LLRW bus 1BR8. Bus IBR8 supplies I several non-safety related loads in LLRW, including room and area i heaters, fims, pumps and MCCs. All loads supplied from 1BR8 are l balance of plant system and are not safety related. MCC 1BR86 is used as I a source for temporary power during refuel outages and normally is not used for plant related processes. Loss of MCC 1BR86 did not affect plant l operation or any safety related system required for operation of the plant.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. MCC 1BR86 is the only load supplied by breaker 1BR826 and is not used for any plant loads. This MCC was i originally installed to support temporary power needs during refuel 1 outages and is not used for safety related or balance of plant loads. This activity did not increase the consequences of an accident evaluated previously in the SAR, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. A failure of this breaker resulting in the loss of the LLRW load center 1BR8 would not affect any safety related equipment or system l operation. The loss of the upstream load center 1BR8 would result in a l transfer of power from an alternate source for the Technical Support Center (TSC). This activity did not change or modify the transfer scheme used to provide power to the TSC. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not create the possibility of an accident 75
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of a different type than any evaluated previously in the SAR. This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This activity did not cause or create a failure made which was not previously evaluated in the UFSAR. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification. A review of the SAR indicated there was no margin of safety involved.
SE 98-29 Residual Heat Removal and Core Spray Systems' Strainer Modification Description and Basis For Chance The potential for the Emergency Core Cooling System (ECCS) pump suction strainer clogging following a Loss of Coolant Accident (LOCA) has been the subject of several NRC bulletins / generic letters. NRC Bulletin 96-03 identifies potential resolution options that could be implemented to ensure the capability of the ECCS to perform its safety function following a LOCA. Included is the installation of a large capacity passive strainer design. This is the resolution option which was implemented at the DAEC. Passive, large-capacity strainers were installed on the Core Spray (CS) and Residual Heat Removal (RHR) Systems' suction lines. The new suction strainers are a stacked disc configuration provided by General Electric (GE). While it is apparent that the final resolution of Bulletin 96-03 requires the installation of these larger strainers, unresolved issues remain. These issues include the pending NRC SERs on the utility resolution guidance (URG) and GE strainer license topical report (LTR), as well as questions regarding containment coatings, and the use of containment overpressure in the net positive suction head (NPSH) calculation. These unresolved issues prevent the close-out of the bulletin. Therefore, the bulletin will be closed-out, and the licensing basis revised, after the open issues between the NRC, BWROG and GE are resolved.
The qualification of affected suppression pool structures and the strairer hardware was consistent with the MARK I Containment DAEC Plant Unique Analysis Report (PUAR). Additional guidance was taken from l NUREG-0661, the Load Definition Report and associated Application Guides, and the Structural Acceptance Criteria as required. Since the
, various Application Guides do not accurately present certain
! hydrodynamic properties of the proposed strainer configuration, these hydrodynamic properties are defined empirically by detailed strainer testing, as discussed in the GE LTR. In addition, guidance was taken from the appropriate Applicatien Guide for the fluid structure interaction
. techniques used in the analysis.
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l l To accomplish the required evaluations, detailed piping, strainer and Torus penetration finite element models were subjected to loads and load combinations developed using the original load generation software and methodologies as specified in the DAEC PUAR, and Application Guides.
The resulting stresses were reviewed for compliance with the DAEC PUAR and found acceptable.
I" i The new strainers (except welds) were designed using ASME Code L Section 111, Class 2-NC allowable stresses. Welds and weld acceptance criteria were in accordance with ASME Code Section III, Subsection NG. l l The structural attachment flange was designed to the rules of ASME Section Ill NF. The strainers, the associated penetration attachments, and 1 the Torus interfaces were designed to current applicable structural loads, ;
including hydrodynamic, seismic, thermal, pressure, and deadweight. l Safety Evaluation Summary I The probability of occurrence of a previously evaluated accident was l unaffected by the installation oflarger strainers. This design change meets or exceeds the design, material, and construction standards applicable to the RHR and CS system, and the suppression chamber. No i j instrumentation was affected, and the systems were not operated outside of their design or testing limits. This modification did not adversely impact .
the integrity of the Primary Containment or its ability to perform its safety
- functions. The modification was designed in accordance with, and meets I the requirements of, the applicabic ASME codes in order to ensure that the l Primary Containment remains capable of performing its function. The i consequences of a previously evaluated accident were not increased. This f modification did not change, degrade or prevent any actions described or assumed in an accident. No assumptions used in the evaluation of the L radiological consequences of any accident described in the UFSAR were
! altered. Both the RHR and CS systems are emergenev core cooling
[ systems and are required to bring the plant to a safe chutdown condition
! after an accident. The replacement of the existing strainers with larger capacity strainers did not have a negative impact on the ability of these systems to perform that function. The initiation signals, system interfaces
~
and interlocks, and pump flow rates were not changed. No fission product barrier was affected. This modification did not impact the Primary Containment System's capability to withstand the pressures and temperatures that could result from any of the postulated accidents for which it is assumed to be functional. Calculations show that stresses are l- within ASME Code allowables for the structural modification. The i probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR was not increased by this I
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(
modification. The replacement strainers meet or exceed the original design specification for materials and construction No protection features were adversely affected and no support system performance was degraded.
- There was no reduction in any system or equipment redundancy or independence. The new strainers perform significantly better than the old
~
strainers when debris is transported to the strainer surfaces. This in turn increases the safety margin associated with the overall RHR and CS system performance. All seismic and hydrodymunic load requirements are met. The containment will maintain its integrity i.nd functional capability when subjected to the loads induced by a design-basis LOCA. The ,
- consequences of a malfunction of equipment importam to safety previously evaluated in the SAR were not increased by this modification.
The only components which could have been affected by this change were the RHR and CS systems, the strainers themselves, and the Torus. This design change did not change the accident response or safety function of these components / systems. This modification did not adversely impact the integrity of the primary containment or its ability to perform its safety
.. functions, and the possibility of an accident of a different type than 4
previously evaluated was not created. This modification did not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the SAR. No new failure modes were introduced since no equipment was rehmated and the systems' function l i.
and operation were not changed. The margin of safety as defined in the !
basis for any Technical Specification was not reduced. Althcugh the ECCS and containment are referenced in the Technical Speciiications, ,
there was no change to the required action for the conditions specified and no change in the surveillance requirements. The LPCI and CS pump flow i rates and test frequency were not modified. A review of the submerged i structures calculation for Torus water level indicated a negligible effect. l The required minimum water level in the suppression pool and the condensate storage tank and the frequency of their verification remain the e same. There was no effect on the average suppression pool temperature.
The bases for the operability limits and surveillance requirements i specified were not affected by the design change. This modification resulted in a strainer that has the capability to operate acceptably with significantly increased post LOCA and operation debris loadings. This in i turn significantly increases the margin of safety for the overall system post LOCA.
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SE 98-32 Temporary Shielding On Reactor Water Cleanup and Residual lleat Removal Piping Description and Basis For Chance This change provided temporary lead shielding on Reactor Water Cleanup (RWCU) System piping and the section of the Residual Heat Removal (RHR) System piping that connects to the RWCU piping in the Drywell.
This shielding was intended for ALARA purposes for general area dose reduction for workers. This shielding included wrapping of the RWCU/RIIR pipes, elbows and the valves by lead blankets and securing these blankets in place using chains and hooks, and hanging the blankets from the pipe supports and structural steel, to shield the vertical run of the piping. The shielding was installed and removed during RFO15 Cold Shutdown conditions. The engineering evaluations for the Temporary Shielding Requests (TSRs) included limitations and recommendations.
The implementation of these limitations was ensured by the administrative controls associated with the TSRs and the Temporary Shielding Program.
The lead shielding on the horizontal sections of the piping was supported entirely by the piping itself, while the shielding on the vertical runs was supported by the pipe supports and structural steel, and the pipes were not loaded.
Safety Evaluation Summary This change did not create an accident because it was installed and removed when the plant was in cold shutdown conditions, and the stresses caused by the shielding were within the code allowable limits. The shielding did not compromise the seismic integrity of any SSC even under the DBA safe shutdown earthquake (SSE) conditions. Any snubbers that were removed for maintenance were govemed by Technical Specification provisions. Therefore, this activity did not increase the probability of occurrence of an accident evaluated previously in the SAR, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. Since the temporary shielding did not alter any assumptions for evaluating the 10 CFR 20 and 10 CFR 100 guidelines described in the UFSAR, the consequences of an accident evaluated previously in the SAR were not increased. Because the reactor was already in safe shut down condition, it was sufficient to show that the shielding did not impact maintaining the plant in safe shutdown condition, the integrity of the reactor coolant pressure boundary was not impacted, and the capability to prevent or mitigate the consequences of an accident which could result in potential offsite exposures in excess of those allowed by 10 CFR 100 was not impacted. The piping did not degrade because the stresses under the most 79
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l L severe dynamic conditions as caused by the safe shutdown earthquake L ;(SSE) were evaluated and found to be within the allowable limits. In the unlikely event of a pipe break, the break would have been enveloped by the LOCA pipe break in the Drywell, and CS and LPCI would have provided inventory makeup following this break per applicable Technical L Specifications and Outage Management Guidelines. As a result, this
. activity did not affect the acceptance criteria of 10 CFR 20 and 10 CFR i 100. Therefore, this activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new failure modes were created as a result of this shielding.
This activity did not create the possibility of an accident of a different type j than any evaluated previously in the SAR. The SSCs associated with the temporary lead shielding still perform as before, and this activity did not lead to any failure mode of a different type than the types evaluated in the SAR. Therefore, this activity did not create the possibility of a malfunction of equipment important to safety of a different type than any ,
evaluated previously in the SAR. The method ofinstallation of the lead !
shielding was reviewed for the ability of the fastening mechanisms to prevent the shields from falling or shifting under static or extreme dynamic conditions, and were found acceptable. Because the stresses in l
the SSCs that supported the lead shields were within the code allowable limits, the margin of safety was not compromised. This activity did not i reduce the margin of safety as defined in the basis for any Technical l Specifications.
SE 98-33 Temporary Shielding On 'B' and 'D' Feedwater Lines L
Description and Basis For Chance Temporary shielding was provided on the 'B' and 'D' Feedwater lines from the reactor pressure vessel to the f,rst elbow on each of the lines.
This shielding was intended for ALARA. purposes for Inservice Inspection l
(ISI) and other RF015 related activities in the Drywell. The shielding was installed and removed during RF015 Cold Shutdown conditions. The engineering evaluations of the TSRs included the limitations and recommendations. The implementation of these limitations was ensured
, by administrative controls associated with the TSRs and the Temporary K
Shielding Program. The lead shielding on the subject piping was supported entirely by the piping itself.
L Safety Evaluation Summarv l
The lead shielding did not create an accident because it was installed and removed when the plant was in Cold Shutdown conditions, and the i stresses caused by the shielding were within the code allowable limits.
f' l- 80 l ... . . .-. - . . .
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f l
l- This change did not compromise the seismic integrity of any SSC even
! under the DBA safe shutdown earthquake (SSE) conditions. Any snubbers that were removed for maintenance were governed by Technical J l Specification provisions. This shielding did not increase the probability of occurrence of an accident evaluated previously in the SAR, and it did not degrade any SSC. The shielding did not alter any assumptions for evaluating 10 CFR 20 and 10 CFR 100 guidelines described in the UFSAR. This activity did not increase the consequences of an accident evaluated previously in the SAR and the probability of occurrence of a !
- malfunction of equipment important to safety evaluated previously in the SAR was not increased. Because the reactor was already in safe shut down condition, it was sufficient to show that the shielding would not impact maintaining the plant in safe shutdown condition, the integrity of the reactor coolant pressure boundary was not impacted, and the capability i e to prevent or mitigate the consequences of an accident which could result l
in potential offsite exposures in excess of those allowed by 10 CFR 100 were not impacted. A pipe break would have been enveloped by the LOCA pipe break in the Drywell. CS and LPCI would provide inventory j makeup following this break per applicable Technical Specifications and !
! Outage Management Guidelines. As a result, this activity did not affect
- the acceptance criteria of 10 CFR 20 and 10 CFR 100. Therefore, this activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new failure L modes were created as a result of this temporary shielding. Therefore, this activity did not create the possibility of an accident of a different type than
- any evaluated previously in the SAR, and this activity did not create the j
!. possibility of a malfunction of equipment important to safety of a different '
type than any evaluated previously in the SAR. Because the stresses in the .
pipe that supports the lead shields were within the code allowable limits,
! - the margin of safety was not compromised. Also, the fastening elements j had adequate cnpacity for their ability to prevent the shields from falling or i shifting undet static or extreme dynamic conditions. Therefore, this activity did not reduce the margin of safety as defined in the basis for any l Technical Specifications.
l SE 98-34 Temporary Shielding On Core Spray Lines Descriotion and Basis For Chance Temporary lead shielding was provided at both the 'A' and 'B' Core Spray lines at the Reactor Pressure Vessel Nozzles. This was intended for
- i. ALARA purposes to support Inservice Inspection (ISI) at these nozzles and other RFO related activities in the Drywell. The shielding was installed and removed during RF015 Cold Shutdown conditions. The J.
engineering evaluations of the TSRs included limitations and i
81
recommendations. The implementation of these limitations was ensured by the administrative controls associated with the TSRs and the Temporary Shielding Program. The shielding on the piping was supported entirely by the respective RPV nozzle.
Safety Evaluation Summary i
The shielding did not create an accident because it was installed and removed when the plant was in Cold Shutdown conditions, and the stresses caused by the shielding were within the code allowable limits.
The shielding did not compromise the seismic integrity of any SSC even under the DBA safe shutdown earthquake (SSE) conditions. Any snubbers that were removed for maintenance were governed by Technical Specification provisions. This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The shielding did not alter any assumptions for evaluating the 10 CFR 20 and 10 CFR 100 guidelines described in the UFSAR. This activity did not increase the consequences of an accident evaluated previously in the SAR, and it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. Because the reactor was already in safe shut down condition, it was sufficient to show that the shielding did not impact maintaining the plant in safe shutdown condition, the integrity of the reactor coolant pressure boundary was not impacted, and the capability to prevent or mitigate the consequences of an accident which could result in potential offsite exposures in excess of those allowed by 10 CFR 100 was not impacted. A pipe break would have been enveloped by the LOCA pipe break in the Drywell, and the LPCI would have provided inventory makeup following this break per applicable Technical Specifications and Outage Management Guidelines. As a result, this activity did not affect the acceptance criteria of 10 CFR 20 and 10 1
CFR 100. Therefore, this activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new failure modes were created as a result of the shielding.
Therefore, this activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The fastening elements had adequate capacity to prevent the shields from falling or shifting under static or extreme dynamic conditions. This activity did not reduce the margin of safety as defined in the basis for any Technical ,
Specifications. l 82
SE 98-35 Temporary Shielding On Control Rod Drive Return Line Description and Basis For Chance l
Temporary lead shielding was provided at the Control Rod Drive (CRD) l return line nozzle. This was intended for general area dose reduction for ALARA purposes to support ISI at this nozzle. The shielding was !
installed and removed during RF015 Cold Shutdown conditions. The l L engineering evaluation of the TSR included limitations and l
recommendations. The implementation of these limitations was ensured by the administrative controls associated with the TSRs and the i
Temporary Shielding Program. The lead shielding on the piping was
{
supported entirely by the piping itself. '
Safety Evaluation Summary i The lead shielding did not create an accident because it was installed and removed when the plant was in Cold Shutdown conditions, and the l stresses caused by the shielding were within the code allowable limits.
The seismic integrity of any SSC even under the design basis safe shutdown earthquake (SSE) conditions was not compromised. Any snubbers removed for maintenance were govemed by Technical Specification provisions. This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The temporary shielding did not alter any assumptions for evaluating the 10 CFR 20 and 10 CFR 100 guidelines described in the UFSAR. This activity did not increase the consequences of an accident evaluated l previously in the SAR. The piping with the lead shielding was evaluated and found to be capable of withstanding the loads under extreme dynamic conditions based on the allowable stresses specified in the applicable codes, and the shielding did not affect any other equipment. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. Because the reactor was already in safe shut down conditions, it was sufficient to show that the shielding would not impact maintaining the plant in safe shutdown condition, the integrity of the reactor coolant pressure boundary l was not impacted, and the capability to prevent or mitigate the consequences of an accident which could result in potential offsite exposures in excess of those allowed by 10 CFR 100 was not impacted. A pipe break would have been enveloped by the LOCA pipe break in the Drywell, and CS and LPCI would have provided inventory makeup i
following this break per applicable Technical Specifications and Outage Management Guidelines. As a result, this activity did not affect the acceptance criteria of 10 CFR 20 and 10 CFR 100. Therefore, this activity did not increase the consequences of a malfunction of equipment 83
important to safety evaluated previously in the SAR. No new failure modes were created as a result of this temporary shielding. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. Because the stresses in the SSCs that support the lead shields were within the code allowable limits, the margin of safety was not compromised. The fastening elements had adequate capacity for their ability to prevent the shields from falling or shifting under static or extreme dynamic conditions. This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
'iE 98-36 Temporary Shielding On Reactor Water Cleanup Discharge Line Description and Basis For Chance Temporary shielding was provided on the Reactor Water Cleanup (RWCU) discharge line for ALARA purposes for general area dose reduction in the Steam Tunnel during Main Steam Isolation Valve preventative maintenance and valve maintenance. The shielding was installed and removed during RFO15 Cold Shutdown conditions. The engineering evaluations of the TSR included limitations and recommendations. The implementation of these limitations was ensured by administrative controls associated with the TSR and the Temporary Shielding Program.
Safety Evaluation Summary The lead shielding did not create an accident because it was installed and removed when the plant was in Cold Shutdown conditions, and the stresses caused by the shielding were within the code allowable limits.
Because the RWCU return system connects to the Feedwater header before entering the Drywell, failure in this line could not cause a potential of draining the vessel. Therefore, this activity did not increase the probability of occurrence of an accident evaluated previously in the SAR.
The shielding did not alter any assumptions for evaluating 10 CFR 20 and 10 CFR 100 guidelines described in the UFSAR. This activity did not '
increase the consequences of an accident evaluated previously in the SAR, i and the probability of occurrence of a malfunction of equipment important l
to safety evaluated previously in the SAR was not increased. Because the ;
reactor was already in safe shut down conditions, it was sufficient to show !
that the shielding would not impact maintaining the plant in safe shutdown f l condition, the integrity of the reactor coolant pressure boundary was not 1 impacted, and the capability to prevent or mitigate the consequences of an 1
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I accident which could result in potential offsite exposures in excess of l those allowed by 10 CFR 100 was not impacted. As a result, this activity i did not offect the acceptance criteria of 10 CFR 20 and 10 CFR 100.
l Therefore, this activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new failure modes were created as a result of this temporary shielding.
This activity did not create the possibility of an accident of a different type l
than any evaluated previoasly in the SAR. No failure modes that could adversely impact the performance of any SSC could be identified by this activity. The SSCs associated with the temporary lead shielding perform as before and this activity did not lead to any failure mode of a different type than the types evaluated in the SAR. Therefore, tMs activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The margin of safety was not compromised. The fastening elcments had adequate capacity for their ability to prevent the shields from falling or shifting under static or extreme dynamic conditions. This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
SE 98-37 Temporary Modification To Cross-tie Service Air To Containment Atmosphere Control Description and Basis For Chance The purpose of this Temporary Modification was to provide a source of Service Air to the Drywell during Refuel Outage (RFO) 15. The Service Air was needed to operate pneumatic hand tools in the Drywell during the outage. This modification involved connecting a hose from a Service Air Valve to a Containment Atmosphere Control System valve. A second hose was then placed on another Containment Atmosphere Control System valve and routed to the interior of the Drywell where it was then used to operate pneumatic hand tools. The routing of the air hose into the Drywell was performed after the plant was in Cold Shutdown in accordance with the Temporary Modificanon. The hose which was introduced into the l Drywell was removed prior to the restoration of Primary Containment.
Safety Evaluation Summary The probability of occurrence of an accident evaluated previously in the SAR was not increased. The installation of a mechanicaljumper between the Service Air System and the Containment Atmosphere Control System was not an accident initiator, nor was routing an air line from the l Containment Atmosphere Control System to the interior of the Drywell while the plant was in Cold Shutdown. Furthermore, the UFSAR states l 85 l _ _ -
L that the total failure of the Compressed Air System will not affect the l operation of safety. equipment. Neither Service Air nor the Containment ;
L Atmosphere Control System are required to mitigate an accident. The installation of this Temporary Modification did not affect the ability of the ,
Containment Atmosphere Control System or the Service Air System to j perform their function. This temporary installation did not impair any other systems from performing their function Therefore, this activity did i not increase the consequence of an accident as evaluated in the SAR, and l this activity did not adversely affect the equipment in any systems :
L . including the systems related to safety. The probability of occurrence of a malfunction of equipment important to safety evaluted previously in the 3
SAR was not increased. This activity did not increase the consequences of '
L a malfunction of equipment important to safety evaluated previously in the l
SAR. The consequences of an equipment malfunction were not more l l
severe as a result of the installation of this Temporary Modification because the mechanicaljumper did not affect equipment in any systems related to safety. The Containment Atmosphere Control System piping utilized is normally isolated from the rest of the Containment Atmosphere y Control System. While the Temporary Modification was in place, the l normal isolation valve remained closed and a blind flange remained
! inserted. Therefore, this Temporary Modification did not change the way l the Containment Atmosphere Control System was operated. This activity L
did not create the possibility of accident of a difTerent type than any previously evaluated in the SAR, and the possibility of a malfunction of
- equipment important to safety of a different type than any previously '
)
evaluated in the SAR was not created. The installation of this Temporary Modification had no impact on the margin of safety as defined in the basis for any Technical Specification.
SE 98-41 Core Reload and Cycle 16 Design Descnotion and Basis For Chance
! The purpose of the core modification package was to develop a core loading pattern which would ensure that Cycle 16 operation could be performed in accordance with the DAEC technical and licensing basis.
The package encompassed Cycle 16 design activities and Reload 15 shuffle activities. A shuffle consisted of moving previously loaded fuel
[ and adding new fuel bundles to the core consistent with the Cycle 16 l reload pattern. Permanently discharged bundles were placed in the spent l fuel pool. The specific loading pattem, Cycle 16 operating strategy, and l transient analysis were performed under the General Electric 10 CFR 50
[ Appendix B Quality Assurance Program. This program was evaluated by
! DAEC Quality Assurance during a series of Source Surveillances and the program was found to meet the aspects of acceptable design control as t
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I i defined by 10 CFR 50 Appendix B The fuel used for Cycle 16 is the same mechanical design (GE10) as that loaded and operated in previous cycles at the DAEC. This fuel is designed and licensed by the NRC via GE's topical report GESTAR-II and GE Fuel Bundle Designs. The Cycle 16 bundles were changed with increases in U235 enrichment and burnable poison loading as described in the Fuel Bundle Design Reports. This increase is bounded by previous analysis for the Duane Arnold Energy Center. Cycle specific analyses were performed to include allowances, via Minimum Critical Power Ratio (MCPR) limits, ibr . specific pieces of plant I equipment out of service, such as Turbine Bypass Valves out of service I and/or Recirculation Pump Trip out of service.
Safety Evaluation Summary l The probability of an accident evaluated previously in the SAR was not increased by the shuffle of fuel assemblies or operation of this fuel during !
Cycle 16. The fuel loaded in Reload 15 and operated during Cycle 16 is I the same fuel (GE10) as that loaded / operated in previous reloads / cycles at l the DAEC. No changes in fuel handling practices or equipment that I would affect the bundle drop accident were made with this core l modification. Additionally, the Nuclear Fuel System as described in the NSOA does not perform any safety action for transients, accidents, or special events. The consequences of an accident previously evaluated in the SAR were not increased by the shuffle of fuel assemblies or operation of this fuel during Cycle 16. The core loading pattern was evaluated in the Supplemental Reload Licensing Report (SRLR) and Nuclear Design Report (NDR) to demonstrate compliance with the licensing basis as described in the SAR. The GE-10 fuel design meets all requirements for fuel and is direct replacement (for mechanical design) of the fuel l previously in use. Although the U235 enrichment was increased in the bundle design for Cycle 16, that enrichment is bounded by the current analysis and does not increase the consequences of an accident. That is, offsite dose does not increase as it is bounded by the current source term used in analysis. The probability of the occurrence of a malfunction of equipment important to safety as evaluated previously in the SAR was not increased. No change was made to equipment important to safety (i.e., the fuel cladding has not changed nor has mechanical bundle design) with this core modification package. Likewise, no change was made to the interaction with equipment important to safety. It was demonstrated that the GE-10 fuel loaded in this reload met all acceptance criteria for fuel designs and was manufactured / constructed under an NRC-approved l Quality Assurance Program. The probability of a failure of the fuel cladding when operated in accordance with the fuel thermal limits is not l-increased from that previously evaluated. Also, the ASME Vessel overpressure analysis demonstrates that the peak RPV pressure is well 87
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I within the design allowable limit. Therefore, the probability of a vessel overpressure and subsequent overstressing of the reactor coolant pressure l boundary was not increased from that previously evaluated. This core modification did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The j modification of the loading pattern was evaluated to ensure that the fuel would perform its intended function during postulated malfunction of equipment. The results of the transient and accident analysis presented in the SRLR demonstrate that the fuel cladding integrity is maintained when the thermal limits, including the analyzed equipment out of service <
penalties and operation outside the " Exclusion Zone" are met. This core l modification did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The possibility of a l malfunction of equipment important to safety of a different type than previously evaluated was not introduced during this reload. This reload did not reduce the margin of safety as defined in the basis of any Technical l
Specification. During Cycle 16 operation, as summarized in the SRLR, the margin to safety is maintained for all Technical Specification parameters as long as the fuel thermal limits are met. The limits described in 10 CFR 50.46 are maintained for Cycle 16 operation. Similarly, Technical Specifications, as appropriate, were met for activities associated with reload activities.
SE 98-42 Temporary Modification For Maintenance On RHR Loop 'A' Torus Suction Isolation Valve Description and Basis For Chance This Temporary Modification involved installing a blank flange to prevent Torus water from reaching the RHR Loop 'A' Torus Suction Isolation Valve during repairs on the valve. This valve is the common Torus suction valve for the 'A' RHR loop. The blank fiange was installed on the flange located inside the Torus on the RHR suction piping, after the suction strainer was removed. This modification was installed only when the plant was in Cold Shutdown, within Technical Specification guidance for the 'A' RHR loop being out of service and in accordance with applicable Outage Management Guidelines.
Safety Evaluation Summary The probability of occurrence of an accident previously evaluated in the i SAR was not increased because there are no accidents previously analyzed in the SAR for the plant conditions and for the affected components i involved with installation of the blank flange. Sufficient capability to mitigate the consequences of an accident previously analyzed in the SAR 88
._ _ . _ _ __ ._ - . __ _ . ~ . . .. .
for the specific plant conditions were maintained via Technical Specifications and Outage Management Guidelines. The remaining systems were capable of mitigating the consequences of an accident and there was no affect on dose consequences. The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR was not increased, and the consequences of a malfunction of equipment important to safety previously evaluated in the SAR were not increased because the entire 'A' loop of R"R was removed from service within the restrictions contained in Technical Specifications and Outage Management Guidelines. There was no other equipment which was affected by this blank flange. The possibility for an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a difTerent type than any evaluated previously in the SAR was not created. Based on a review of Technical Specifications and bases, the UFSAR and the NSOA, there was no margin of safety defined in these documents that was affected by this blank flange installation during the specified times and plant conditions. Therefore the margin of safety was not reduced.
SE 98-43 Temporary Modification For Monitoring Residual Heat Removal Service Water (RHRSW)/ Emergency Service Water (ESW) Pit Level Description and Basis For Chance To enhance the Operator's ability to maintain RHRSW/ESW Pit level, a temporary, non-safety related indicator was provided so that the Operator could monitor RHRSW/ESW Pit level from the front panel area of the Main Control Room. This indicator was connected to the ' A' and 'B' RHRSW/ESW Pit Level Recorders' instrumentation loop at a convenient location between the Power Supplies and the Level Recorders. The indicator was mounted in the front panel area of the Control Room and was physically restrained such that it would not damage any other equipment.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. RHRSW/ESW Pit level indication is not an accident initiator for accidents previously evaluated in the SAR. This activity did not increase the consequences of an accident evaluated previously in the SAR, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. RHRSW/ESW Pit level indication is non-safety related and is not used to mitigate the consequences of an accident previously evaluated 89
in the SAR. There was no impact on equipment important to safety. Dose consequences were not increased. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and the consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased. The possibility of an accident of a different type than any evaluated previously in the SAR was not created. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. Per review of the UFSAR and Technical Specifications, there are no margins of safety in the basis for any Technical Specification concerning RHRSW/ESW Pit level indication. The margin of safety as defined in the basis for any Technical Specification was not reduced.
SE 98-44 Temporary Modification For Maintenance On Control Building Load Center Description and Basis For Chance This Temporary Modification inserted a gagging device on a non-safety related Service Air Supply Header Isolation Control Valve to maintain the valve in the open position in support of maintenance activities performed on the Control Building Load Center. The valve needed to remain open to provide compressed air to various plant equipment, including the Main l Steam Line Plugs which were installed for refueling activities. When the Control Building Load Center is deenergized, the solenoid valve to the Service Air Supply Header Isolation Control Valve deenergizes causing the Control Valve to revert to its fail closed position. This Temporary '
Modification inserted a gagging device to prevent the Control Valve from closing during the period of time when the solenoid valve was deenergized. A back-up power source was made available to supply the solenoid valve with electrical power approximately one hour after the normal electrical source was removed from service for maintenance activities. The gagging device was used only during the period when no power existed to the solenoid valve.
Safety Evaluation Summary The insertion of a gagging device on the Service Air Supply Header did not change the way the Service Air System is normally operated and it did not change the probability that a loss of plant air would occur. The Service Air Header Supply Isolation Control Valvc was installed to isolate in the unlikely event that the system header pressure drops below a pre-determined value. If, however, a break were to occur in the service air header downstream of the Service Air Header Supply Isolation Control 90
Valve, and the isolation of the gagged valve was not possible, the safety related components in the plant that must have air to them would still be operable due to the reliable source of back-up compressed air. It was also possible to manually isolate the Service Air System to mitigate a pipe failure in the S;rvice Air System. Therefore, this activity did not increase
- the probability of occurrence of an accident evaluated previously in the SAR, and it did not increase the consequences of an accident evaluated previously in the SAR. Even if the instrument and service air header pressure were to diminish as the result of a pipe failure, the equipment needed to mitigate this accident would still be available due to the reliable back-up Compressed Air System. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. The Service and Instrument Air Systems continue to operate in a normal manner, which is to provide the plant equipment with clean compressed air. Therefore, this activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This activity did not create the possibility of an accident of a different type than previously evaluated, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The Instrument and Service Air Systems are not discussed in the Technical Specifications.
The insertion of the gagging device would not reduce the margin of safety as defined in the Technical Specification ifit were installed while the plant was on-line. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 98-45 EMAs To Change Post Accident Sampling System (PASS) IIcat Tracing Temperature Controllers Descrintion and Basis For Chance EMAs were completed to replace the 5 existing temperature controllers on the heat tracing on the two PASS Drywell sample lines and the two PASS Torus sample lines. The UFSAR was also updated. Redundant, non-Class IE heat tracing has been provided which is sized to hold a line temperature of approximately 200-215 F and is capable of continuous operation. All heat-traced sample lines are thermally insulated. Previously the heat tracing was to hold line temperature at 250 F. This temperature was based on limiting plateout of gaseous iodine in the PASS sample lines. Due to the Noble Metals Chemical Addition (NMCA), heat tracing to this high temperature can cause undesirable effects on the accuracy of PASS hydrogen and oxygen measurements.
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j- Safety Evaluation Summary The PASS is not an accident initiator. It provides sampling capabilities post-accident. The change to heat trace (IIT) temperature is necessary to ensure more accurate hydrogen and oxygen samples. Therefore, changing the heat tracing from 250 F to 200-215 F did not increase the probability of occurrence of an accident evaluated previously in the SAR. This activity did not increase the consequences of an accident evaluated previously in the SAR. The reduction in heat trace temperature is necessary to improve the accuracy of the hydrogen and oxygen samples taken by the PASS. Changing the heat tracing from 250*F to 200-215 F did not increase the consequences of a previously evaluated accident; rather this change provides a more representative sample to the operator l which will aid the operator in taking appropriate actions in the event of an accident. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased.
The PASS is not a safety-related system. Changing the heat tracing from
{
250*F to 200-215 F did not increase the consequences of a malfunction of l equipment important to safety evaluated previously in the SARiand the possibility of an accident of a different type than any evaluated previously L in the SAR was not created. This activity did not create the possibility of L a malfunction of equipment importat to safety of a different type than any l evaluated previously in the SAR, and the margin of safety as defined in l
the basis for any Technical Specification was not reduced.
- SE 98-47 EMA To Relace Drywell Cooling Loop 'A' Well Water Return
- Header Drain Valve Description and Basis For Chance The Drywell Cooling Loop 'A' Well Water Return Header Drain Valve was replaced to comply with the HLE-34 line classification and Quality Level I valve requirements. The valve was removed and replaced with an l ASME Ill EBE-GT. The previous valve was a 1 inch EBD-GT. The L replacement valve is a 3/4 inch EBE-GT. The piping configuration consists of a 1 inch by 3/4 inch reducing coupling, a 1 inch by 3/4 inch reducing insert and a 3/4 inch EBE-GT valve. The section of pipe downstream of the valve is capped to provide redundancy and is classified t as JBD. This modification complies with applicable Design Codes. The piping configuration remained unchanged other than the reduction in size.
p There were no pipe supports affected with the valve replacement, therefore j the system's seismic class was unchanged.
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Safety Evaluation Summarv This change did not increase the probability of occurrence of an accident l evaluated previously in the SAR. This activity met the design, material, and construction standards applicable to the component that was changed.
The change in valve size did not affect the ability of the valve to establish containment isolation or pressure boundary integrity. This activity did not increase the consequences of an accident as previously evaluated in the SAR. The dose consequences remain unchanged as a result of the valve L replacement. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in
- ' the SAR. This activity met the original design specifications for material and construction practices and did not degrade the components reliability.
The new valve and the old valve are both manually operated and their normal valve position is closed. This change did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The consequences of malfunction are unchanged with the change in valve size. The configuration of the new valve has no impact on the seismic qualification of the piping system. The possibility of an accident of a different type than any evaluated previously in the SAR was not created. Both the old and new valve are rated for the same operation in pressure and temperature. This change did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The change in valve size did not affect the margin of safety as the safety function can be performed by either size valve. This change in valve size had no effect on the margin of safety.
SE 98-49 EMA To Replace Turbine Stop Valve SV-3 Before Seat Drain Valve Description and Basis For Chance The Turbine Stop Valve SV-3 Before Seat Drain Motor Operated Valve (MOV) had previously been repaired due to a worn seat surface. The seating surface was again damaged. It was recommended the valve be changed to avoid recurrence of the same scenario and to ensure continued I plant operation. An exact replacement for the valve was no longer available. The existing globe valve was removed and replaced with a gate valve. The motor operator was not changed. i Safety Evalustion Summarv
,. Using a gate valve meets the functional requirements for open and shut operation. The new valve meets design requirements. Therefore this modification did not increase the probability of occurrence of an accident 93
L
, evaluated previously in the SAR. The alTected valve is the isolation L boundary between the stop valve and the condenser. It is opened by l operator action to allow the stop valve seat to drain to the Low Pressure Condenser. The gate valve provides the appropriate function of the MOV which is either Open or Close.- The MOV is not required for the throttling ;
function which is usually associated with globe valves. This modification l did not cause the system to experience radiological releases in excess of limits set by 10 CFR 100 and 10 CFR 50 Appendix A guidelines. Since the MOV is not used for providing reactor coolant, primary, or secondary containment pressure boundaries, changing the valve type from globe to gate did not increase the consequences of an accident evaluated previously in the SAR, and the probability of occurrence of a malfunction of equipment important to safety evaluated in the SAR was not increased.
This valve is classified as safety significant, not safety related. The MOV provides system flow control by operator action of placing it in the open or close position. Gate valves are designed for this type ofoperation. Globe valves are usually used for the throttling functions, which is not a !
I requirement of this MOV. The chances of failure of the new valve in the I close position are no greater than the old valve. This modification did not L increase the consequences of a malfunction of equipment important to l- safety evaluated previously in the SAR, and the possibility of an accident l of a different type than any evaluated previously in the SAR was not l- created. The failure modes of the gate valve are the same as those of the globe valve. The new valve was installed at the same location and uses l the same operator as the previously existing equipment. This activity did not create the possibility of a malfunction of equipment important to safety i of a different type than any evaluated previously in the SAR. The margin of safety was not reduced by changing the valve type from a globe to a i gate.
SE 98-51 EMA To Add Drywell Ventilation Ductwork Description and Basis For Chance l
! The Peak Recording Accelerograph (PRA) is a monitor that measures the magnitude of a seismic event. The monitor on the 'B' Main Steamline pipe support was designed to operate in a temperature range of-20 F to 300 F. The monitor showed signs ofinternal damage resulting from operating at high temperatures. The monitor was mounted directly to a steel pipe support bracket in the Drywell. The Steamline was well insulated and a temperature indicator strip which was on the monitor indicated the temperature did not exceed the operating limits. Inspection of the removed monitor showed damage resulting from high temperature on the steamline side. The purpose of this activity was to slightly alter the
- air flow pattern inside the Drywell to add a duct outlet to direct cooling air 94
l i ;
l over this monitor. This activity added a 1 foot long,4 inch diameter, air duct to keep a constant supply of cool air to the seismic monitor. The 4 inch air duct connects to the 14 inch ventilation riser that delivers
' approximately 1000 CFM to the area above the well seal plate.
Approximately 30 CFM is diverted to cool the seismic monitor. The new ductwork was installed in accordance to the specification for ventilating ,
ductwork. The Dry,vell Ventilation P&lD was also revised to indicate this '
change. ;
l Safety Evaluation Summarv
]
This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The Drywell HVAC system is not a safety system, and cannot fail in such a way as to initiate an accident or '
transient previously evaluated in the SAR. The Drywell ventilation
! system does not have any design function to mitigate the consequences of an accident or transient. Therefore, the consequences of the accidents evaluated previously in the SAR were not increased. This change to the i Drywell ventilation duct did not impact the ability of a structure, system, l
or component important to safety to perform its intended function. The l
overall affect on average Drywell temperature and circulation is insignificant. The new duct was designed and installed in accordance with the design specification for ventilation ducts in the Drywell; it is seismically supported, and it does not affect the structural integrity of the L main HVAC duct. The lower operating temperature of the seismic l' monitor improves its reliability, and helps ensure that the monitor is available to record a potential seismic event. Therefore, this change had no adverse impact on a safety function, and it did not increase the consequences of a malfunction of equipment important to safety evaluated
.)
previously in the SAR. The Drywell HVAC system is not a safety system, j and cammt fail in such a way to initiate an accident or transient. The new l
Juw wod reets existing specifications for installation in the Drywell. No ;
plausible acciais or transients of a different type were identified. This change had no adv rs - impact on existing safety equipment, and it did not l create a possibility f a malfunction of equipment important to safety of a
- different type than that evaluated previously in the SAR. This change did L
not affect the Drywell heat loading and the average Drywell temperature L was not affected. The ability to measure and control Drywell temperature
- l. was not impacted. This change posed no impact to any Technical Specifications, or safe operation of the plant. Therefore, no margin of safety was reduced.
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SE 98-52 ECP 1608 - Replacement Of Drywell and Torus Sampic Lines IIcat l Tracing l
l Description and Basis For Chance
! This modification replaced heat tracing (HT) on sample lines to Drywell and Torus hydrogen and oxygen monitors with new heat tracing and l
insulation to raise the sample line temperature to 200-215 F. This change was needed to improve the accuracy of the hydrogen and oxygen indications. Inaccuracies could be introduced at temperatures below 200 F, due to volumetric contraction, and at temperatures above 215 F l due to recombination of hydrogen and oxygen caused by the presence of noble metals. The controls for the new heat trace was provided in new panels. This modification also removed existing heat tracing from sample
! lines to the radiation monitors. This was needed to reduce the sample temperature to an acceptable value for these monitors, since the radiation monitors and the hydrogen and oxygen monitors share a portion of l common (higher temperature) sample piping. A portion of the PASS sample line was also heat uaced to 200-215 F (see SE 98-45).
To reduce the amount of piping requiring HT and insulation (and in turn reduce radiation exposure during installation), the ' A' Torus sample piping was rerouted by cutting the 'A' side sample piping and reconnecting it to the 'A' side return piping. The 'B' I l Torus sample piping was rerouted from the suppression chamber i nozzle by cutting the 'B' side sample piping and reconnecting it to l the 'B' side return piping. Additional pipe supports were installed on both the 'A' and 'B' side sample piping from both the Torus and the Drywell. During refueling outage (RFO) 14, Noble Metal Chemical Addition (NMCA) was implemented at the Duane L Arnold Energy Center (DAEC). After a loss of coolant accident I (LOCA), these noble metals could enter the containment and be deposited in the Containment Atmosphere Monitoring (CAM) sample piping. This could also have an impact on the accuracy of
! hydrogen and oxygen measurements. Line temperature must be l kept below 215 F due to the " noble metal effect," but yet high enough to ensure a dry sample.
Safety Evaluation Summary The CAM system is not an accident initiator. It provides indication of hydrogen, oxygen and radionuclides in containment during normal operation and post-accident. Since the radiation monitors and the
, hydrogen and oxygen monitors share common sample lines, removal of the llT on the radiation monitor sample lines was necessary to allow the 96
l sample to cool down from the new higher value prior to reaching the radiation monitors. This modification ensures proper operation of the CAM system ensuring the operators have accurate indications of conditions in containment afler a loss of coolant accident (LOCA).
Therefore, this modification did not increase the probability of occurrence of an accident evaluated previously in the SAR. There is no increase in the radiological consequences of any previously analyzed UFSAR accident, and the changes made by this activity did not change, degrade or prevent actions described or assumed in an accident discussed in the UFSAR. The Drywell temperature is below 200 F within 100 seconds and remains below 200 F for the duration of the accident. The Suppression Pool temperature peaks at approximately 200 F at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> post-accident. Therefore heat tracing and insulating to 200-215 F results in an accurate reading except for the first 100 seconds. It is recognized that a LOCA creates a high Drywell pressure condition and isolates the analyzers. Therefore, a delay will exist between the accident and the beginning of monitoring. While the UFSAR states that the hydrogen and oxygen monitors provide accurate hydrogen and oxygen indications over a l
range of 100-300 F, the monitors are not required by NUREG 0737 until 30 minutes after safety injection. IIeat tracing to 200-215 F will not provide an accurate indication at these high containment temperatures, but after 30 minutes have passed and the containment temperature has decreased, the indications will be within the accuracy required by RG 1.97. Other events are postulated that result in temperatures in excess of 215*F. However, these events do not result in the significant generation of hydrogen or oxygen. The temperatures for these events may result in inaccuracies, but the overall system accuracy (both instrumentation 3 accuracy and the supply piping accuracy) is expected to remain within established allowances. This modification did not increase the consequences of a previously evaluated accident. The only safety related system affected by this modification is the CAM system. The design, fabrication, and testing of this modification was performed in accordance with applicable codes and standards. The components used meet the system design r quirements and were installed and tested using approved plant proceLres. The modification was designed and installed in accordance with applicable codes. The modification has been seismically evaluated to ensure the piping is able to support the additional incremental loading of the modification. The effect of the higher temperatures on the solenoid valves in the sample lines was evaluated and the results found acceptable. Therefore, this modification did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This modification did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and the possibility of an accident of a different t 97
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l type than any evaluated previously in the SAR was not created. This l modification did not create the possibility of a malfunction of equipment !
! important to safety of a different type than any evaluated previously in the l t
SAR, and the margin of safety as defined in the basis for any Technical !
Specification, was not reduced, l SE 98-53 EMA To Replace lsolation Valves To Reactor Feed Pump Cooler l
Description and Basis For Chance l
l Two isolation valves to the Reactor Feed Pump Motor Cooler have had l problems closing completely as the result of silt and rust buildup on the internals from the General Service Water (GSW) System. The valves (a gate valve on the supply side to the Feedwater Pump, and a globe valve on !
the return side of the Feedwater Pump motor cooler) were having !
difficulty closing completely in the past years. A butterfly style valve replaced the existing valves in an attempt to improve the intended function ofisolation in the associated system. This change did not affect any seismic piping or supporting structures, or have a major affect on the flow characteristics (flow rate or pressure drop) to the Feedwater Pump motor cooler.
Safety Evaluation Summary l This activity did not increase the probability of occurrence of an accident I evaluated previously in the SAR. This activity involved the installation of new butterfly valves in the GSW System. The GSW System is not l contained in the NSOA and is not a direct contributor to any accidents in l the SAR. This activity met the design, material, and construction standards applicable to the component that were to be changed. Thus, I l there was not a functional impact on the Feedwater Pump, so the l probability of a Feedwater pump trip was not increased, and the consequences of an accident as previously evaluated in the SAR was not increased. This activity did not increase the probability of occurrence of a l malfunction of equipment important to safety evaluated previously in the SAR. This activity met the original design specifications for material and i construction practices and did not degrade the components reliability or performance requirements. The configuration remains unchanged other than a difTerent style of valve used in the system. There are no seismic l
requirements for the lines for which these valves are involved and there is no change in the seismic qualities of the piping system. This activity did not increase the consequences of a malfunction of equipment important to
- safety evaluated previously in the SAR. This change did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The possibility of a malfunction of equipment important to l
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safety of a different type than any evaluamd previously in the SAR was not created. GSW is not discussed within Technical Specifications. GSW is primarily a support system. The installation of new isolation valves for GSW to the Feedwater Pump motor did not reduce the margin of safety, as the probability and consequences of a Feedwater Pump trip are unchanged.
SE 98-63 Well Water Flow Balancing Description and Basis For Chance The well water flow requirement for the main plant intake coils has been changed from 850 gpm to 480 gpm. This change in flow required isolation of the Control building Chiller bypass throttle valve. This required an evaluation to update the UFSAR, Well Water System P&lD, and Operating Instruction. A study performed during the summer of 1996, indicated that 480 gpm (instead of 850 gpm) was adequate for main plant cooling. The original design well water requirements for plant HVAC equipment was 1376 gpm. The calculation assumed 840 gpm for the main plant intake coils and 180 gpm for the Drywell Cooling System. The UFSAR listed the flow requirement for intake coils as 850 gpm and Drywell cooling at 448 gpm. This represented a significant increase in the demand on the Well Water System (approximately 268 gpm). Increasing well water flow to account for the increase in demand was not practical due to higher costs involved. The Drywell Cooling System was originally l intended to have two 100% capacity loops, each requiring 180 gpm. To I limit the Drywell temperatures, it was necessary to run both loops.
l Additional coolers in the upper elevations in the Drywell have increased !
well water demand to 448 gpm. The coolers were sized assuming a 135 F inlet air temperature. This is now the DAEC Technical Specification maximum temperature. In order to accommodate available well water to various plant needs, in the summer of 1996, the well water flow to the main plaat intake coils was adjusted to 480 gpm to ensure that the Drywell cooling system received its design flow. A Temporary Modification was established to monitor temperature at various locations throughout the summer. No significant consequences of operating the intake coils at reduced flow were observed. Reducing the well water flow to one-half of the design flow prom 850 gpm approximately to 420 gpm) results in approximately a 10 F increase in the air temperature of the coil. The reduction in well water flow results in a general area Reactor Building and Turbine Building temperature of approximately 100 F on a hot summer day. These values are very conservative and are within the nominal design temperature for vanous locations in the two buildings.
99
i Safety Evaluation Summary i
l This activity did not increase the probability of occurrence of an accident l
l evaluated previously in the SAR. A reduction in well water flow to the !
Main Plant Intake Coils could result in higher Reactor Building temperature. The general access areas of the Reactor and Turbine I Building are designed to be nominally at 90* F on a typical hot summer day. The change in flow through the main plant coils could result in an overall rise in the general access areas by as much as 10 F. The potential l increase to 100 F is not a concern. Equipment performance is not i adversely impacted by the increase in temperature. All general access areas are expected to remain safe for occupancy and are not expected to
, raise above 104 F (maximum normal temperature). There are no accidents related to well water which have been evaluated in the SAR.
The Well Water System is not safety related. This activity did not {
increase the consequences of an accident evaluated previously in the SAR. l There was no increase in the radiological consequences of any previously l analyzed UFSAR accident, because this activity did not alter any I assumptions previously made in evaluating the radiological consequences of an accident, and it does not play a role in mitigating the radiological ,
consequences of an accident described in the UFSAR. All containment '
isolation functions for well water remain unaffected. This activity did not I increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The change in valve i line-up did not increase the possibility of a transient. The potential increase of higher Reactor Building air temperature was fully evaluated and is within parameters of nominal design temperatures for various locations in the Reactor Building and Turbine Buildings. The Well Water System and/or main plant intake coils are not safety related and the ,
changes did not effect any equipment which has a function important to safety. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. All containment isolation functions for well water remain unaffected. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction
, of equipment important to safety of a different type tha any evaluated l previously in the SAR was not created. The margin of saf ety as defined in the basis for any Technical Specification was not reduced because there are no bases for any Technical Specifications associated with the Well Water System.
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SE 98-76 Temporary Modification For Inducing Differential Pressure Across Low Pressure Coolant Injection inject Check Valves Description and Basis For Chance This temporary modificatien removed a pipe cap and installed a temporary pressure monitoring / venting rig to piping downstream of permanent -
isolation valves, RIIR Loop 'B' LPCI Inboard Injection Isolation Valve Leakage Test Connection Outboard Isolation and RHR Loop 'A' LPCI Inboard Injection Isolation Valve Leakage Test Connection Outboard Isolation. The Leakage Test Connection Outboard Isolations are
, considered to be part of the Primary Containment and reactor coolant 4
pressure boundary. The rig was physically connected outside of the Primary Containment pressure boundary and reactor coolant pressure boundary. This rig consisted of a small section of 1/2 inch tubing / piping, a pressure indicator, one valve downstream of the pressure indicator and piping / tubing routed to a floor drain.
l 5 The Leakage Test Connection Isolation Valves (and their inboard isolation l valves) were opened on an intermittent basis to pressurize the rig i
(including the temporary pressure indicator) up to the temporarily-installed valve. The temporarily-installed valve was opened and closed to l vent the LPCI Inject Line Piping between the LPCI Inject Check Valves
- and the LPCI Inboard Inject Valves to various pressures (not below 900 l l
psic;) to create a differential pressure across the LPCI Inject Check Valves.
4 The permanent isolation valves were then closed. This provided a means to determine the pressure that would stop the check valves from chattering and the period of time for which the chattering would stop. These pressures and times could then be implemented into routine plant ,
operation until the next refueling outage to prevent any long-term damage !
to the LPCI Inject Check Valves due to chattering.
The various modes of the RHR System remained operable because the amount of flow able to pass out of the subject tubing / valve / pressure indicator arrangement was insignificant when compared to the amount of 1-flow which the IUiR System is capable of delivering to the reactor vessel.
Additionally, venting the pressure of the subject piping to not-less-than 900 psig ensured that the subject piping would not flash to steam and remained full of water.
Safety Evaluation Summary The probability of occurrence of an accident previously evaluated in the SAR was not increased, and the consequences of an accident previously evaluated in the SAR was not increased because the temporary rig was 101
adequately qualified for pressure rating, and opening the permanently-installed isolation valves to measure pressure and vent the subject piping was under the direct control of an Operator (which complies with Technical Specification allowable actions for inoperable containment isolation valves) to provide quick re-isolation if needed. The capability of the RIIR System to maintain its safety functions was not adversely affected. There was no affect on the fission product barriers or dose consequences. The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR was not increased. The rig did not adversely affect the LPCI Inject piping's seismic criteria, separation criteria and environmental qualifications. No significant additional loads which adversely affect those previously analyzed in the original design were added. No equipment protection features were deleted or modified. Support system performance necessary for reliable operation of the Primary Containment and reactor coolant pressure boundaries were not affected. System / equipment redundancy or independence was not affected since the rigs are in opposite divisions.
The rig did not affect any other equipment important to safety (particularly the containment isolation valves which continued to provide isolation capability). The consequences of a malfunction of equipment important to safety previously evaluated in the SAR was not increased. The two permanent isolation valves (inboard and outboard), while closed, provided the required isolation. When the two permanent isolation valves (inboard and outboard) were open, while venting, the LPCI Inject Check Valves provided the isolation capability for the Primary Containment and reactor coolant pressure boundaries considering a single failure of the installed rig. Additionally, the valves were under direct control of an Operator to provide quick re-isolation if necessary (complies with Technical Specification allowable actions for inoperable containment isolation valves). The possibility for an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The margin of safety was not reduced.
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Section B - Procedure / Miscellaneous Changes l
- This section contains brief descriptions of Procedure / Miscellaneous changes completed during the period March 1,1997 through September 30,1998, and summaries of the safety evaluations for those changes, pursuant to the requirements of 10 CFR Section 50.59(b). All changes were reviewed against 10 CFR 50.59 by the Duane Arnold Energy Center (DAEC) Operations Committee. None of the changes involved unreviewed safety questions.
SE 97-05 T7:hnical Specification Bases And UFSAR Changes Concerning Standby Gas Treatment (SBGT) And Standby Filter Unit (SFU)
Descrintion and Basis for Chance The Standby Gas Treatment (SBGT) and Standby Filter Unit (SFU) sections of the Technical Specification Bases and UFSAR were updated to clearly state testing requirements. Revisions were also made to the purchase requirements for new carbon and High Efficiency Particulate Air (HEPA) filters.
Previously the Technical Specification Bases and UFSAR stated the to ting of the SBGT System and SFU charcoal adsorbers would be in acco dance with USAEC Report DP-1082 and RDT Standard M-16-lT.
These two documents were current when this information was added to the Bases /UFSAR. They were called out in Regulatory Guide 1.52, Revision 1, dated July 1976, as the appropriate standards. Revision 2 of Regulatory Guide 1.52 replaced the reference to DP-1082 with ANSI /ASME N509 and ANSI /ASME N510. DP-1082 was the Final Progress Report on the Savannah River Test Facility in 1966 to develop a Standardized Nondestructive Test of Carbon Beds for Reactor Confinement Applications. The N509 & N510 documents were intended to take what the industry had learned and standardize the design, inspection, and testing. Therefore, where appropriate, these newer standards have been incorporated into the DAEC Surveillance Program. ASTM D3803 replaces RDT Standard M-16-lT. ASTM D3803, was first approved in 1979, and was revised in 1989 to improve the repeatabil y of the testing.
Technical Specification Bases Tables described the physicJ oroperties and test standards for new carbon. These tables were similar to 1ath 2 from Regulatory Guide 1.52, Revision 1. The standard for testing new carbon
, was ASME N509-1980 Table 5-1. Table 5-1 was modified to be DAEC
( site specific and replaced the Technical Specification Bases tables. The physical properties of carbon are exactly per N509-1980 Table 5-1.
( Performance requirements of the new carbon is as follows:
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- The methyl iodide test at 30 C and 95% Ril (relative humidity) is the limiting test for the SFUs and is listed in the standard table.
(Testing the SFUs to 3% penetration ensures a factor of safety to -
the 90% given in the UFSAR.)
e The SBGT System has heaters to reduce the relative humidity to 70% and has an efficiency of 99% credited in the UFSAR '
Regulatory Guide 1.52 states that the carbon is to be tested to a maximum methyl iodide penetration of 0.175%. Therefore, a methyl iodide test at 30"C,70% relative humidity, with a 6" bed depth with a maximum penetration of 0.175% is performed. This !
is the limiting test for the SBGT System.
l- Previously Technical Specifications Bases required testing with blemental
- i. iodine. This requirement was deleted. The suppliers of carbon have developed carbons that meet the requirements of ANSI /ASME N509-1980 Table 5-1 which include an elementaliodine test. Documentation of the i results are not necessary for the specific lot of carbon supplied to DAEC.
For SBGT, the carbon sampling method was updated to state that the design allows removal directly from the bed. Any reference to a grain thief was deleted because other methods of obtaining a representative sample might be used. The design of the SBGT room does not allow the use of a grain thief, so other methods of getting a representative sample ,
have to be used. The intent of the grain thief was to allow a sample to be i taken from an area in the carbon bed that " sees" normal system flow. The Surveillance Test Procedure ensures that the sampling is done in an area l that " sees" normal system flow. I L Specific revision numbers of standards are no longer listed in the l
- l Technical Specification Bases, and are controlled in lower tier plant
!- documents.' Technical Specification Bases reference to ANSI N101.1-1972 was deleted since the requirements of N510 adequately address these testing requirements. The requirement for the llEPA filters to withstand iodine removal sprays was deleted because it is not applicable to the DAEC. The requirement to test the HEPA filters at a USNRC approved location has been deleted. Testing is conducted at a facility approved in I conjunction with DAEC's Quality Assurance Program. Since Regulatory l
Guide 1.52 provides methods that are acceptable to the NRC staff for
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implementing the regulations, specific testing methods that are consistent i with the Regulatory Guide are acceptable, and are not considered an j unreviewed safety question.
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Safety Evaluation Summary The SBGT Systems and SFUs are not accident initiators, and cannot cause an accident. The updated testing requirements will ensure that the systems will perform as described in the UFSAR. The SBGT System and SFUs are accident mitigation systems. The changes described for the system testing and component purchases work together to ensure that the systems will perform at a level equal to or better than that asstuned in the UFSAR.
The systems will continue to ensure that the radiological consequences of an accident are bounded by that described in the UFSAR. The performance of the SBGT System or SFUs will not affect the performance of any equipment important to safety. They mitigate airborne dose. These tests will detect potential problems that could be indicative of equipment malfunctions within the given systems. The testing requirements described will continue to ensure that the equipment will perform as intended and as described in the UFSAR. The SBGT Systems and SFUs will function as assumed in the UFSAR with the new testing requirements; therefore, the ability of these systems to mitigate a malfunction is not reduced. None of the new requirements negatively affect the systems.
These requirements ensure that the systems are maintained fully capable of performing their intended functions. The SBUT System and SFUs cannot create an accident of a difTerent type than previouMy evaluated. The systems will continue to ensure that radiological consequences as described in the UFSAR remain unchanged. Improving the SBGT System and SFU testing willimprove the performance of the systems, and will not adversely affect any plant systems. No new failure modes are introduced by these testing requirements. The new testing requirements will ensure that the systems perform as assumed in the UFSAR under conditions that 1 are likely to exist during an accident. Therefore, the margin of safety will not be reduced. I SE 97-15 Drywell Cooling Water System P&ID Revision Description and Basis For Chance A handswitch and associated lights which were for a future Cleaner Injection System which has not been installed in the backwash portion of l the Drywell Cooling Water System were removed under a previous modification. Subsequently it was identified that these components still appeared on the Drywell Cooling Water System P&lD. In addition some editorial errors were identified on the Backwash Loop 'B' legend and on the loop designator for a valve on the same diagram.
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- Safety Evaluation Summary These diagram changes did not affect the operation of the system and did not increase the probability of an accident previously evaluated in the i SAR. Drywell cooler capability or leakage was not impacted by this change. These changes did not increase the consequences of an accident !
previously evaluated in the SAR. The diagram correction did not increase i the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR. This change did not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR, nor did it create the possibility of an accident of a difTerent type than any evaluated previously in the SAR and the possibility of an accident of a different type than any evaluated previously in the SAR was not created. The possibility ofleakage or the ability to cool the Drywell was not affected. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The system functions as it previously did. The ability of the system to maintain the Drywell temperature was not affected. This was a documentation change only and the margin of safety as defined in the Technical Specifications was not reduced.
l SE 97-30 UFSAR Change For Reactor Recirculation (Recire) System Motor-Generator (MG) Sets Minimum Speed Description and Basis For Chance l l
Prior to startup from Refueling Outage 14,it was determined the as-left minimum speed for the 'B' Recirc MG set (approximately 22%) did not ;
match the UFSAR. Additionally the 'A' Recire MG set was running at 1 24.7% speed at " minimum" speed. The minimum MG Set speed sets the minimum Reactor Recirculation Pump speed. The speed indicating controller will limit the output if either the Recirculation Pump discharge valve is not fully open or total Feedwater flow is less than 20% of rated.
This limited output signal will reduce the generator speed to the minimum speed. This limiting action is to prevent pump overheating should the discharge valve be closed and to protect the Recirculation Pump against possible cavitation due to low Feedwater flow. Minimum MG Set speeds set at 22% and 24.7% prevents pump overheating during the short periods of time the discharge valve is closed, and prevent cavitation.
The Feedwater flowing into the reactor annulus during operation provides subcooling for the fluid passing to the Recirculation Pumps, thus determining the additional net positive suction head (NPSH) available beyond that provided by pump location below the vessel water level. If 106
Feedwater flow is below 20%, the Recirculation Pump speed is automatically limited. This limit is chosen to prevent pump cavitation even for operation with the suction pressure available only from the !
reactor vessel water column above the pump. The MG Set minimum speed setting ofless than 28.7% MG Set speed will maintain sufficient Recirculation Pump and Jet Pump NPSH under all operating conditions including Single Loop Operation The UFSAR was revised to indicate minimum speed is set by the scoop tube positioner electrical stops and is 20-28% speed.
Safety Evaluation Summary A minimum MG Set speed between 20 and 28% could not increase the p probability of occurrence of an accident evaluated previously in the SAR.
The MG Sets will run back to or run at minimum speed when required.
l This action did not initiate an accident. A minimum speed between 20 and l 28% could not increase the consequences of an accident evaluated
! previously in the. SAF,. The slightly higher MG Set and Pump speed l _ produces a slower flow reduction so that margins to fuel thermal limits are
! greater for a complete loss of Feedwater than for a two Recirculation l
L Pump trip. The Recirculation runback response is bvanded by the events evaluated as non-limiting in the UFSAR. The simultaneous runback of both Recirculation Pumps is discussed in the UFSAR as a non-limiting I event. The failure of a Recirculation speed controller to maximum speed is discussed in the UFSAR as a non-limiting event. A minimum speed between 20 and 28% will not increase the probability of occurrence of a i malfunction of equipment important to safety evaluated previously in the SAR. This change did not cause Recirculation System NPSH concerns. A l minimum speed between 20 and 28% could not increase the consequences ,
of a malfunction of equipment important to safety evaluated previously in !
the SAR. The minimum speed runback is not a required safety action for l any analyzed event. A minimum speed between 20 and 28% could not create the possibility of an accident of a different type than any evaluated previously in the SAR. This change did not cause Recirculation System NPSH concerns. This change did not create the possibility of a malfunction of equipment important to safety of a different type than any
- evaluated previously in the SAR. A minimum speed between 20 and 28%
could not reduce the margin of safety as defined in the basis for any Technical Specification. The minimum speed runback is not a required '
safety action for any analyzed event. The MG Set minimum speed setting is not discussed in the basis for any Technical Specification.
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I SE 97-31 Revision To Equipment Radwaste System P&lD Description and Basis For Channe The details of the instrument air supplies thr two control valves were not I i
shown on the P&ID for the Equipment Radwaste System. It was determined that an isolation valve, pressure control valve (PCV), Ulter, and line lubricator needed to be added for each control valve. These components were added to the P&lD to reflect current plant conGguration.
1 Safety Evaluation Summarv i l
This activity did not increase the probability of occurrence of an accident i previously evaluated in the SAR. The addition ofinstrument air detail to the P&ID did not impact the operation of the affected valves. These document changes did not affect the operation of the system and they did l not increase the probability of an accident previously evaluated in the l SAR. The control valves operate as they did previously. This change did not increase the probability of an inadvertent liquid radwaste release, and the consequences of an accident evaluated previously in the SAR were not increased. Adding detail to the air supply lines did not increase the l
probability of occurrence of a malfunction of equipment important to i safety previously evaluated in the SAR. The malfunction modes and their probability have not changed. This activity did not increase the ;
consequences of a malfunction of equipment important to safety evaluated I previously in the SAR. These changes did not increase the consequences of a liquid radwaste release resulting from malfunction. The possibility of an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The ability of the system to prevent inadvertent liquid radwaste releases in excess of 10 CFR 20 limits will not be affected.
This documentation change did not reduce the margin of safety as defined in the Technical Specifications.
SE 97-36 Reactor Recirculation System P&lD Revision Description and Basis For Channe The details of the Instrument Air supply to the Reactor Recirculation System Sample Line Outboard Isolation Valve were not shown on the Reactor Recirculation System P&lD. An air isolation valve did not appear on the P&lD. An isolation valve and a pressure control valve also needed to be added to the P&lD for the control valve air supply. These 108 l
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components were added to the P&lD to reflect the current plant ,
configuration. I I
Safety Evaluation Summary The addition ofinstrument air detail to the P&lD does not impact the operation of the affected valve. These document changes did not affect the operation'of the system and do not increase the probability of an accident previously evaluated in the SAR. The Reactor Recirculation System Sample Line Outboard Isolation Valve operates as it did previously and the probability of a release of radioactive material is not increased.
Adding detail to the air line on the P&ID did not affect the operation of the system. The P&lD revision did not increase the consequences of an unacceptable release of radioactive material. Adding detail to the air j supply line will not increase the probability of occurrence of a malfunction '
of equipment important to safety previously evaluated in the SAR. The malfunction modes and their probability did not change. Changes to the P&lD did not increase the consequences of an unacceptable release of radioactive material resulting from malfunction. Increasing the detail on the P&lD did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The possibility of an unacceptable release of radioactive material was not affected. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. This change only increased the air supply detail on the P&lD and did not make any changes to the system or its operation. The ability of the system to l '
prevent or mitigate an unacceptable release of radioactive material was not impacted. The system functions as it previously did. The ability of the system to reliably isolate all pipes necessary to establish the Primary Containment barrier was not reduced. This is a documentation change only and the margin of safety as defined in the Technical Specifications has not been reduced.
SE 97-44 Condensate Demineralizer P&lD Revision Description and Basis For Chance The following discrepancies identified in the P&lD for the Condensate Demineralizer System were corrected:
i e An instrument air header spare connection valve should have been shown closed but was shown open.
- The handswitch for the main drain for the 'E' filter was inadvertently removed when the drawing was redrawn.
109
I Control air pressure regulators shown in the 'E' loop were shown with identification numbers for the ' A' loop.
Safety Evaluation Summary The Condensate Demineralizer System is not specifically designed for any design basis event since it is not a safety system. The Condensate Demineralizer System is also not required to meet the single failure criteria. Because this docwnent change did not affect the system performance in a manner that could lead to an accident, the probability of an accident previously evaluated in the SAR is not increased. The Condensate Demineralizer System is not required for the mitigation of any accidents defined in the NSOA and UFSAR. This document revision did not prevent the Condensate Demineralizer System from performing its design basis function as defined in the SAR. Therefore, this change did not prevent nor degrade any essential safety function assumed by the NSOA to mitigate the consequences of design basis accidents. This change does not increase the consequences of an accident evaluated in the SAR and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased.
These drawing corrections did not challenge a fission product barrier more severely than those already analyzed in the UFSAR and the NSOA. The design bases for the condensate demineralizer System is not affected by this change. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and the possibility of an accident of a different type than any evaluated previously in the SAR were not created. The possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR was not created. This document revision did not change or affect any setpoints or surveillance requirements as required by Technical Specifications. Therefore, this change did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-45 Special Test Procedure (SpTP) 196 - Control Building / Standby Gas Treatment System Instrument Air Compressors Description and Basis For Chance The purpose of this Special Test was to provide detailed instructions concerning the performance of an air leak check for the Control Building / Standby Gas Treatment System safety related air compressors, 1K-3 and 1K-4. The 1K-3 and 1K-4 Air Systems are nearly identical in that they have nearly the same number ofloads and amount of piping.
110
However, the 1 K-3 compressor duty cycle was approximately 34% greater than that of 1K-4. This Special Test was written to determine the air consumption through the major branches of the 1K-3 and 1K-4 Air Systems. The test data between the iK-3 and 1K-4 Air Systems were compared, then the sources of the high air consumption in the 1K-3 System were identified.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The failure of the Safety Related Air System and the systems it supports are not precursors of any accidents ;
afTected by this test and are not initiators of any accidents previously l evaluated in the SAR. This test was conducted when the plant was in Cold Shutdown and the affected plant systems were not required by plant Technical Specifications to be operable to mitigate the consequences of an accident evaluated in the SAR. Therefore, this activity did not increase the consequences of an accident evaluated previously in the SAR, and this activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The purpose of the testing was to identify and remedy excessive sources of air usage which result in high duty cycles to 1K-3 and 1 K-4. The consequences of this testing was evaluated against the Design Safety l Standards, the Nuclear Safety Criteria for Boiling Water Reactors, the Nuclear Safety Criteria, the Classification of DAEC Structures, Systems, and Components, and the NSOA. This test operated the Safety Related Air System well within the design criteria established in the documents listed above. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. Throughout the performance of this test the Saiety Related Air System did not exceed specified limits regarding air quality and system pressure. Stainless steel compression fitting tees were permanently installed in the Air System to allow connection of temporary air supplies to maintain air supplies to certain loads when their normal safety related supply was isolated. These fittings are Quality Level 1 stainless steel and conform to design requirements. The fittings were determined to be suitable for use on the existing copper piping. After completion of testing, the temporary air supplies were removed and the tees were capped. The risk of additional system leakage resulting from these fittings is considered
- negligible compared to the number of existing fittings in the system. This l activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and it did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This testing did not involve changes to iK-3 and 1K-4 or to their operation.1K-3 and 1K-4 operated normally 111
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i throughout the test. The testing did not change the way the system performs other than the temporary isolation of air to systems that were not l required to be operable while in Cold Shutdown. The installation of compression fitting tees did not introduce a new failure mode for the system. The test involved operating plant systems by normal means other i
than the temporary isolation of systems loads. The installation of the compression fittings did not introduce a different failure mode to the system. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-54 (Revision 1) UFSAR Changes and Revision 33 of Fire Plan Volume I Description and Basis For Chance L
The following changes were made to UFSAR Table 9.5-1 and Volume 1 of the DAEC Fire Plan resulting in Revision 33:
Changed the scope of the operability and surveillance requirements to address only those Fire Protection Systems and features required for compliance with 10 CFR 50 Appendix R. Therefore, the operability and surveillance requirements pertaining to various detection, suppression and hose station systems that are not required for Appendix R compliance have been removed. Fire '
Protection systems which are required for Appendix R compliance as a result of the Appendix R re-baseline analysis have been added.
UFSAR Table 9.5-1, Fire Plan section 12.0 and the Fire Plan Bases section of the Fire Plan Operability Requirements have been revised accordingly.
. Changed the plant's carbon dioxide tank volume from a 90%
minimum level to an 80% minimum level. This reduces frequent and unnecessary deliveries of carbon dioxide while maintaining sufficient volume to provide two full discharges, with margin, to the Cable Spreading Room.
. Removed the surveillance requirement to demonstrate redundant fire pump operability on a daily basis when one fire pump is considered inoperable. This is consistent with Amendment 174 which removed daily operability checks for safety related systems.
. Included fire dampers in the list of fire rated assemblies required for Appendix R that have operability and surveillance requirements.
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e Revised the Fire Plan fire watch requirements to allow the Fire j
Marshal to evaluate whether a fire watch is required or other measures may be used if the fire watch is required to enter a high- !
radiation area, a contaminated area or an area of the DAEC where the fire watchers presence could endanger the safe operation of the DAEC or where a personnel safety hazard is likely to exist. !
. In the Operability Requirements, included llose Station 36 as a compensatory measure for inoperable Cable Spreading Room suppression. This liose Station, as well as Station 35, support !
compliance with the NRC Safety Evaluation Report dated June 1, I 1978.
Administrative changes were made which reflect plant common practices ,
or describe methods of meeting NFPA standards. Changes in titles, '
department names and organization charts were made to reflect the current organization.
Safetv Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. Fire is not an entry condition, basis or an assumption for any accident previously evaluated in the UFSAR and the NSOA. The changes did not increase the probability of any of the following events occurring: a fire, inadvertent actuation of a suppression system or loss of essential liVAC or other plant equipment credited in the safe shutdown analysis. This activity did not increase the consequences of an accident evaluated previously in the SAR. The probability of a malfunction was not changed, and the consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not l
increased. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The changes resulted in system operability and reporting requirements to reflect Appendix R compliance strategies but did not change the DAEC maintenance and surveillance of all Fire Protection System., using industry standards and practices as a guide. This change did not reduce the margin of safety as defined in the basis for any Technical Specification. Fire i Protection systems do not form the basis for any Technical Specification safety margins.
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r SE 97-56 Control Rod Drive Hydraulic System P&ID Revision i
Descrintion and Basis For Chance l This change revised the P&ID for the Control Rod Drive Hydraulic I
- j. System by deleting extraneous and wrong information depicting a 3/4 ,
l- inch-DBA nipple as 1/2 inch-EBB. According to the original plant l installation, no 1/2 inch nipple was part of the components installed. l Hence, the indication of the nipple as 1/2 inch-EBB in the P&lD was an error.
l l Safety Evaluation Summary I This activity did not increase the probability of occurrence of an accident
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I' evaluated previously in the SAR. This change only corrected a documentation error in the P&lD while showing the piping class code for a component. The installed component is superior to the one wrongly shown in the P&lD. The installed component envelops the code i
requirements mentioned in the DAEC UFSAR. Since the installed component meets / exceeds the code requirements mentioned in the UFSAR, the document correction reflecting the plant installed
- configmation does not increase the consequences of an accident evaluated previously in the SAR. This change did not increase the probability of l occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This change did not create the possibility of an accident of a f
different type than any evaluated previously in the SAR. This change did l not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the SAR. The margin of safety is not reduced because the actual installed component was of superior class as compared to the one erroneously depicted in the P&ID.
SE 97-58 UFSAR Revision For Fire Door Inspection Requirements Description and Basis For Chance Table 9.5-1 of the DAEC UFSAR stated " fire doors shall be verified to be functional semi-annually to verify operation of auto-closing devices and latch mechanisms". This UFSAR commitment was based on Section 4.9 of the NRC Safety Evaluation Report (SER) dated June 1,1978. This section of the SER indicates that the DAEC will periodically inspect fire doors to verify their status. The requirement is that fire doors will be
- inspected . semi-annually to verify that self-closing mechanisms and latches are in good working order.
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I This UFSAR change involved the following two types of doors:
- Access doors for cable and piping chases where the access hatch consists of a normally closed, non-standard size, locked fire door providing access to a confined space. These doors were not supervised and therefore per the UFSAR and SER required inspection on a weekly basis.
Locked fire doors which are located entirely within a locked high radiation area. Only the Steam Tunnel Airlock Fire Door met this condition. The door was supervised and therefore per the UFSAR and SER required inspection on a semi-annual basis.
The inspection frequency for these doors has been changed to a once per fuel cycle basis. This change involved a revision to the DAEC UFSAR.
Safety Evaluation Summary Fire is not an entry condition, basis or an assumption for any accident previously evaluated in the SAR. The revision of the UFSAR to exclude locked doors located entirely within high radiation areas and chase access doors, did not increase the probability of an accident. These doors are not normally used during plant operation. The chase doors meet the access door criteria as defined in NFPA 80. They are smaller than a conventional door and may require crawling through. The chases are not normally accessed, are not traffic areas and transient materials are not stored in these chases. The Steam Tunnel Airlock Fire Door is located entirely within a high radiation area and is a Secondary Containment airlock door. This door must be closed and verified closed by the airlock logic circuit to allow egress from the Steam Tunnel. NFPA 80 indicates that " fire doors are of value only if properly maintained so that they close or are closed at the time of fire" The NFPA provides no recommended inspection frequency. The NFPA indicates that an inspection program "should be implemented and should be the responsibility of the property management". These doors are normally closed, labeled as fire doors and are locked. Their function is to remain closed in the event of a fire. They are not located in areas of high humidity or in areas which could otherwise cause rapid deterioration of the fire door. The char.ge in inspection frequency to once per fuel cycle will not reduce the effectiveness of these doors to contain the spread of a fire. While fire is a possibility, fire is not i
an entry condition, basis or an assumption for any accident previously evaluated in the SAR. Therefore, this change did not increase the probability of occurrence of an accident evaluated previously in the SAR, 115
and it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This change does not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. These doors are not pressure rated and are not required to remain closed in the event of high differential pressure conditions to contain the effects of a Iligh Energy Line Break (HELB) or other line break. The change in inspection frequency to cyclic did not reduce the effectiveness of these doors to contain the spread of a fire. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a ditTerent type than any evaluated previously in the SAR was not created.
Fire Protection features at the DAEC are not addressed by the Technical Specifications. Therefore, this change did not impact the margin of safety as defined in the basis for any Technical Specification.
SE 97-60 Service Water System in Pumphouse P&lD Revision Description and Basis For Chance As verified by a walkdown to determine actual plant configuration and according to an Operating Instruction, the valve V46-0238 is the Radwaste Dilution Vent Valve, and valve V46-0199 is the Radwaste Dilution Drain Valve. Ilowever, the P&lD incorrectly identified the vent as V46-0198 and the drain as V46-0238. The P&lD was revised to correctly depict V46-0238 as the vent valve and V46-0199 as the drain valve.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. This activity did not make any changes to the plant. Only a document change was made to correctly identify two non-safety related valves. These valves are identical and meet their service conditions. The SAR was not impacted by this change. This activity did not increase the consequences of an accident evaluated previously in the SAR, and it did not increase the probability of l occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and it did not create the possibility of a malfunction of equipment important to safety of 4
a different type than any evaluated previously in the SAR. This change did not reduce the margin of safety as defined in the basis for any Technical Specification since it involved only a document change to l
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correctly identify two non-safety related and non-Technical Specifications related valves.
SE 97-62 Auxiliary Boiler Heating System P&lD Revision '
Description and Basis For Chance The following discrepancies were identified in the P&lD for the Auxiliary i Boiler Heating System:
. The Operating Instruction (OI) indicates the position of the Demineralized Water Float Control Inlet and Outlet valves are normally closed. The P&lD showed these valves open since the time of original installation.
. A drafting error was identified which wrongly showed the Boiler Blowdown Line Drain, as normally open. The valve should have been shown normally closed per the applicable 01.
. The P&ID had the pipe line 4"-GBD-31 (to turbine steam seal) not shown connected to a 8"-GBD-29 line.
l The P&lD has been revised to show the valves nonnally closed, and the
- pipe linejoined. I Safety Evaluation Summary This change revised a P&lD regarding the normally closed position of certain non-safety related valves, in accordance with the current 01. It l also corrected a typographical error made during revision to this P&lD.
This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR, and it did not increase the consequences of an accident evaluated previously in the SAR. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased, and the consequences of a malfunction of equipment important to safety evaluated previously in the l- SAR is not increased. This change did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and i it did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This l change did not reduce the margin of safety as defined in the basis for any Technical Specification.
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SE 97-63 UFSAR Change To Clarify Descriptions of Post-Accident Radiological Source Terms Descriotion and Basis For Chance ,
l This activity consisted of updates to the UFSAR description of the source tenn assumptions used in post-accident radiological dose calculations.
These updates clarified that the analyses have used source term isotope lists that are conservative with respect to the source term isotopes listed in
' Technical Information Document, TID-14844. The previous descriptions indicated that the TID source term isotope list was used directly.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. This activity involved clarification of the SAR descriptions of assumptions about radiological source terms used in the analysis of accidents. It had no impact on the initiators or probability of an accident. This activity did not increase the consequences of an accident evaluated previously in the SAR. The actual source terms used in the analyses, although not identical to the TID-14844 source term, are both more conservative and more plant specific than those that would have been obtained from Table IV of TID-14844. This activity brought the UFSAR description of the assumptions into agreement with the actual assumptions used in the analyses. The use of more conservative source term assumptions indicates that the consequences could be decreased; however, the differences between the source terms are considered to be insignificant. There is no basis to conclude that this change would increase the consequences of an accident previously evaluated in the UFSAR. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased.
Since the analyses supporting the qualification of equipment have been based on assumptions of a source term that is equal to or greater magnitude than the TID-14844 source term, the equipment may be qualified to radiation doses that exceed the actual accident total integrated dose. Therefore, the probability of equipment malfunction is not increased, and the consequences of a malfunction of equipment important to safety evaluated previously in the SAR is not increased. The use of conservative source terms for qualification of safety related equipment makes it less likely that equipment will fail, but has no impact on the consequences of equipment failures beyond those already evaluated in the SAR. This change did not create the possibility of an accident of a different type than any evaluated previously in the SAR. This change had
- no impact on the types of accidents, transients, or malfunctions that could occur. This activity did not create the possibility of a malfunction of 118
equipment important to safety of a different type than any evaluated previously in the SAR. This change did not reduce the margin of safety as defined in the basis for any Technical Specification. Since the analyses used source terms that are more conservative than the TID-14844 source term, the DAEC has personnel dose predictions and equipment qualification requirements that have increased margins of safety over what would be required.
SE 97-65 Containment Atmosphere Control System P&ID Revision Description and Basis For Chance A discrepancy was identified between the P&lD for the Containment Atmosphere Control System and the OI for this system, regarding the normally open/ closed position of a drain trap inlet isolation valve. The P&lD showed it as closed, but it is required to be normally open, per the
- 01. This was considered a drafting error, and the P&lD has been revised ,
to depict the valve as normally open.
l Safety Evaluation Summarv !
This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. Tms change only corrected a drafting ,
i error in the P&lD to corrcetly depict the position of the non-safety related valve as normally open, in accordance with the 01. The consequences of an accident evaluated previously in the SAR are not increased, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR is not increased. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and the possibility of an accident of a different type than any evaluated previously in the SAR is not created.
The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created.
This change had no impact on the margin of safety as defined in the basis for the Technical Specifications.
SE 97-66 Revision To Control Rod Drive Ilydraulic (CRD) System P&lD Description and Basis For Chance A discrepancy in a note for a CRD Flow Stabilizing Valve on the CRD System P&lD was corrected. The note stated,"SV1843B closes on rod withdrawal when 'B' stabilizing valves are selected..." It was determined this error existed since the note was added by a previous drawing change.
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l l The note has been revised to read, "SV1843B closes on rod withdrawal when 'A' stabilizing valves are selected...".
l Safety Evaluation Summa _ry This activity did not increase the probability of occurrence of an accident i
evaluated previously in the SAR. This change only corrected a documentation error in the P&lD. The CRD stabilizing valves operate as they did prior to this change. Only the note on the P&ID was revised.
l This activity did not increase the consequences of an accident evaluated previously in the SAR, and it did not increase the probability of occurrence of equipment important to safety evaluated previously in the SAR. This change did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The electrical function of the stabilizing valves was not alTected.
l The consequences of a malfunction of equipment important to safety evaluated previously in the SAR did not increase. The possibility of creating a new accident was not created, and the possibility of a malfunction of equipment important to safety of a different type than l previously evaluated in the SAR was not created. The margin of safety was not reduced. The CRD System was not physically altered from its design intent.
SE 97-69 Containment Atmosphere Monitoring (CAM) System P&ID Revision l
Descrintion and Basis For Channe The control valves that are in the sample pump bypass lines for CAM l System Loop A and Loop B, were not shown on ti e P&lD. These valves were not assigned unique equipment identification . ambers in the past.
l The P&lD at the time of original startup did not show these valves or the associated pipe lines. The P&ID has been revised, and the equipment has l been labeled in the plant to show and/or identify these valves. In addition, the Post Accident Sampling System return line was shown tapping into the Torus sample return line in the wrong location. This error was made in a previous P&lD revision. The P&lD was revised to correct this error.
! Safety Evaluation Summary This change did not make any alteration to the plant. It only revised the plant documentation to label two non-safety related control valves and corrected a drafting error in the CAM System P&lD. No new failure modes were created by this change; therefore, this activity did not increase
, the probability of occurrence of an accident evaluated previously in the SAR. This activity did not impact 10 CFR 100 limits, and therefore, it did 120
not increase the consequences of an accident evaluated previously in the SAR. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased, and the consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased. No new failure modes were created by this change, and therefore, this change did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and it did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. Surveillance Test Procedures and Technical Specifications were not affected. Therefore, this change did not reduce the margin of safety as defined in the basis for any Tecimical Specification.
SE 97-73 Well Water System P&ID Revision Description and Basis For Chance The Low Tank Level Cutout Switch for the Well Water Chemical Injection Pump and Tank is an accessory for the Well Water Chemical Injection Tank. The tank, along with the switch was installed by a previous modification. The switch and the associated wiring was never shown on the P&lD. Providing a plant ut.m.e identification for the switch and showing it on the P&ID is helpful for r.ny maintenance activity to be carried out on the switch. Discrepancie.e were also identified between the Well Water System P&lD and the DA'dC Fire Protection System. The P&lD showed several Fire Hose Stations (FHS), with valves shown in the open position. In the plant these valves are closed and no fire hoses are connected to them. Furthermore, the Well Water System is not credited with supplying water to the plant's hose stations. Therefbre, the P&lD was revised to reflect the actual plant configuration. A change was also required to correct an editorial type error concerning a reference listed in the drawing.
l Safety Evaluation Summary This modification revised plant documentation to provide a unique plant identification to a non-safety related level switch and show this level switch and associated wiring on the P&lD. This modification also revised i the P&lD for the Well Water System in accordance with the actual plant l configuration by deleting the FHSs that are not part of the Fire Protection System. The Well Water System is non-safety related. The Well Water System is not an accident initiator or mitigator. IIence this activity did not increase the probability of occurrence of an accident evaluated previously in the SAR, and it did not increase the consequences of an accident t
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evaluated previously in the SAR. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The possibility of an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a different 1
type than any evaluated previously in the SAR was not created. This change and the systems affected by this change are not described in Technical Specifications. Therefore, this activity does not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-76 UFSAR Change Revising T-scal Replacement Interval For Drywell Vent and Purge Valves and Reactor-Torus Vacuum Breakers Descrintion and Basis For Chance This UFSAR Change revised the replacement interi al from 4 to 7.5 years for the T-seals in the Drywell Vent and Purge Vdves and the Reactor-Torus Vacuum Breakers. The functions of these valves are for Primary Containment pressure boundary and Primary Containment isolation. The vent and purge valves also have a function for containment pressure control. The vacuum breakers have a pressure relief function.
The seals are replaced periodically to prevent seal failure which could result in an excessive radioactive material release from Primary Containment or oxygen in-leakage into containment. The leakage integrity test history of these valves has been very good, with no degradation in leak tightness and consequently T-ring seal integrity being observed. The continued demonstration of seal integrity, via leak rate tests, helps to ensure that the possibility of a sudden, catastrophic failure of a T-ring seal remains highly unlikely.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The T-seals are not initiators of any accident evaluated in the SAR. Changing the T-scal replacement frequency did not affect the capability to close and seal, and it did not l increase the probability of an accident occurring. Since 1982, significant information has been gathered concerning the perfonnance of these valves
! and in particular the T-seal. This information has been gathered through periodic replacement inspections, local leak rate testing, quarterly leakage integrity surveillance testing, and through inservice failures. In addition, an engineering evaluation has been performed to determine the service hfe. Based on this information, the replacement frequency of 7.5 years (or 122 l
five 18-month cycles)is appropriate. This activity did not increase the consequences of an accident evaluated previously in the SAR. The accidents as identified in the NSOA for which the Containment Vent and Purge Valves or Torus-Reactor Building Vacuum Breakers are required are the control rod drop accident, the LOCA inside containment, and the LOCA outside of containment. The safety actions for these valves include reactor vessel isolation, establishing Primary Containment, and establishing Secondary Containment. In all actions the valves are required to close to avoid the unacceptable result of a radioactive material release exceeding 10 CFR 100 limits. Since the valves are capable of functioning as intended under worst case conditions they will also function as intended under the lesser conditions of a control rod drop accident or LOCA outside of containment. The probability of occurrence of a malfunction of ;
equipment important to safety evaluated previously in the SAR did not increase. The continued demonstration of seal integrity, via the leak rate tests, helps to ensure that the possibility of a sudden, catastrophic failure of a T-ring seal remains highly unlikely. Changing the replacement !
frequency did not increase the probability of occurrence of a malfunction of the T-ring seal, and the consequences of a malfunction of equipment i important to safety evaluated previously in the SAR were not increased. I Changing the frequency of T-seal replacement did not create the possibility of different types of malfunctions, and the margin of safety as ,
defined in the basis for any Technical Specification was not reduced. I SE 97-77 Fuel Pool Cooling and Cleanup System P&ID Revision l l
Description and Basis For Chance i
1 An error was identified in the P&lD for the Fuel Pool Cooling and Cleanup System. In depicting the symbol for the Condensate Make-up l
Line Check Valve to Fuel Pool Cooling and Cleanup Check Valve the ;
symbol incorrectly showed the direction of the flow away from the Fuel l Pool Skimmer Surge Tank, whereas the flow in the actual plant configuration goes into the tank. A previous drawing revision depicted the l check valve indic ating flow in the wrong direction. The change package did not contain any mention ofintended technical changes. Hence, this was considered an inadvertent error. The P&lD has been revised to correctly depict the check valve direction.
An enhancement was made to correctly depict the T-junction between two lines. Thejunction area of the lines has been redrawn as an enhancement to the P&lD, to avoid any possible confusion. The P&ID incorrectly indicated that the boundary of Seismic Category I classification of the RHR Discharge to Fuel Pool Storage line extended beyond a manual isolation valve. This was contrary to the UFSAR position on the seismic 123
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l classification boundary for this line. The UFSAR indicates that this line is !
Seismic Category I from the RHR System up to and including the closed l
isolation valve. Hence, the flag showing the Seismic Category I boundary in the P&ID was shifted to the valve.
Two Piping Class Sheets were also revised. These sheets indicated that !
" the entire line was Seismic Category I. This was contrary to the UFSAR I position on the seismic classification boundary for this line. The UFSAR indicates that this line is Seismic Category I from the RHR System up to and including the closed Fuel Pool Cooling System RHR Return Line Isolation Valve. Therefore, the drawing was revised to provide a clarification note indicating that the remaining portion of the line is not Seismic Category I.
1
, Safety Evaluation Summary j i
This activity revised a P&ID and two piping class sheets. This is in accordance with the current plant configuration, plant calculations, and the
- Licensing basis documents. The changes made by this activity did not initiate any accident. This activity does not increase the probability of occurrence of an accident evaluated previously in the SAR. These changes did not impact the 10 CFR 100 limits in any way. Therefore, the consequences of an accident evaluated previously in the SAR were not i
increased. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This ,
, activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. No new failure modes were created by the changes mentioned in this activity. These changes did not cause a malfunction of equipment. Hence, the proposed activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-80 Circulating Water System P&lD Revision Description and Basis For Chance A concern was identified that several equipment identification tags for the Cooling Tower Level Element Air Compressors Receivers and associated
] equipment, and several valves located in the Cooling Tower Breaker
, House, did not match with the P&lD, but did match the Operating
- instructions. The discrepancies were confirmed by a walkdown of the 124
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L Cooling Tower, Pump Ilouse and Ilreaker Houses. The walkdown identified additional discrepancies between the plant and the P&lD. No changes were made to the plant. Only plant documents, an equipment database and identification tags were affected to make them conform to L the current plant configuration.
- Safety Evaluation Summarv l The equipment affected by the changes are non-safety related, and are part L
of the Circulating Water System and the Cooling Tower Systems. Neither of these systems is an accident initiator or mitigator, and the changes did not degrade these systems. Therefore, this activity did not increase the probability of occurrence of an accident evaluated previously in the SAR.
This activity did not impact 10 CFR 100 limits in any way because the subject systems are not included in the NSOA. This activity did not
, increase the consequences of an accident evaluated previously in the SAR, and it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. These j changes did not increase the consequences of a malfunction of equipment
!- important to safety evaluated previously in the SAR. No new credible accident mode is possible for the systems affected by this change. This l- change did riot create the possibility of an accident of a different type than i any evaluated previously in the SAR, and it did not create the pos_sibility of a malfunction of equipment important to safety of a different type than l any evaluated previously in the SAR. The change and the systems affected by the change are not described in Technical Specifications.
Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
SE 97-81 Radwaste Sample System P&lD Revision Description and Basis For Chance I
The Radwaste Sample System P&lD required revision to reflect the plant configuration by indicating the status of certain non-safety related instruments that were either removed, or removed and bypassed in the past.
Safety Evaluation Summarv L This change was related to the decommissioning of certain non-safety p related instruments in the Radwaste Sample System. This change shows
[ the instruments on the P&lD as deleted, in accordance with the current L plant configuration. The bottled grab method of taking samples as required by the UFSAR, and in accordance with the Radwaste Haadling L
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l Procedures can still be used. Therefore, the overall system performance was not affected in any way by this change. Also, the Radwaste Sample System is not an initiator of an accident, so this activity did not increase
- the probability of occurrence of an accident evaluated previously in the SAR. The Radwaste Sampling System beyond the first root valve is
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outside the scope of Regulatory Guide 1.143, that governs the requirements of the Radwaste System and its components. Theafore, this change did not increase the consequences of an accident evaluaud previously in the SAR. The equipment associated with this change is not 1 safety related because the subject sampling is for the liquids within the process and not related to the liquids to be released to the environs. No safety action was impacted because the process sampiing is not safety related. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. There are no radiological consequences due to an assumed malfunctioning of the equipment in the Radwaste Sample System. This l activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR and it did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the i
SAR was not created. Technical Specifications do not specify any margin of safety for the Radwaste Sample System or its components. Also, no surveillance tests are specified for the components / system affected by this change. This change did not reduce the margin of safety as defined in the basis for any Technical Specifications.
SE 97-82 Diesel Generator System P&ID Revision Descrintion and Basis For Chance l
The pipe clamps for the Lo-Lo Alarm Switches for the Standby Diesel Generator (SBDG) Day Tanks were not of the type shown on the isometric piping drawings. The installed clamps are a Unistrut model. The drawings showed U-bolts. A review ofdocumentation showed a lack of evidence for the basis of the replacement of the U-bolts by Unistrut l- clamps. The isometric piping drawings were revised to show the Unistrut clamps.
An error was discovered in the P&ID showing the size of the isolation valves for the Lo-Lo Alarm level switches for the SBDG Day Tanks.
While the installed size of these valves is 2 inch, the P&lD incorrectly showed the size as 1 inch. The correct size,2 inch, is based on design
, drawings (which indicate the connecting pipe as 2 inch), the Equipment Database, and walkdown information. This error was also found in the 126
earlier P&lD version, contained in the original plant startup records.
l Typically the valve size is not indicated on the P&lD. The size indication for the subject valves was deleted from the P&lD, and the line from the L
tank up to the valve is indicated as 2 inch. The seismic boundary shown at the isolation valve was deleted because the complete line, up to the . Level Indicating Controller (LIC), is Seismic Category 1.
Safety Evaluation Summary l
This activity involved revising drawings to reflect the current plant i
configuration. Because the tubing was evaluated to meet Seismic I ;
category requirements even without the clamp, the change is acceptable.
l The other change shows the piping designation as 2 inch for the piping at the isolation valve for the LIC on the P&lD. This was also based on the current plant configuration. The safety function of the valve is to ensure l the integrity of the system pressure boundary. Because the piping was i installed to meet the design, material, and construction standards !
applicable to the Diesel Oil System, the valve meets the safety function, ,
! and, the change in valve size on the P&lD has no impact. Therefore, this ;
l activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. This activity did not change the type of l the tube clamp, and changing the instrument isolation valve size did not
- l. impact the safety function, integrity of the system pressure boundary,
- because the tubing meets the Seismic Category I requirements, and the valve size has no impact on this safety function. Therefore, the system l was not degraded in any way, and it still functions as it was intended. This
! change did not impact the radiological and other consequences of any l
accident analysis described in the UFSAR in any way. Therefore, this activity did not increase the consequences of an accident evaluated previously in the' SAR. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a l' malfunction of equipment important to safety evaluated previously in the L SAR. Even in the event of failure of the affected equipment, the only L credible scenario would be loss of one day tank for the system. This is j enveloped by the SAR. This activity did not create the possibility of an
- l. accident of a different type than any evaluated previously in the SAR, and this activity did not create the possibility of a malfunction of equipment
[ important to safety of a different type than any evaluated previously in the
!~ SAR. The only credible failure mode that can impair the equipment l impacted by this change is rupture of the tubing in a seismic event. This i has been analyzed and found to meet the seismic design requirements and
} the safety standards defined in the UFSAR. This activity did not reduce i
the margin of safety as defined in the basis fbr any Technical Specifications, e
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i SE 97-83 Fire Protection System P&lD Revision l
Descrintion and Basis For Chance l
This change revised the Fire Protection P&lD to show a non-safety related Deluge System Priming Line Supply Isolation Valve in the normally closed position to agree with Operating Instructions, Surveillance Tests, and actual plant configuration.
Safety Evaluation Summary There are no safety related functions associated with this Deluge System.
The Fire Protection System is not an accident initiator. Therefore, this activity did not increase the probability of an occurrence of an accident evaluated previously in the SAR. There are no radiological consequences associated with this change. This activity did not increase the consequences of an accident evaluated previously in the SAR, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased by this activity.
The consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased. The possibility of an accident of a different type than any evaluated previously in the SAR was not created by this activity, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated I previously in the SAR was not created. The Fire Protection System is not part of the DAEC Technical Specifications; therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical i
i Specification, SE 97-87 Procedure To Test Performance Of Control Building Chillers l
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- Description and Basis For Chance The purpose of the Equipment Monitoring Procedure is to collect data from the Control Building Chiller that is in service, to verify the adequate l operation of the Chiller Condenser, and to maintain this data for trending.
i This procedure was completed as part of DAEC's response to NRC
! Generic Letter 89-13 " Service Water Problems Affecting Safety-Related Equipment". The Generic Letter was issued due to the high safety significance of Service Water Systems and the relatively high frequency of j Service Water System failures and degradations in the industry. The
, Generic Letter in part requires Licensees to " confirm by periodic test, the heat transfer capability of all safety-related components cooled by the
- Service Water System". It was determined that the Control Building Chillers are within the scope of the Generic Letter requirement since 128 i
Emergency Service Water (ESW) is used to remove heat from the Chiller Condenser and Oil Cooler during safety related operation.
The Equipment Monitoring Procedure is used to assess the heat transfer capability of the Chiller Condenser while ESW is supplying the cooling removal for the system. The procedure was written to test each Chiller while in service carrying its normal load. The Chiller being tested is still considered operable. The following data is collected and used to determine adequate operation of the Chiller Condenser: ESW flowrate, differential pressure, inlet and outlet temperatures to/from the condenser, chill water flow, inlet and outlet temperatures to/from the evaporator, Chiller load, motor current, compressor discharge pressure and refrigerant temperature to condenser. The procedure was written to operate the Chiller and temporarily install instrumentation to collect this data.
Safety Evaluation Summary The Control Building Chillers and the Closed Loop Chilled Water System are not accident initiators. The Chiller being tested or a redundant Chiller can be started to provide Control Building cooling. This activity does not increase the probability of an accident that has been previously identified and evaluated in the SAR. Operating load and cooling water adjustments are made during the performance of this procedure. The Chiller is not operated outside its normal operating band. In the event that problems are encountered during Chiller adjustments, the Chiller can trip. The trip of a Control Building Chiller is an event that has been anticipated. The Chiller trip will not impact the ability of the Control room HVAC System to isolate the Control Room in the event of an accident previously analyzed in the SAR. This activity will not increase the consequences of an accident evaluated previously in the SAR. This activity will not increase the probability of occurrence of a malfunction of equipment important to safe + evaluated previously in the SAR. Although very unlikely, this activity could cause the trip of the operating Control Building Chiller which is considered an important to safety component. The trip would most likely be caused by high compressor discharge pressure and once cleared, the affected Chiller could be restarted. A redundant Chiller is available and can be started immediately, if necessary. There are adequate cautions and wamings to alert the operator of the steps that could cause a Chiller trip. By following the guidance in the procedure for this activity, it is unlikely that a Control Building Chiller trip will occur. In any case, the design function of the Control Building IIVAC System is not affected.
This activity does not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The loss of a Control Building Chiller due to this activity cannot cause any event different than that analyzed. This activity cannot cause a failure of a 129 i
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I different mode than previously evaluated. A Chiller malfunction of a L different type is not credible. This activity did not reduce the margin of i safety as discussed in the Technical Specification Basis for the Control
! Building Chillers. The controls included in the procedure ensure the Chiller remains operating within the normal operating bands. Since the Chiller is operated in this manner the margin of safety is not reduced.
SE 97-88 ' Control Rod Drive Hydraulic System P&lD Revision I i
Descrintion and Basis For Chance The P&lD for the Control Rod Drive (CRD) Hydraulic System was revised to show the correct plant location of CRD Drive Water and CRD l Cooling Water differential pressure instrumentation.
Safety Evaluation Summary This drawing change did not cause any alteration to the plant. Plant l documentation was revised to show the correct physical plant locations of l l CRD Drive Water and CRD Cooling Water differential pressure instrumentation and to add DAEC equipment identifications. No physical changes to the system were made. This instrumentation is non-safety l related and perform their intended design function. This activity did not l increase the probability of occurrence of an accident evaluated previously in the SAR. There are no radiological consequences associated with this activity. This , 'ivity did not increase the c,nsequences of an accident l evaluated previously in the SAR. The design function (to provide '
differential pressure indication), individual components and the physical I plant arrangement of the subject non-safety related CRD Drive Water and CRD Cooling Water differential pressure instrumentation were not l impacted by this activity. This activity only corrected drawing errors that l l have existed since plant construction. Therefore, this activity did not l increase the probability of occurrence of a malfunction of equipment ;
important to safety evaluated previously in the SAR, and this activity did l not increase the consequences of a malfunction of equipment important to I safety evaluated previously in the SAR. The possibility of an accident of a I l different type than any evaluated previously in the SAR is not created.
l This activity did not create the possibility of a malfunction of equipment l important to safety of a different type than any evaluated previously in the l SAR. No component mentioned in the Technical Specifications was affected. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
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i
l l SE 97-90 UFSAR Change To Criticality Analysis For Progressive and Remote (PsR) Racks and Advance Fuel Designs Descrir> tion and Basis For Chance The UFSAR was changed to incorporate the information associated with the par racks presented in HOLTEC Report 111-971708, " Criticality Safety Evaluation of the Spent Fuel Storage Racks in the Duane Arnold Energy Center for Maximum Enrichment Capability", dated August of 1997. The criticality analysis of the spent fuel pool racks manufactured by Progressive and Remote (par) Ilandling Technologies assumes a maximum uranium-235 enrichment in a fuel lattice (a :;ircinch segment of a fuel bundle) of 3.1 percent. However, a portion of the fuel bundles at the DAEC have a maximum lattice enrichment of up to 3.68 percent U-235.
Therefore, an analysis was performed to document the fact that the par racks are capable of storing fuel more highly enriched than 3.1 percent.
The results of this analysis were incorporated into the UFSAR.
Safety Evaluation Summary The probability of occurrence of the accident / abnormal condition of dropping a fuel assembly on or around the spent fuel pool racks as evaluated in the UFSAR was not increased by this activity because no modification in fuel handling equipment or fuel storage equipment occurred. This change did not increase the probability of occurrence of an accident evaluated previously in the SAR. The accident / abnormal condition of dropping a fuel assembly on or around the spent fuel pool racks has been evaluated with the result being that the reactivity consequences are negligible. The analyses perfonned show that the requirement to maintain k-effective less than 0.95 is satisfied for normal and abnormal conditions. Any further type of bundle mispositioning or accident within the spent fuel pool has been analyzed and will have a k-efTective ofless than 0.95. This change did not increase the consequences of an accident evaluated previously in the SAR. Equipment important to safety was neither modified nor impacted by this change. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased, and the consequences of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. No new types of accidents were introduced. This change did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously. The l design of the spent fuel pool racks is to maintain the fuel configuration l substantially subcritical. This design is maintained intact by the l introduction of fuel of the type specified in the llOLTEC analysis into either design of the spent fuel pool rack at the DAEC. The margin of 131 l
l safety as defined in the basis for the Technical Specifications is not reduced. This margin is based on the requirement to keep the k-effective of fuel in the spent fuel pool racks less than 0.95. This UFSAR change was based on this requirement.
if SE 97-91 ASME Section IX Qualification Of Welders and Procedures Descriotion and Basis For Chance ASME,Section IX, contains requirements for the preparation and qualification of welding procedure specifications, and the qualification of welders and welding operators. ASME Section IX, QW-452.1, does not provide for the use of side bend test coupons for thickness' less than 3/8" ;
in thickness (0.375" thick) QW-452.1 requires face and root bend test- l specimens be used for test coupons thickness' less than 3/8". In preparation for Refueling Outages RFOl3 and RF014, thirty six DAEC welders were qualified by side bend testing NPS 2 Sch 160 pipe test :
coupons. NPS 2, Schedule 160 pipe is 0.344" tick, The rule of QW-452.1 is based upon a mechanical phenomenon combined i with an arbitrary 3/8" value for the width of the test specimen. A side bend test specimen is bent through the width of the specimen while a face or root specimen is bent through the thickness of the specimen. As the thickness of a side bend specimen falls below 3/8" while the width of the specimen remains at 3/8", there is a point at which the side bend test specimen becomes unstable and will buckle. The arbitrary dimension of 3/8" minimum thickness for side bend specimens was set as a built in safeguard for the Code user. This rule was intended to prevent the Cede user from preparing specimens which could possibly fall into an unstable condition where buckling would occur. Side bend tested specimens provide a superior qualification examination compared to the required face and root bend test specimens. The side bend test is a full-through-weld-thickness examination to determine the welders ability to deposit sound metal. The required face and root bend tests only examine the surface of the root and cap welds, but do not evaluate the full-through-weld-thickness. The DAEC Research Project has demonstrated that the 0.344" side bend test specimens are technically acceptable based on sound engineering principles, and the subsize side bend testing performed meets the intent of ASME,Section XI.
Safety Evaluation Summarv Pressure boundary welds were made in accordance with the design specifications. Welders qualifications were determined to be acceptable.
Therefore, welds made by welders qualified with subsize side bend 132
,~ . .- -- . -- -.-.-.-.---. .--_-.--.--- _
L l
( i l
l' L coupons do not increase the probability of an accident evaluated previously in the SAR, and there is no increase in the consequences of an q
accident evaluated previously in the SAR. This activity did not increase j
the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the j consequences of a malfunction of equipment important to safety evaluated
!- previously in the SAR. The welder qualification method used, meets or l exceeds the intent of ASME,Section IX. There is no possibility of L creating an accident of a different type than any evaluated previously in i
the SAR, and there is no possibility of creating an equipment malfunction I of a different type than any evaluated previously in the SAR. Welder L
qualification testing (side bend testing) is superior to root and face bend l tests. This method of welder qualification does not reduce the margin of l safety as defined in the basis for any Technical Specification.
l j_ SE 97-92 Condensate and Demineralized Water System P&lD Revision Description ani Basis For Chance l
l It was identified that the bearing cooling water tubing for the Condensate
{
L Service Jockey Pump was not shown on the Condensate and '
L Demineralized Water System P&ID. However, this same type of tubing was shown on the P&ID for the adjacent Condensate Service Pumps. 'A i field walkdown confirmed the existence and configuration of this tubing.
. The tubing was probably not shown on the P&lD because it was l considered skid mounted and had no in-line components. In addition, the l equipment name in the equipment database for the valve supplying bearing frame cooling to the Condensate Service Jockey Pump was revised >
for clarification. Also, the P&ID for the Condensate and Demineralized
- Water System was revised to show the Condensate Service Pumps L Discharge Cross-tie Valves as open. These valves had been shown closed on the P&ID since initial issue. The valve position reflected on the P&lD should be normal operation position, which is open.
j Safety Evaluation Summarv This activity only revised plant documentation and did not propose any physical alteration to the plant. These Condensate and Demineralized Water System components are all non-safety related. The changes made did not degrade the system or in any way impact system / component design function. One of the Condensate Service Pumps serves as a backup for the Condensate Service Jockey Pump. For this to happen, Condensate Service
. Discharge Valves and the Cross-tie Valves must be open. The changes j made to the P&ID reflect the design intent of these system components.
Therefore, this activity did not increase the probability of occurrence of an
{
i 133
l l
l accident evaluated previously in the SAR. Since the changes made to the P&ID reflect the design intent of these system components, there were no increased radiological consequences associated with this activity. This j activity did not increase the consequences of an accident evaluated previously in the SAR. No equipment important to safety is impacted by this activity since these non-safety related components perform as intended to support the refueling, normal service and emergency demands l of the system. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in j the SAR, and the consequences of a malfunction of equipment important l to safety evaluated previously in the SAR were not increased. No new failure modes were introduced by this activity since the affected j components perform their intended design functions. This activity did not l create the possibility of an accident of a different type than any evaluated previously in the SAR. This change only corrected drawing errors and l added additional information not previously shown. No new failure modes were created and no equipment important to safety was affected.
This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. No component mentioned in the Technical Specifications was affected. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
l SE 97-93 UFSAR Change For Discrepancies Between Blume Analysis and UFSAR Response Spectra Description and Basis For Chance L
! The UFSAR required update to clearly present the seismic requirements for design activities at the DAEC. The Seismic Category I structures were analyzed by J. A. Blume in 1973 for the seismic loads on the structures,
! and the development of the floor response spectra to be used for seismic design of piping and equipment. The following issues were addressed to l ensure that the J. A. Blume analyses were evaluated against the i appropriate criteria and the UFSAR update was complete:
- There were apparent discrepancies between the J. A. Blume analyses and the UFSAR with regard to the design seismic motion.
These apparent discrepancies were grouped into the following categories:
l I (1) J. A. Blume used only one input (time history) motion for building analysis and generation ofin-structure floor response l spectra, while the UFSAR presented six response spectra.
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1 I
l-p The J. A. Blume analyses use a factor of two-thirds to (2)
L determine the vertical accelerations, while the UFSAR stated that a factor of 80% shall be used for structures on bedrock, lean concrete fill or about 10 feet of soil over bedrock.
L e- The response spectra curves in the PSAR are not the same as the L
response spectra curves in the FSAR and UFSAR.
. Response spectra generated from the time history input used for the building analyses have only been compared to the PSAR response spectra for structures founded on rock or lean concrete fill.
The UFSAR and various design documents were revised to clearly present the seismic requirements for design activities at the DAEC. The seismic classification and qualification of structures, systems and components were not affected by this evaluation. I Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident cvaluated previously in the SAR. This evaluation merely assesses the adequacy of the J. A. Blume earthquake analyses of Seismic Category I structures with respect to the SAR criteria, and the seismic classification and qualification of structures, systems and components were not affected. I The effects on safety equipment due to a seismic event are adequately bounded by the J. A. Blume analyses. No new failure modes were identified or created by this evaluation. The probability that an earthquake ,
will occur was not afTected. There were no changes to plant hardware or !
operating procedures. This activity did not increase the consequences of an accident evaluated previously in the SAR. The J. A. Blume earthquake analyses of Seismic Category I structures are adequate for design activities and there are no physical changes. Thus, the seismic classification and qualification of structures, systems and components were not aff^cted.
The response to an accident and mitigation of the consequences were not affected by this evaluation. Therefore, the consequences of an accident evaluated previously in the SAR were not increased, and the probability of occurrence of a malfunction of equipment important to safety during a seismic event as evaluated previously in the SAR was not increased. The response to a malfunction of equipment important to safety and mitigation of the consequences were not affected by this evaluation; therefore, the l consequences of a malfunction of equipment important to safety evaluated
[ .
previously in the SAR were not increased, and the possibility of an
! accident of a different type than any evaluated previously in the SAR was
! not created. This activity did not create the possibility of a malfunction of L
I 135
equipment important to safety of a different type than any evaluated previously in the SAR. There was no effect on the reactor coolant pressure boundary integrity, ECCS performance or snubber operability, Therefore, the margin of safety as defined in the basis for any Technical Specification was not reduced.
SE 97-94 Residuallleat Removal System P&lDs and Core Spray System P&lD Revisians Descrintion and Basis For Chance Errors were identified in the depiction of various pipe lines on the Residual Heat Removal (RHR) System P& ids. All these pipe lines join the main line,"RHR Test Line." The branch lines were not shown on the P&lD in the same sequence as they are actually connected to the main line and as depicted on the isometric drawing. The error in the P&lDs existed from the time of the original plant startup. The RHR System P&lDs were revised to correct this error. The P&lDs for RIIR and Core Spray were also revised regarding drawing reference arrows and coordinates for continuation of the lines. The description given at the reference arrows for some of the lines was changed to indicate the origin and destination of the lines more accurately. Another drafting error on the RHR System P&lD involved a reducer not being shown between an 18 inch line and a 12 inch line upstream of the 'B' RHR Heat Exchanger Inlet Throttle Valve. This reducer is shown on the isometric drawing. This error had also existed on the P&lD from the time of the original plant startup, and has been corrected.
Safety Evaluation Summary This document change did not alter the design, state, or the designation of the pipe lines. Therefore, the system performance is not affected in any way by this change. This change did not increase the probability of occurrence of an accident evaluated previously in the SAR, and this activity did not increase the consequences of an accident evaluated previously in the SAR. While the change pertaining to the continuation arrows and cross reference to the other P&lDs enhances the drawings, the addition of the reducer was in accordance with the applicable codes and standards. As such the system's performance was not affected in any way by this change. Therefore, this activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This change had no impact on the functional aspect of the piping, but was only intended to make it easy to read the P&lD. There was no 136 l
i change to the system or equipment that could adversely afTect them. This is considered an enhancement to the P&lD because the change is fully in -
L accordance with the UFSAR. No other accidents other than those already evaluated in the SAR could be made possible by this change. Therefore, ,
this activity did not create the possibility of an accident of a different type l . than any evaluated previously in the SAR. No new failure modes other
!' than the ones evahiated previously in the SAR can be identified by this
! change. Therefore, this activity did not create the possibility of a '
malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. Technical Specification requirements for RHR and Core Spray were not affected by this change. The systems
! still function as intended. Therefore, this activity did not reduce the i margin of safety as defimed in the basis for any Technical Specification. ,
i l' . .
l SE 97-97 Auxiliary Heating System P&ID Revision L Description and Basis For Chance The vent lines from the casing of the two Turbine Building Heating
! System Hot Water Loop Recirculation Pumps were erroneously shown on
! the P&lD as draining into a funnel drain, located afler normally closed valves. However, in the plant, these vent lines are connected to drain lines. In addition, it was also found that the subject vent lines are 1/2" not ,
3/4"as was depicted in the P&lD This change revised the P&lD to show the vent pipe lines in accordance I with their actual configuration in the plant.
Safety Evaluation Summary The Turbine Building Heating System Hot Water Loop System is neither an initiator of an accident nor a mitigator. This change pertains to the rerouting and change in size of the normally closed vent and drain lines for ;
( the casing of the Hot Water Loop Recirculation Pumps. The overall '
l performance of the system was not affected in any way by this change.
l This activity did not increase the probability of occurrence of an accident
! evaluated previously in the SAR. The subject system is not required to l mitigate any accident. This change is non-safety related. This activity did l not increase the consequences of an accident evaluated previously in the
- i. SAR. This change did not cause any malfunction of the equipment in this L system. This change did not increase the probability of occurrence of a l malfunction of equipment important to safety evaluated previously in the SAR. Because the system is not required to support any safety related system, there are no radiological consequences due to an assumed
(
malftmetioning of the equipment due to this change. This activity did not L 137 l
1 l ,,
increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new accident was created by this change. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The components associated with this activity were not subjected to any new failure modes by this change. This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The Turbine Building HVAC, its subsystems and components are not included in the Technical Specifications. This change did not impact overall system performance. This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
SE 97-98 Offgas System P&ID Revision Description and Basis For Chance An isolation valve in the drain line to the Off Gas Equipment Drain Sump was not shown on the Offgas System P&lD. This drain line branches out from the sample line that goes from the Hydrogen Analyzers to the Offgas Sample Racks Return To Steam Jet Air Ejectors Valve. This isolation valve is normally closed. This valve was also shown without an equipment identification number on other design drawings as well as in the plant. In addition, a pressure indicator which was identified on the P&lD was shown on another design drawing as a pressure point.
The isolation valve as well as three other valves in the associated lines were installed for a modification in 1976, as part of rerouting the Hydrogen Analyzer discharge line to sample return line. While three of the above four valves were given plant identification numbers by a document change in 1992, the fourth valve was apparently missed. This valve, is now identified as V41-0241, Hydrogen Analyzer Sample Line Drain Isolation, and the design drawing has been corrected to properly identify the pressure indicator.
Safety Evaluation Summa _ry The subject document change involved showing the instrument drain line and the isolation valve for the Hydrogen Analyzers in the Off Gas System on the P&ID and assigning the labels on a design document. The piping involved is non-safety related. The Off Gas System is not required to prevent or mitigate an accident. The Hydrogen Analyzers are not impacted by this change because the condensate drain line and the isolation valve enhance the performance of the analyzers. The system still functions as intended. Therefore, this activity did not increase the 138 l
l probability of occurrence of an accident evaluated previously in the SAR, l and it did not increase the consequences of an accident evaluated
) previously in the SAR. An assumed malfunction of the equipment will I not impact the system in any way because the valve is normally closed and it fails closed. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR has not been increased. An operational failure or a component failure of this j system does not result in a site boundary dose that is an appreciable fraction of 10 CFR 100 doses. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR and it did not create the possibility of an accident of a different type than any evaluated previously in the SAR. This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The proposed change has no impact on the surveillance tests for the equipment associated with this change. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-101 Revisions To UFSAR Chapters 1 And 11 Description and Basis For Chance Changes have been made to Chapters 1 and 11 of the UFSAR to keep it up to date with current practices relative to Radwaste processing and Radwaste System operation. The changes included the following:
. Typographical error.
. Clarification of requirements.
Elimination ofinconsistencies with other UFSAR Sections.
. Deletion of obsolete references to vendor equipment or processes no longer in use at the DAEC.
. Add reference to basic regulatory requirements such as site disposal requirements and 10 CFR 61.
All of the changes were made to reflect current practices and configurations which were previously evaluated via the design control process as applicable. Some clarification have been made to reflect current commitments to regulations contained elsewhere in the SAR and these are being incorporated into the text to reduce confusion.
139
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The changes did not affect the inputs into any accident analysis performed for the DAEC. Therefore, this UFSAR change did not increase the probability of occurrence of any accident evaluated previously in the SAR and the consequences of any accident evaluated previously in the SAR was not increased. The afTected equipment is not classified as nuclear safety related. The changes did not create any situation where equipment important to safety would be compromised. The Radwaste System is classified Quality Level 2, which invokes quality assurance requirements for the pressure boundary components of the Radwaste System only. None of the changes affected the pressure boundary of the Radwaste System. Therefore, this change did not affect the probability of occurrence of a malfunction of any equipment which has a safety function. The changes did not result in any increased radiological exposure to plant personnel or the public. This change did not increase the consequences of a malfunction of any equipment important to safety evaluated previously in the SAR and the possibility of an accident of a different type than any evaluated previously in the SAR was not created. This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The bases for Technical Specifications associated with the Radwaste System are not affected by this change. The margin of ,
safety is defined for this system by bounding criteria which are contained ;
in the SAR and was not affected by these changes. This activity did not !
reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-107 Technical Requirements Manual Revision For Allowance Of Manual l
Rod Block Description and Basis For Chance Technical Specification Amendment 223 to convert DAEC Technical Specifications to the Improved Technical Specifications (ITS) per NUREG-1433 included relocation of certain Technical Specification requirements to the Technical Requirements Manual (TRM). The TRM is a licensee-controlled document incorporated by reference into the UFS AR and subject to revision per 10 CFR 50.59.
This change revised the TRM to relax the actions for inoperable rod block channels to allow a manual rod block to be inserted as an alternative to placing the inoperable channels in the tripped condition. The resulting requirements continue to ensure that process variables and SSCs are j 1
140 i
maintained consistent with the conditions and assumptions of the DBA and transient analyses. The changes did not involve any design changes to the plant and are consistent with current plant operations.
Safety Evaluation Summary This change to the TRM was to allow a Reactor Manual Control System (RMCS) rod block to be inserted as an alternative action to placing an inoperable required channel in trip. As such, the changes did not result in any hardware changes or permanent physical alteration of the plant and they had no impact on the design or function of any SSC. The SSC process variables, characteristics, and functional performance were maintained consistent with the event initiator and initial condition assumptions for the DBA and transient analyses. Therefore, the likelihood of event initiation remained as previously analyzed and there was no increase in the probability of occurrence of an accident evaluated previously in the SAR. Because the requirements did not satisfy the 10 CFR 50.36 screening criteria for inclusion in the Technical Specification, they did not impact, either directly or indirectly, any event analyzed in the SAR. The changes did not change that conclusion as they did not introduce any new modes of equipment operation (i.e., new Required Actions are within the equipment's performance capability and are compatible with existing plant operations.) Accordingly, the changes did not increase the consequences of an accident evaluated previously in the SAR, and there was no increase in the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. There was no increase in the radiological consequences of an accident. The changes did not increase the consequences of a malfunction l of equipment important to safety evaluated previously in the SAR. No new credible accidents were introduced. Accordingly, the changes did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and they did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The changes did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-109 More Restrictive Requirements Added To Technical Requirements Manual Description and Basis For Chance Technical Specification Amendment 223 to convert DAEC Technical Specifications to the Improved Technical Specifications (ITS) per NUREG-1433 included relocation of certain Technical Specification requirements to the Technical Requirements Manual (TRM). The TRM is l
141
i I
a licensee-controlled document incorporated by reference into the UFSAR and subject to revision per 10 CFR 50.59.
These changes to the TRM involved additions and more restrictive modifications to the verbatim relocated Technical Specifiation LCOs, Actions, and Surveillance Requirements, including reductions in Action Statement Completion Times and increases in Surveillance Frequencies.
l The resulting requirements continue to ensure that process variables and
- l. SSCs are maintained consistent with the conditions and assumptions of the l DBA and transient analyses.
l l Safety Evaluation Summary I These changes did not result in any hardware changes or permanent physical alteration of the plant and the changes had no impact on the design or function of any SSC. The SSC process variables, characteristics, l and functional performance were maintained consistent with the event b initiator and initial condition assumptions for the DBA and transient i analyses. Therefore, the likelihood of event initiation remained as previously analyzed and there was no increase in the probability of occurrence of an accident evaluated previously in the SAR. Because the L requirements did not satisfy the 10 CFR 50.36 screening criteria for inclusion in Technical Specifications, they did not impact, either directly or indirectly, any event analyzed in the SAR. The changes in these requirements did not change that conclusion as they did not introduce any new modes of equipment operation (i.e., new Surveillances and Required Actions are within the equipment's performance capability and the test Frequencies and Completion Times are compatible with existing plant operations.) The changes did not increase the consequences of an accident L evaluated previously in the SAR, and there was no increase in the '
l probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. There was no increase in the r radiological consequences of an accident. Therefore, the changes did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new credible accidents were introduced. Accordingly, the changes did not create the possibility of an l accident of a different type than any evaluated previously in the SAR, and
- the changes did not result in any changes to plant equipment or operation which could introduce a new equipment failure mode. The changes did
- not create the possibility of a malfunction of equipment important to safety L of a different type than any evaluated previously in the SAR, and they did not reduce the margin of safety as defined in the basis for any Technical Specification.
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SE 97-111 Technical Requirements Manual Revision To Relax Calibration Frequency For Source Range and Intermediate Range Monitoring l l
Description and Basis For Chance i l
Technical Specification Amendment 223 to convert the DAEC Technical l Specifications to the Improved Technical Specifications (ITS) per )
NUREG-1433 included relocation of certain Technical Specification l requirements to the Technical Requirements Manual (TRM). The TRM is a licensee-controlled document incorporated by reference into the UFSAR and is subject to revision per 10 CFR 50.59.
This change revised the TRM to relax the calibration frequency on the Source Range Monitoring (SRM) and Intermediate Range Monitoring l
(IRM)"Not Full In" rod block Functions from 18 to 24 months. !
Safety Evaluation Summary This change did not result in any hardware changes or permanent physical alteration of the plant. The changes had no impact on the design or function of any SSC. The SSC process variables, characteristics, and functional performance were maintained consistent with the event initiator and initial condition assumptions for the DBA and transient analyses. The likelihood of event initiation remains as previously analyzed and there was no increase in the probability of occurrence of an accident evaluated previously in the SAR. Because the requirements did not satisfy the 10 CFR 50.36 screening criteria for inclusion in the Technical Specifications, l they did not impact, either directly or indirectly, any event analyzed in the SAR. The changes in these requirements did not change that conclusion as they did not introduce any new modes of equipment operation (i.e., new Surveillance Frequency is within the equipment's performance capability and are compatible with existing plant operations.) Accordingly, the l changes did not increase the consequences of an accident evaluated previously in the SAR. The purpose of an instrument calibration is to
- compensate for any " drift" in the instrument setpoint over time. Because this trip function is performed by a limit switch, it does not exhibit time-dependent drift. Therefore, the changes did not adversely affect any plant equipment and there is no increase in the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This trip function is not part of the primary success path of any analyzed event. Thus, there was no increase in the radiological consequences of an accident. The changes did not increase the l consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new credible accidents were introduced. The changes did not create the possibility of an accident of a different type 143 i
l r
than any evaluated previously in the SAR. The changes did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR, and the changes did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-112 Drawing Revisions For Low Level Radwaste Processing and Storage Facility (LLRWPSF) Room Layout / Room Designation Changes Description and Basis For Chance Various drawings were required to be revised to incorporate changes that had taken place in the past. Changes were not shown on the drawings regarding room / wall / door configurations for the LLRWPSF first and second floors. The changes were associated with addition of new doors and walls, installing new openings in non-masonry walls that are not load bearing, and changes in designation of certain non-processing areas in the LLRPSF.
Safety Evaluation Summary The Low Level Radwaste Processing and Storage Facility is not an initiator of any accident because it is not associated with any safety related equipment. The document changes, such as addition of new doors and walls, installing new openings in non-masonry walls that are not load bearing, and changes in designation of non-processing areas in LLRPSF cannot create any accident. The changes are not a tire hazard. This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The document changes are Quality Level IV. No equipment associated with the changes are required to mitigate an accident. There is no impact on 10 CFR 20 limits. This activity did not increase the consequences of an accident evaluated previously in the SAR. These changes cannot cause any malfunction of equipment. The LLRPSF does not house any safety related equipment.
This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased. The possibility of an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The LLRPSF and the equipment housed by it are not included in the Technical Specifications or surveillance requirements. This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
144
SE 97-114 Revision To UFSAR Chapter " Isolation Valve Closing Devices and Circuits" Description and Basis For Chance The following changes were mate to G. taole contained in UI'3AR Chapter " Isolation Valve Closing Devices and Circuits":
Changed "RHR system shutdown cooling discharge isolation valves" to "RHR system shutdown cooling inboard discharge isolation valves". This clarified the specific valves which have an isolatio 1 function which should be identified in the UFSAR. The "RHR system shutdown cooling outboard discharge isolation valves" were removed from the list contained in this Chapter of the UFSAR. These valves are located outside of the Drywell on the Low Pressure Coolant Injection (LPCI) inject lines. These lines penetrate Primary Containment, contain a check valve and communicate with the Reactor Recirculation loop. The inboard valves receive a Group IV isolation signal from the Primary Containment Isolation System logic via the Residual Heat Removal System (RHR) logic only when the RIIR System is in the Shutdown Cooling Mode of operation. They also receive open or close signals from the LPCI-Loop-Select logic. The outboard valves do not receive a Group IV isolation signal, but do open or close based on signals from the LPCI-Loop-Select logic.
. Removed " Core spray injection isolation valves", from this section of the UFSAR. These valves are located outside of the Drywell on the Core Spray inject lines. These lines penetrate Primary Containment, contain a check valve and communicate directly with the Reactor Vessel. Both the inboard and outboard valves do not receive any automatic isolation signals. They receive open signals from the initiation logic.
- Changed " Design closure times" to " Design maximum closure times."
The UFSAR discusses " Isolation" valves. Maximum closure time is the critical parameter for isolation.
Safety Evaluation Summary Since the subject valves meet all design basis and licensing basis criteria i contained in the SAR, the probability of occurrence of an accident is not increased. The isolation capability of the subject valves does not affect the probability of occurrence of an accident. The consequences of an accident previously evaluated in the SAR are not increased. The subject valves can l
145 I
. - - - - - - . =- _ _.
l influence the consequences of an accident due to their isolation functions; however, they will isolate as required by the design basis and the licensing basis. The probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR is not increased. The I
subject valves are equipment important to safety due to their isolation functions; however, they will isolate as required. The consequences of a malfunction of equipment important to safety previously evaluated in the l SAR are not increased. The possibility for an accident of a different type than any evaluated previously in the SAR is not created. This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The margin of safety was not affected because the subject valves meet all design basis and licensing basis criteria contained in the SAR.
SE 97-115 Technical Requirements Manual Revision To Extend Calibration Frequency I For Safety Relief Valve and Safety Valve Tailpipe Pressure Switches Description and Basis For Chance l
Technical Specification Amendment 223 to convert the DAEC Technical i Specifications to the Improved Technical Specifications (ITS) per l NUREG-1433 included relocation of certain Technical Specification
! requirements to the Technical Requirements Manual (TRM). The TRM is i
a licensee-controlled document incorporated by reference into the UFSAR and subject to revision per 10 CFR 50.59. I This specific change relaxed the Technical Specification calibration l frequency on the Safety Relief Valve (SRV) and Safety Valve (SV) ;
l tailpipe pressure switches from 18 to 24 months. The resulting i
- requirements continue to ensure that process variables and SSCs are L maintained consistent with the conditions and assumptions of the DBA !
and transient analyses.
Safety Evaluation Summary These changes did not result in any hardware changes or pemianent l physical alteration of the plant and the changes had no impact on the
! design or function of any SSC. The SSC process variables, characteristics, and functional perfonnance were maintained consistent with the event initiator and initial condition assumptions for the DB A and transient analyses. Therefore, the likelihood of event initiation remains as i previously analyzed and there was no increase in the probability of occurrence of an accident evaluated previously in the SAR. Because the
- requirements did not satisfy the 10 CFR 50.36 screening criteria for inclusion in the Technical Specification they did not impact, either directly 146
t or indirectly, any event analyzed in the SAR. These changes did not introduce any new modes of equipment operation (i.e., new Surveillance Frequency is within the equipment's perfonnance capability and are compatible with existing plant operations.) Accordingly, the changes did not increase the consequences of an accident evaluated previously in the SAR. The purpose of an instrument calibration is to compensate for any
" drift"in the instrument setpoint over time. The extension in calibration frequency affects the expected drift in the pressure switch setpoints.
However, the increase in drift was accounted for in the setpoint calculations for the LLS function. Therefore, the changes did not adversely affect any plant equipment and there was no increase in the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The indication function of these pressure switches is not part of the primary success path of any analyzed event. Thus, there was no increase in the radiological consequences of an accident. The changes did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new credible accidents were introduced. Accordingly, the changes did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The changes did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR, and the changes did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 97-117 . Residual Heat Removal System P&ID Revision Descriotion and Basis For Chance .
The local handswitch for the RHR/ Core Spray Fill Pump did not have a label or identification number. Plant documents needed to be updated to reflect the new number. The start /stop handswitch for the RHR/ Core Spray Fill Pump is now shown on the P&lD. The schematic was also revised to better reflect the configuration of the installed switch. The function .of the RHR/ Core Spray Fill Pump has not changed.
Safety Evaluation Summary This change did not increase the probability of occurrence of an accident evaluated in the SAR. The RHR/ Core Spray Fill Pump still operates as intended. Only the start /stop switch was added to the P&lD. This activity did not increase the consequences of an accident evaluated previously in the SAR. Monitoring of the discharge lines was not affected. This change did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The consequences of a malfunction of equipment important to safety evaluated 147
l
- _ previously in the SAR was not increased by this change. The possibility l L of creating a new accident was not created. The fill pump still operates as
. intended. The possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the SAR was not created. The margin of safety was not reduced. The manual switch is the L only control for the pump and is required for its operation. Monitoring of i the discharge lines as described in the Technical Specifications was not affected.
. SE 97-120 Control Building Air Conditioning System P&lD Revision
' Descriotion and Basis For Channe None of the components associated with the Control Building Exhaust Isolation Dampers were labeled in the field. A flow control valve, two
! pressure indicators and a pressure control valve (PCV) were identified that were not shown on the P&ID. The flow control valve and the two F pressure indicators were not shown because they are vendor supplied components furnished with the pneumatic operator. The PCV was not addressed on any documents or drawings uet could be located. This valve ,
is installed in the 100 psig Instrument Air supply to the operator and I regulates control air pressure, providing the 80 psig required to open the L damper. This system component was apparently added in the mid 1980s due to excessive damper noise with 100 psig control air. Although no documentation could be located for the original installation, a new L pressure regulator of the same type was recently installed due to air leaking from the old one. A Commercial Grade Dedication / Upgrade Evaluation was performed on this new PCV, and reasonable assurance was provided that the component would perform its intended design function.
Equipment identification number, PCV-6107Al, was assigned to this pressure regulator and the P&ID was revised to show it.
l Safety Evaluation Summary i
The Control Building Isolation Control /HVAC System is not an accident
[ initiator, but is required to provide Control Room environmental control in L
the event of accidents. This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The Control Building Isolation Control /HVAC System functions to mitigate the l consequences of an accident in order to protect Control Building L inhabitants in emergency situations. There are no increased radiological consequences associated with this activity. The addition of a pressure
, regulator in the 100 psig air supply to a damper operator does not change
{ the design, function, or method of performing the function of Control Building Exhaust Isolation Damper. Damper operators are designed for I
l 148 l
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I-I air opening, spring closing and are capable of opening dampers in eight to ten seconds and closing dampers in five seconds. Air pressure of 80 psig j minimum and 110 psig maximum is required to open the damper. The PCV is set at 80 psig, providing the required air pressure to the operator.
Failure of this pressure regulator will result in either full air pressure (100 psig) to the operator (acceptable since operator is designed for 110 psig max.) or lack / loss of required air pressure (80 psig min.), in which case the damper fails closed as designed providing maximum Control Building isolation. In addition, isolation dampers are mounted in pairs and are designed such that either may close or open fully without regard for the l position of the other. Therefore, this activity did not increase the consequences of an accident evaluated previously in the SAR, and it did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. This change does not prohibit the system components from performing as designed to mitigate the consequences of an accident since the required air pressure is still available to open the damper and the damper still fails closed on loss / lack of required air pressure as originally designed. This activity did not i increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and the possibility of an accident
! l of a different type than any evaluated previously in the SAR was not l created. No new failure modes were created by this activity. This activity l l did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The LCOs and Surveillance Requirements associated with Main Control Room ventilation are included in DAEC Technical Specifications. The changes addressed by this activity did not affect the ability of the Control Building ;
Isolation Control /HVAC System to maintain positive Control Room pressure as required by the Technical Specifications. This activity did not reduce the margin of safety as defined in the basis for any Technical Qecification.
SE 97-121 UFSAR Change For Containment Penetration Seal Material Descrintion and Basis For Chance The seals for several containment penetrations are shown in UFSAR Figures. The seal material was obsolete and the UFSAR Figures: . quired change to allow the use of an equivalent material. This change allowed the use of alternate materials on the seals for the CRD removal hatch and L the male stabilizer assembly inspection port. The function of the seals are to provide part of the primary containment pressure boundary. The
- containment specification requires that the seals be either infiatable or j solid (or a combination) and fabricated of silicone rubber in continuous f rings. The original seals were fabricated from Parker silicone compound I
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- . S418-6 with a durometer rating of 60. The replacement material is Parker l silicone compound S613-60 with a durometer rating of 60. Both seals are manufactured to Aerospace Material Specification (AMS) 3303 (silicone, general purpose). Both materials are designed for continuous service in the temperature range of-60 to 450 F, for air and gas service, and for static service. These requirements are appropriate for the identified uses.
Therefore, the design pressure and temperature ratings are unchanged as
, the materials are technically equivalent.- The ability of the penetration to j function (seal) is unchanged. The change in material did not impact the i
ability to test the penetration. The passive primary contaimnent integrity can be obtained by either seal material. Therefore, this change was acceptable for this function.
Safety Evaluation Summarv This activity did not increase the probability of occurrence of an accident l evaluated previously in the SAR. The seal performs a passive safety function. The material change did not affect the ability to seal. Primary containment does not cause accidents as previously analyzed in the SAR.
This activity did not increase the consequences of an accident evaluated previously in the SAR. The probability of a malfunction is not changed, and the consequences of a malfunction of equipment important to safety l~
evaluated previously in the SAR were not increased. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the l SAR was not created. The function and design of the seals are similar in all aspects but material. The materials were evaluated to be technically equivalent. This change did not reduce the margin of safety as defined in the basis for any Technical Specification. The change in seal material did not affect the margin of safety. The primary containment integrity margin as defined in the bases to the Technical Specifications has not been
! reduced by this change.
i SE 97-122 UFSAR Change For Cooling Tower Basin Low Level Alarm Description and Basis For Chance UFSAR Section 10.4.5.3 discusses flooding of the Turbine Building Basement that could be caused by a ruptured Circulating Water System expansionjoint. This UFSAR change deleted the reference to a six inch
( Cooling Tower Basin level drop causing an alarm in the Control Room.
! Instead, the UFSAR was revised to state,"An 18 inch drop below normal
[ operating level, will alarm in the Control Room alerting the operator of a
- failure in the River Water Supply system or a leak in the Circulating Water
]
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I System". This activity did not change the existing low basin level alarm set point or any other existing set point. It reflects a low level setpoint change made in 1987 to reduce the number of Control Room alarms.
Changing the Cooling Tower basin low level alarm did not impact the plants response to an unlikely Circulating Water line rupture in the Turbine Building. Adequate time and alarms are provided to trip the Circulating Water pumps before any damage occurs. The plant will automatically scram due to turbine trip from low condenser vacuum. The safe shutdown capability of the plant is not threatened in the unlikely event of a Turbine Building Flood scenario. This UFSAR change did not
, add or delete any automatic or manual function. It did not introduce an i unwanted or previously unreviewed system interaction. No seismic, enviromnental qualification, or quality group classification was affected.
Safety Evaluation Summary l
[ This activity did not increase the probability of occurrence of an accident ;
reflect a reduction in the Cooling Tower basin low level alarm. This l change did not impact any of the accidents previously evaluated in the l l UFSAR. Lowering the low level alarm set point improved system reliability by activating the alarm only when an actual low level condition i exists. Plant operators still have sufficient time to react and initiate I corrective actions per existing plant procedures. Lowering the Cooling
, Tower basin low level alarm setpoint did not increase the consequences of i
an accident previously evaluated in the SAR. The cooling tower low level alarm is one of a series of alarm features used to indicate problems with the River Water Supply System or a leak in the Circulating Water System.
This change did not cause plant operators to react in a different way than documented in plant procedures. This change did not impact how the .
' plant responds to a loss of circulating water pit level or makeup water to l
the circulating water pit. The Circulating Water System and Cooling l Towers are not safety related. The cooling tower basin low level alar.n j setpoint does not have any impact on any safety related equipment r.
I '
DAEC. Therefore, this activity did not increase the probability r,f occurrence of a malfunction of equipment important to safety evaluated l previously in the SAR, and it did not increase the consequences of a
! . malfunction of equipment important to safety evaluated previously in the SAR. The low level alarm setpoint was lowered to remove spurious alarms. No other alarms, indications or procedures have been changed. l This change did not impact plant response or other alarms received during
. a Circulating Water System leak or actual tower basin low level condition
! during plant operation. Therefore, this activity did not create the l possibility of an accident of a different type than any evaluated previously l in the SAR. This activity did not create the possibility of a malfunction of 15)
l l
- j. equipment important to safety of a different type than any evaluated t previously in the S AR. Lowering the cooling tower basin low level alarm I did not impact the plant's margin of safety. Therefore, this activity did not I reduce the margin of safety as defined in the basis for any Technical l Specification. j l SE 97-123 Well Water System P&lD Revision Description and Basis For Chance i
The Well Water System P&lD showed a valve in the bypass line for a cooling water bypass control valve as normally closed. However, per the 1 valve lineup in the system Operating Instruction this valve nomially is in ,
- throttled open position. Therefore, the drawing wcs revised to show the '
symbol correctly. A historical document review indicated that the subject
! valve was shown incorrectly, as normally closed on the P&lD, at the time of the original plant startup. This was considered a drafting error. I j Safetv Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. This change only revised the plant documentation to show the correct normal position of a non-safety related valve. The Well Water System is not safety related. The change was in accordance with the Operating Instruction for this system. The Well Water System is not an accident initiator or mitigator. This change did not impact the 10 CFR 100 limits in any way because the NSOA does not identify the Well Water System as a system required to perform any safety i function. Therefore, this activity did not increase the consequences of an accident evaluated previously in the SAR. This change to the position of the non-safety related valve did not cause any malfunction of the valve because the valve is designed to envelop this condition and it meets the plant design codes and standards. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. No SSC was affected by this l change. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. This activity did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The Well Water System is not addressed in the Technical Specifications. Per the UFSAR the Well
, Water System requires no testing. This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
152
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Descrintion and Basis For Chance l
, The Duane Amold reactor pressure vessel (RPV) was originally designed
{. for normal and abnormal cyclic loading conditions expected to occur, or l postulated to occur, over its 40 year design life. The Reactor Vessel .
Transient Design Table listed 18 separate events and the design number of occurrences of each event for a 40 year design life. This table formed the J
basis for the original fatigue analysis of the RPV.
-A review of plant records and operator logs was recently completed to determine the actual number of design cyclic events that have occurred i
through fuel cycle 14 (Fall 1996). As expected, the review indicated that some events are occurring at a higher rate than anticipated, while other
, events are occurring at a lower rate. No cyclic events were found to have i
exceeded the design number of occurrences at this time. In addition, it was determined that some events listed in the UFSAR Table did not cause l fatigue usage to the RPV (e.g., daily or weekly reductions in power) while i there were other events not listed in the Table which did cause fatigue usage (e.g., CRD isolation).
i Accordingly, an updated transient design life for the Duane Arnold RPV i has been developed based on past operating history through fuel cycle 14
- (approximately 22 years). Significant pressure and thermal transients j which affect the fatigue usage of the RPV, the number of cyclic events that l- have occurred through fuel cycle 14, the expected number of events for a 40 year life, and the design number of events have been identified.
l The revised Table lists the significant pressure and thermal transients 1
which affect the fatigue usage of the Duane Arnold RPV and the design number of occurrences of each cyclic event. The revised Table represents
- i. the design cyclic events for which the RPV has now been analyzed based j on a new analysis. The wording in the Reactor Vessel Section of the d
UFSAR was also changed to be consistent with Improved Technical l
Specifications and current procedures.
The following changes were made to the Table:
- The following events were deleted since they do not contribute to fatigue usage to the RPV: Daily reduction to 75% power; weekly reduction to 50% power; and control rod worth test (referred to as control rod sequence exchange at DAEC). For each of these events, 153
l i
the RPV pressure and temperature does not change, and therefore, there is no cyclic stress. '
The following events were added since they result in non-negligible fatigue usage to the RPV: Aborted startups (has essentially the same effect as a full startup/ shutdown cycle); CRD isolation (has the same effect as a scram); and Single CRD scram (has the same effect as a scram),
t
. The design number of startup/ shutdown events was increased from 120 to 160. The number of events was increased because these events are occurring at a rate slightly higher than originally estimated.
l l Only the total number of scrams are counted. Scrams are not counted ;
by the initiating event. This is because all scrams produce essentially 1 the same fatigue usage to the RPV.
l l The effect of the updated RPV transient design life on the original RPV l
fatigue analyses was assessed. Fatigue usage factors were recalculated for the cyclic events and design number of cycles provided in the revised Table. In the assessment, the design number of aborted startups (30) was added to the design number of startup/ shutdown cycles (160) and the RPV was evaluated for a total of 190 startup/ shutdown cycles. The CRD penetrations were analyzed for an additional 50 CRD isolations and 10 single CRD scrams.
The new fatigue usage factor for each component of the RPV analyzed in the original fatigue analysis was tabulated along with the fatigue usage factor from the original analysis. All fatigue usage factors based on the cyclic events provided in the revised Table were calculated to be s 1.0.
The Duane Arnold RPV meets the ASME Code fatigue requirenuts for the updated transient design life.
Safety Evaluation Summary The updated reactor vessel transient design meets the design criteria for the reactor in that the applicable design criteria as established by ASME Section III are still met (fatigue usage factors are all s 1.0). Therefore, the probability of occurrence of an accident evaluated previously in the SAR has not been increased. The consequences of an accident that can be affected by the updated reactor vessel transient design analysis as defined
- i. in the NSOA is an unacceptable result of a system stress in excess of that allowed by industry codes. Since the updated reactor vessel transient design still considers these accidents and since the allowable fatigue usage 154 t
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factor still meets the applicable industry code, the consequences of an accident is not affected by the updated reactor vessel transient design.
Therefore, the consequences of an accident evaluated previously in the SAR have not been increased. The RPV serves as a passive pressure boundary for the Reactor Coolant System. All fatigue usage factors for the updated transient design are less than ASME Code allowables.
Therefore, the probability of occurrence of a malfunction ofequipment important to safety evaluated previously in the SAR has not been increased. No physical changes were made to plant equipment. No new failure modes or accidents are postulated by this change. Therefore, the consequences of a malfunction of equipment important to safety evaluated previously in the SAR has not been increased. No changes were made to design conditions or operating procedures. The possibility of an accident of a different type than any evaluated previously in the SAR was not increased. The RPV serves as a passive pressure boundary for the Reactor Coolant System. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not increased. Technical Specifications stated that the stresses in the RPV meet the stress intensity and fatigue limits of ASME Section III. The RPV has been reanalyzed for the new design transient events given in the revised UFSAR. The calculated fatigue usage factor for all RPV components analyzed in the original fatigue analysis is less than 1.0, which is the allowable fatigue usage factor in the original analyses. Thus, the margin of safety, which is defined as the range of a physical parameter (e.g., the fatigue usage factor) above the acceptance limit (1.0 in this case) has not been reduced.
SE 98-04 UFSAR Change For Refueling Interlocks Description and Basis For Chance This UFSAR Change modified the description of the Refuel Interlock #3 features. This revised description of features allows a better understanding of how to functionally test the interlock as required by Technical Specifications.
Previously, Refuel Interlock #3 was tested by demonstrating that refuel bridge travel was interrupted if more than one rod was withdrawn.
Performance of this test was a three step process: 1) a rod was withdrawn with the Mode Switch in Refuel,2) then the Mode Switch was turned to Startup which allowed pulling a second rod, and 3) then the Mode Switch was turned back to Refuel. However, the UFSAR was meant to describe the Refueling Interlock Effectiveness, and not the required method of testing.
i 155
l i
L The description of Refuel Interlock #3 was broken into two parts and is
' described as follows:
1 3.a With the Mode Switch in REFUEL, refueling platform travel
(- towards the core is prevented when the following two conditions
.l t exist concurrently:
1 1
- 1. One rod withdrawn, deselected, and then the same or J. alternate rod selected with the Mode Switch remaining .
continuously in REFUEL.
j
- 2. The refueling platform position switch is open (platform near or over the core). !
3.b - With the Mode Switch in SHUTDOWN or RUN, refueling l platform travel towards the core is prevented when the following )
l two conditions exist concurrently: .
- 1. One or more rods withdrawn.
- 2. The refueling platform position switch is open (platform l near or over the core). j l
With the Mode Switch in the Refuel position, Interlock #3 can be demonstrated to satisfy the scenario of the more than one rod out condition blocking the motion of the Refuel Bridge, when near the core, without pulling a second rod. By showing that movement of either the Rod Select
, Switch or the Mode Switch will disable the pemaissive for Refuel Bridge j travel if near the core, it demonstrates that the scenario of changing the l Mode Switch to withdraw more rods and then being able to move the !
Refuel Bridge is unnecessary. I 1
l- The revised testing simply deselects a selected rod and selects a second l
!- rod. This action results in an interlock action of blocking the motion of l the refueling platform if near or over the core, it is the same action that I l previously was tested by withdrawing a second rod.
l Safety Evaluation Summary The new description of Refuel Interlock #3 refines the description of features so that it is better understood for testing. The resulting revised interlock testing will be as effective but keep the plant configuration in a ,
more conservative lineup and in accordance with the current Technical Specifications. No hardware, logic, or setpoint changes were made, and therefore no physical changes were made to the facility. Interlock #3
- protects against more than one rod out when moving the Refuel Bridge 156
- e e v-7q u ws- yw y- w-y y e v gyy - ==, -
1 near the core, which guards against the Design Bases requirement that inadvertent criticality is prevented during fuel handling operations. The revised description details the conditions which could result in a second rod not full in, and the Mode Switch in Refuel. Testing these conditions ;
confirms that Interlock #3 will guard against inadvertent criticality.
Therefore, the probability of occurrence of an accident evaluated previously in the S AR was not increased. This change did not increase the consequences ofinadvertent criticality, and therefore did not increase the consequences of an accident evaluated previously in the SAR. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased, and the consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased. The possibility of an accident of a different type than any evaluated previously in the SAR was not created. This change did not create the possibility of a malfunction of ,
equipment important to safety of a different type than any evaluated previously in the SAR. The margin of safety was not reduced because the revised language of the Refuel Interlock #3 describes actions that would be necessary to accomplish a simulated condition of more than one rod out ,
in the Refuel Mode. By testing these actions, we test the scenario of more than one rod out. The margin of safety as defmed in the basis for any ,
Technical Specification is not reduced.
SE 98-05 Fire Plan Revision 34 Description and Basis For Chance l J
Revision 34 to the DAEC Fire Plan involved deleting the italicized text in l '
the following excerpt from the Fire Plan,"The implementation of the Quality Assurance program assures that the Fire Protection Systems and Features listed under " Operability Requirement" and those portions ofthe fire protection systems which protect safety related structures, systems or components will meet the Quality Assurance requirements of Reference 3.11 Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance." 10 CFR 50 Appendix R, Section I, Introduction and Scope states, "...the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to ;
mitigate the consequences of design basis accidents... Redundant systems i used to mitigate the consequences of other design basis accidents but not necessary for safe shutdown may be lost to a single exposure fire..."
While those systems that contribute to " defense in depth" are important to !
overall fire safety, elevating the systems' level of maintenance and reliability out of balance with the systems' ability to ensure safe shutdown is an inappropriate use of plant resources.10 CFR 50 Appendix R i
157
contains specific requirements for Fire Protection equipment required to ensure safe shutdown within sections III.0, III.J, Ill.L and III.O. In addition, specific exemptions from the requirements ofIII.G, III.J, III.L and 111.0 specify Fire Protection equipment required to ensure safe shutdown.' This change involved the removal of Licensing commitments to specific Fire Protection equipment which are not required to satisfy the requirements of 10 CFR 50 Appendix R or approved exemptions from the requirements of 10 CFR 50 Appendix R. To comply with 10 CFR 50 Appendix R sections Ill.G, III.J, III.L and Ill.0, a systematic evaluation.of fire areas was performed using the procedures detailed in the Fire Hazards
' Analysis. These compliance assessments are documented in the Master Document List (MDL) calculations and detail the plant equipment, manual operator actions, exemptions and Fire Protection Systems required to ensure safe shutdown of the plant._ The Fire Protection Systems identified
~
by the compliance assessment calculations or in approved exemptions are listed in the Fire Plan and have specific Operability and Surveillance requirements. The remaining Fire Protection Systems provide " defense in depth", however they do not provide protection of equipment required to safely shutdown the plant. Removal of Licensing commitments for these systems did not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Maintenance and testing of the systems are either specified by the property protection insurer or at a level equivalent to standard industrial Fire Protection Systems. Modifications to the Fire Protection Systems are controlled by the DAEC design change process specified in Administrative Control Procedures and any changes i are evaluated to ensure the systems' contribution to overall defense in depth is not reduced or is compensated by an increase in protection.
Safety Evaluation Summary The removal of these commitments did not impact any of these analyses including their entry conditions, bases or assumptions. The changes in licensee commitments to include only equipment required for compliance with 10 CFR 50 Appendix R does not change the probability of an accident as fire is not an entry condition, basis or assumption for any of these events. The changes did not increase the probability of any of the following events occurring: a fire, inadvertent actuation of a Suppression System or loss of essential llVAC or other plant equipment credited in the DAEC safe shutdown analysis. These changes did not adversely affect the plant's ability to safely shutdown in the event of a fire. Therefore, the removal of these commitments for the Fire Protection Systems did not increase the probability of occurrence of an accident evaluated previously in the SAR. Since fire is not one of the accidents analyzed in the UFSAR or NSOA, the changes would not increase the consequences of those accidents. The removal of these commitments did not affect fission 158
. . - - - - , - , m - m.< - - .,ev-
- e,v-r- - - - - + - ,ev-
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i product barriers. Consequently, the changes did not challenge the l integrity of any fission product barriers or increase the radiological consequences of any accident evaluated previously in the SAR or NSOA.
The changes did not degrade the ability of Fire Protection Systems and features to perform their intended functions. The changes did not adversely affect procedures governing maintenance and modification activities for Fire Protection Systems and features which provide " defense l in depth". Therefore, this change did not increase the consequences of an i accident evaluated previously in the SAR. The removal of these commitments apply only to Fire Protection equipment and systems. The requirement to prevent damage to redundant safe shutdown equipment as a result ofinadvertent operation or rupture of Suppression Systems was not affected by this change. Fire Protection equipment is not considered equipment important to safety and the removal of these commitments did not increase the probability of a malfunction of equipment important to safety and the changes did not incre= th: =.~ ....nces of a malfunction of equipment important to safety evaluated previously in the SAR.
Without interaction between Fire Protection Systems and safety related
! systems, structures or components, an accident of a different type than evaluated in the SAR is not possible. Fire Protection Systems and features provided to protect against fires are evaluated in the DAEC Fire Hazards Analysis. The removal of these commitments did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and it did not create new failure modes or increase the probability of Fire Protection Systems and features failing. Commitments to install, test
! and inspect the Fire Protection Systems using industry or property insurance provider guidelines were unaffected by this change. The removal of these commitments did not create the possibility of a !
malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. This change did not involve a change to a margin of safety. Additionally, Fire Protection features at the DAEC are l not addressed by the Technical Specifications or the Technical Specification Bases. Therefore, removal of these commitments did not impact the margin of safety as defined in the basis for any Technical l Specification.
i SE 98-09 UFSAR Change For Residual IIcat Removal (RIIR) Shutdown l Cooling Discharge Isolation Valves' Stroke Times l
Description and Basis For Chance l This UFSAR Change revised the RilR system shutdown cooling discharge j isolation valves, MOl905 and MO2003 design closure times from 18 i
seconds to 25 seconds. This was done since the subject valves' opening time was relaxed to a maximum of 28 seconds (even though the valves' i
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l nominal design stroke time is 18 seconds) per the LOCA Analysis. Since the closing and opening times are essentially identical due to the design of ]
i l the valves, the closing time contained in the UFSAR effectively limited
! {
the opening time as well. Thus, the allowable closing times for the subject ;
valves were increased in accordance with the design basis. The allowable '
opening times for the subject valves were previously increased via incorporation of the revised LOCA Analysis into the UFSAR.
The subject valves are the inboard LPCI inject valves and the outboard containmed/ reactor vessel isolation valves. These valves receive isolation signu b the RHR Logic to isolate on PCIS isolation signals whenever {
the RHR Sys, tem is in the Shutdown Cooling Mode of operation. These !
valves are listed in an Administrative Control Procedure as containment i isolation valves, but are exempted from Appendix J leak testing.
2 Safety Evaluation Summarv
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l The probability of occurrence of an accident previously evaluated in the SAR has not increased because the suoject valves' closure is not an accident which has been previously analyzed in the SAR. The closure (or l lack of closure) of the subject valves is not considered to be an accident.
i The consequences of an accident evaluated previously in the SAR are not ;
increased because the valves' closing time meets the design basis !
requirements for the reactor vessel and Primary Containment isolation functions for the subject valves. The increase in maximum allowed valve l i closure time did not degrade or prevent reactor vessel isolation actions l which are part of the SAR because the valves still close in sufficient time to isolate the reactor to minimize fluid flow from the reactor to avoid uncovering the core and avoid core damage. Additionally, the Primary Containment isolation function is not adversely affected since there is no l Primary Containment isolation design basis impact concerning the i maximum allowable valve closing time. Thus, there was no effect on dose i consequences. It is the subject valves' primary function to protect the fuel rod cladding by closing in sufficient time to isolate the reactor vessel to minimize fluid flow from the reactor vessel to avoid uncovering the core.
Additionally, they isolate the reactor vessel by closing to limit the release of radioactivity and also isolate the Primary Containment by closing to limit the release of radioactivity (although these are not their primary functions with respect to valve closure time requirements). The subject valves have no effect on Secondary Containment isolation functions. The increase in maximum allowed valve closure times did not degrade or prevent isolation actions because the subject valves still close in sufficient time to accomplish their design basis functions. Thus, there is no effect on p the fission product barriers or dose consequences. The probability of l- occurrence of a malfunction of equipment important to safety previously i
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L evaluated in the SAR was not increased. The subject valves are safety related but are not adversely affected by slower closure times. The subject valves still meet all seismic criteria, separation criteria and environmental qualifications. No additional loads not analyzed in the original design l were added, and no equipment protection features were deleted or i
modified. Support system performance necessary for reliable operation of the subject valves was not affected, and system / equipment redundancy or independence was not affected since the subject valves are in opposite ,
divisions. The frequency of operation of the subject valves was not affected and there was no increased or more severe testing requirements imposed on the valves. The inboard check valves continue to provide redundant isolation capability and still close in sufficient time to isolate the reactor to minimize fluid flow from the reactor to avoid uncovering the 3 core and avoid core damage. The consequences cf a malfunction of i equipment important to safety previously evaluated in the SAR were not increased because the maximum allowed valve closing time meets the requirements of the design basis. This activity did not create the l possibility of a malfunction of equipment important to safety of a i different type than any evaluated previously in the SAR because the valve !
maximum allowed closing time meets the requirements of the design basis. The margin of safety is not reduced because, based on a review of- )
the Technical Specifications, Technical Specification Bases, UFSAR and l
NSOA, there is no explicit margin of safety defined for these valves' closing times. The implicit margin of safety is for the ability of the valves '
to close in sufficient time to avoid uncovering the core and avoiding core damage by maintaining peak clad temperatures below allowable values.
Since the core remains covered, this margin remains unaffected. Thus, the l margin of safety is not reduced.
1 SE 98-10 Group 3 Isolation Logic UFSAR Change Descriotion and Basis For Chance L A discrepancy was identified between wording in the UFSAR and actual plant configuration. The UFSAR stated that all actuation signals for the l Group 3 logic can be overridden by key-lock handswitches. The triple low j l
reactor water level actuation signal cannot be overridden with a key-lock handswitch.' The wording in the UFSAR was changed from " Key-lock switches provided for override of ALL actuation signals," to " Key-lock ,
switches provided for override of all actuation signals excent Low-Low- l Low reactor water level signal."
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l Safety Evaluation Summary This change to the UFSAR more accurately states the isolation logic .
override capability for the mini-purge isolation valves. Triple low reactor '
vessel water level isolation in the Group 3 logic does not have isolation override capability. This is conservative with respect to the isolation !
logic, since it will remain in the tripped condition upon actuation without the capability of overriding it. Therefore, the probability of occurrence of an accident is not increased and the consequences of an accident are not I increased. If a triple low vessel water level signal is received, the mini- I purge isolation logic is actuated and cannot be overridden as the other i Group 3 logic signals can. Therefore, the consequences of an accident with respect to 10 CFR 100 can not be increased. This change did not increase the probability of occurrence of a malfunction of equipment l
important to safety. The consequences of a malfunciion of equipment l important to safety evaluated previously in the SAR were not increased and the consequences of a malfunction with respect to 10 CFR 100 were not increased. This change did not introduce any new failure modes or create the possibility of an accident of a different type. The plant equipment response remains the same and therefore consequences with j respect to 10 CFR 20 and 10 CFR 100 remain unaffected by the change. l Therefore, this change did not create the possibility of a malfunction of equipment important to safety of a different type. This change did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 98-11 Runback Of Recirculation Pump On Feedwater Pump Trip UFSAR Change Description and Basis For Chance The UFSAR Section," Runback of Recirculation Pump on Feedwater Pump Trip" stated," automatic runback of the recirculation pumps on a feedwater pump trip results in a reactor power reduction that is within the capabilities of the feedwater system with only one pump operating. The correction for the loss ofone feedwater pump is designed to be fast enough to prevent a reactor scram." This is not necessarily true. Based on experience, the automatic runback of the recirculation pumps on a feedwater pump trip does not allow the Feedwater System to respond fast enough to prevent a reactor scram. Therefore, the wording was modified to read," automatic runback of the recirculation pumps on a feedwater pump trip results in a reactor power reduction which may not be within the capabilities of the feedwater system with only one pump operating. The correction for the loss of one feedwater pump may not be fast enough to prevent a reactor scram." Also,"See section 7.9.4.3 of the initial FSAR,"
162
was deleted because it stated that a scram will not occur with a single feed pump trip.
Safety Evaluation Summary This change to the UFSAR only clarifies current plant response and does not increase the probability of occurrence of an accident evaluated previously in the SAR. The consequences of an accident are not increased as a result of this change to the UFSAR. This change did not cause an increase in the consequences of an accident because it is bounded by the transient described for Loss of Feedwater Flow in the UFSAR. The scram will shut the plant down to a " safe" condition and 10 CFR 100 and 10 CFR 20 limits will not be impacted. Therefore, the consequences of an accident evaluated previously in the SAR were not increased. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety. The Loss of Feedwater Flow transient is already described in the UFSAR and has already been evaluated. No physical changes occurred in the plant as a result of this change. The consequences of a malfunction of equipment important to safety are not increased as a result of this change to the UFSAR. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. By stating that the plant may potentially scram on a loss of a feedwater pump, no accidents of a different type were created.
No new failure modes were created by this change because this activity is not an accident initiator. No new malfunction is anticipated for the CRD system as a result of the proposed change that already has not been analyzed. Therefore, this activity did not create the possibility of a malfunction of equipment important to safety of a different type. The recirculation pump runbacks are not described in the Technical Specifications' bases and are a non-safety related function. In addition, the plant response to a feed pump trip is to scram conservatively; therefore, this change may actually increase the margin of safety as defined in the basis for any Technical Specification.
SE 98-13 Revision To P&ID For Steam Jet Air Ejectors Description and Basis For Chance During a review of an Operating Instruction, it was noted that the P&lD for the Steam Jet Air Ejectors showed the Condenser Vacuum Pump Moisture Separator Outlet Isolation Valve and the Condenser Vacuum Pump Seal Water Supply Pressure Indicator Root isolation Valve as normally closed. These valves are normally open. The function of the valves is to isolate the Offgas System flow from the Condenser Vacuum Pump and to isolate the Condenser Vacuum Pump Seal Water Supply 163
L Pressure Indicator, respectively. The valves were installed as original l plant equipment. The P&ID has been revised to indicate the valves' position as normally open.
Safety Evaluation Summan' j This change involved showing the isolation valves for the Mechanical Vacuum Pump on the Steam Air Ejector System P&lD in the normal operating state. The piping and valves are non-safety related. The change L - to the isolation indication on the P&lD did not impact the Steam Air l
Ejector System. The system still functions as intended. This activity did not increase the probability of occurrence of an accident evaluated 3
previously in the SAR. The Steam Air Ejector System is not required to !
prevent or mitigate an accident. This activity did not increase the consequences of an accident evaluated previously in the SAR. An assumed malfunction of this equipment would not impact the system in
! any way, because the valves are not required to isolate, or to be isolated.
l This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in tl e l i SAR. An operational failure or a component failure of this system does l l
not result in a site boundary dose that is an appreciable fraction of 10 Cl!R i 100 doses. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No credible accident could be created by this change, because the valves are passive, normally open. This activity did not create the possibility of an accident of a different type than any evaluated previously l- -in the SAR. No failure modes that could impact the performance of the j system could be identified by this change. This activity did not create the ;
possibility of a malfunction of equipment important to safety of a different i type than any evaluated previously in the SAR. The proposed change has no impact on the surveillance tests for the equipment associated with the change. This activity did not reduce the margin of safety as defined in the basis for any Technical Specifications.
1 SE 98-19 Revision To UFSAR Sections 3.6.1 Through 3.6.3 Description and Basis For Chance L This UFSAR change incorporated miscellaneous editorial changes in sections 3.6.1 through 3.6.3. These changes resulted from a revision of the Pipe Break Inside Containment Design Basis Document. The changes L incorporated previously omitted an more accurate information with respect to the supporting design documentation. In addition, the DBA
[ LOCA total break area was corrected as well as the inside diameter of the
- maximum secondary piping break sizes corresponding to piping breaks in J
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the Recirculation piping, the Main Steam and RIIR discharge piping and the RIIR suction piping.
l Safety Evaluation Summary Updating the DBA LOCA break area to include the correct value did not increase the probability of occurrence of an accident previously evaluated in the SAR. Revising the inside diameter of the piping sizes for maximum secondary piping breaks permitted did not increase the probability of occurrence of an accident since the only piping which was not previously encompassed by the previous data (when taking into consideration the previous DBA break area) was the High Pressure Coolant injection System piping. The UFSAR indicates that this piping is restrained with respect to the Recirculation piping as a source. Furthermore, by reducing this maximum inside diameter the resulting break area is now below the total area of the DBA LOCA. The probability of an accident evaluated previously in the SAR was not increased and the consequences of an accident evaluated previously in the SAR were not increased. The original ,
design for protection against the dynamic efTects associated with
)
postulated rupture of piping systems inside Primary Containment, was i maintained. The probability of occurrence of a malftmetion of piping or other equipment which is important to safety previously evaluated in the l SAR was not increased by this activity. This action did not directly I impact the operation of equipment important to safety, but clarified the I
guidance outlined in the SAR to prevent the possibility of over pressurization inside Primary Containment. As a result, the consequences of a malfunction of equipment important to safety were not inernsed, and the possibility of an accident of a different type than an, evaluated
- previously in the SAR was not created. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
This activity ensured that the maximum containment pressure conditions previously evaluated for the maximum DBA area were not implicitly exceeded by the information being included in the UFSAR.
l SE 98-20 Replacement Of Noble Metal Chemical Addition Rods l
Description and Basis For Chance The following noble metal chemical addition (NMCA) activities were performed for once-burned, twice-burned and thrice-burned fuel bundles during Refuel Outage (RFO) 15:
d 9
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e Replacement of an undamaged pre-treated NMCA fu'.:1 rod from a once-burned fuel bundle with a new non-treated NMCA fuel rod.
e ' Storage in the spent fuel pool of this fuel rod and another undamaged fuel rod from a thrice-burned fuel hundle that was l treated with noble metal during RFO 14. These rods were stored until shipment to General Electric (GE).
l e 1 Storage in the spent fuel pool of the thrice-burned fuel bundle with one fuel rod permanently removed and a maximum of two '
replacement spacers.
- Combined Instrument Measurement System (COINS) examination I
of the once, twice, and thrice burned bundles involving partial bundle disassembly and the use of the fuel prep machines.
The purpose of removing the fuel rods and spacers was to allow General Electric to confirm that the treatment of NMCA on the fuel rod surfaces l
and other Zircaloy bundle components does not accelerate Zircaloy '
corrosion, increase the uptake of hydrogen or change the nature of the fuel deposits in a negative manner. ,
Safety Evaluation Summary There was no increase in the probability of occurrence of an accident l
and/or abnormal condition. Based on a review of the Refueling Accidents, ;
the probability of either dropping or damaging a fuel assembly during removal or installation of the fuel bundle on or around the spent fuel pool racks was not increased. These activities did not increase the consequences of a fuel failure or fuel handling accident evaluated j previously in the SAR. The radiological consequences of the fuel i
handling accident bounds this activity. The maximum dose possible from l- this activity is less than the previously evaluated fuel handling accident l identified in USFAR Chapter 15. Activities for disassembly and reassembly of the once, twice, and thrice burned fuel bundles was performed by qualified personnel in accordance with General Electric's approved fuel handling procedures. The fuel bundles returned to the reactor core were reassembled in the same configuration. Storing the fuel l pins in the approved nine-pin container did not increase the consequence l of an accident. The nine-pin holder is designed to store fuel pins or water rods in a configuration similar to a fuel bundle. No fuel handling
- equipment was modified to perform this activity. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR.. Replacing the fuei rod with a fuel rod designed to the same specification, only with a lower 166
1 enrichment, did not affect the overall perfonnance of the fuel rod or fuel bundle. A thorough design review was performed by GE which determined that the replacement fuel rod meets the design safety basis.
The bundle is mechanically and neutronically equivalent to the original l bundle. NRC Generic Letter 90-02, Supplement 1, states that I reconstitution of a fuel assembly to replace damaged and leaking fuel rods is not considered to be an unreviewed safety question if the repaired fuel assembly constitutes a previously approved design. The thrice-burned bundle was permanently discharged and the structural and seismic configuration of the bundle with one fuel rod missing was not compromised since the tie rods. were not removed. The replacement l spacers are of the same design and are located in the same position on the bundle. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The radiological consequences of any malfunction to the bundle is unchanged.
The radiological consequences of a dropped (and ruptured) fuel rod caused by the malfunction of tools used to grapple the rod are bounded by the dose from the fuel handling accident. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and this activity did not create the possibility of a malfunction of equip. ment important to safety of a different type than any evaluated previously in the SAR. The margin of safety as defined in the basis for Technical Specifications was not reduced by replacing the new fuel rod or storing the fuel rods in the spent fuel pool in a nine-pin holder. The margin to safety was not reduced for any core thermal limits as a result of replacing the fuel rod in the once-burned fuel bundle. The replacement fuel rod meets all of the design characteristics of a GE 10 tuel bundle.
SE 98-21 (Revision 1) Leakage Reduction Program UFSAR Change Description and Basis For Chance During review of Technical Specifications it was discovered that the requirements for quarterly walkdowns for the Leakage Reduction Program needed to be clarified, specifically, whether the Scram Discharge Volume (SDV) and the Containment Atmosphere Monitoring System (CAMS) required quarterly walkdowns. Further review also identified that the Containment Atmosphere Sampling Subsystem (CASS) of the Post Accident Sampling System (PASS) was not inspected for leakage on a cyche basis nor was a quarterly walkdown performed. An UFSAR change was completed to clarify that quarterly walkdowns will not be performed on the SDV, CAMS, or PASS-CASS. This UFSAR change l also added the cyclic walkdown for PASS-CASS. The procedure for the l PASS-CASS was added to the CAMS cyclic inspection. A change was l also made to clarify that cyclic inspections on gaseous systems will be 167 l
performed by measuring leakage, and walkdowns will be performed as necessary to eliminate the sources of excess leakage.
Safety Evaluation Summary The accidents that have been evaluated in the SAR that may be affected are a LOCA inside or outside of containment. Not performing quarterly walkdowns of the SDV, CAMS and PASS-CASS does not increase the probability of these accidents. During normal operation, the scram discharge header is vented and drained. Hence, a visual inspection for leakage is not possible during normal operation. The Leakage Reduction Program indicates that the quarterly walkdowns will be general inspections for visible leakage. A general inspectioa for visible leakage on a gaseous system provides no meaningful data. Performance of cyclic testing on the PASS-CASS does not increase the probability of the
< occurrence of an accident, and the consequencer c f a malfunction of equipment important to safety is not increased sir.ce the system is isolated from the containment during testing and tlie passive Primary Containment function is maintained. The consequences af an accident evaluated previously in the SAR were not increased. The probability of a malfunction of equipment important to safety was not increased, the possibility of an accident of a diffenmt type was not created, and the possibility of a malfunction of equipment important to safety of a different type was not created since the system is isolated fmm the contaimnent during testing ar.d the passive Primary Containmer.t function is tested to ensure its integrity prior to return to service. Tht: margin of safety as dermed in the basis for any Technical Specificatian was not reduced.
Performance of cyclic testing on the PASS-CASS ensures the margin cf safety is maintained.
SE 98-22 Reactor Building Auxiliary Heating System P&lD Revisten Description end Basis For Chance The tap off of tl.e Reactor Building heating loop for the Reactor Bailding Heating Hot Water Supply Expansion Tank was shown in the wrong position on the P&lD for the Reactor Building Auxiliary Heating System.
The tap off was shov/n on the wrong side of the Reactor Building 855' Level Hot Water Retuni Header Isolation Valve. The Expaneion Tank Drain Valve was shown on the Reactor Building Heat I cop Air Removal Fitting and not on the Expansion Tank. The identification number for the Reactor Building Heat Loop Air Removal Fitting was also incorrect. The P&ID has been revised to correct these errors.
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l l- Safety Evaluation Summary This change did not increase the probability of an accident previously evaluated in the SAR. The Reactor Building Hot Water Supply Expansion Tank operates in the same manner as it did previously. The consequences of an accident previously evaluated in the SAR were not increased, and the probability of occurrence of a malfunction previously evaluated in the SAR was not increased. The operation of the Reactor Building heating loop was not affected by this change. The consequences of a malfunction of equipment important to safety previously evaluated in the SAR were not increased. The possibility of an accident of a different type than any evaluated previously in the SAR was not created. The possibility of a release of radioactive material was not created. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The margin of safety as l defined in the Technical Specifications was not reduced.
SE 98-25 UFSAR Change For Feedwater Turbine Trip On High Level Description and Basis For Chance
! This change revised the UFSAR description of the Feedwater Controller Failure - Maximum Demand (FWCF) transient in Section 15.1.1 to correct misleading statements about the Reactor Pressure Vessel (RPV)
High Water Level setpoint (Level 8) being in Technical Specifications, and add information about the role of the Level 8 trip of the Main Turbine (and Feedwater pumps) based upon a General Electric sensitivity study.
- This change was the result of the NRC's review ofImproved Technical Specifications (ITS) and the determination that the trip did not satisfy the 10 CFR 50.36(c)(2)(ii) screening criteria for inclusion in the ITS. No physical changes to the trip logic or setpoint were made, nor were any operation, maintenance or testing procedures affected by this change.
Safety Evaluation Summary This change did not increase the probability of occurrence of an accident evaluated previously in the SAR. No physical changes to the FW Control System were made by this change. Only the description of the FWCF j event in the UFSAR to accurately reflect the role of the turbine trip on l RPV high level (Level 8) was updated and the incorrect cross-reference to .
[ the setpoint being in the Technical Specifications was deleted. Therefore, the pmbability of a FWCF transient was not increased. The probability of an RPV overtill event was not increased. No changes in operation, maintenance or steveillance testing practices or frequencies for these 4
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l components were made by this change. The probability of occurrence of a l malfunction of equipment important to safety was not increased, the l consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased, and the possibility of a new or different type of accident was not created. The possibility of a new or different type of nialfunction was not created. The margin of safety as defined in the basis for any Technical Specification was not reduced. The Level 8 trip is not essential to protect any reactor safety limits (either fuel thermal limits or RPV overpressure) provided that: 1) the current trip setpoint is not lowered to the point where thermal power is still increasing during the FWCF transient at the time of the turbine trip and scram; and,
- 2) that the Operators insert a manual scram on high water level when '
operating below the automatic scram cut-out on first stage turbine pressure
(~30% power). As these provisions are currently being met, the existing l
margin of safety is not reduced. However, by including these provisions ;
in the UFSAR description, we have ensured that any future changes are l properly evaluated to maintain the existing margin of safety.
SE 98-26 Welder Qualifications Description and Basis For Chance l
ASME,Section IX, contains requirements for the preparation and qualification of welding procedure specifications, and the qualification of welders and welding operators. ASME Section IX, Article III, contains specific requirements for welder and welder operator qualification. A
)
review of contract welder qualifications for Refuel Outage (RFO) 13 and RFO 14 revealed that three welders were not fully qualified for work I performed.
1 Safety Evaluation Summarv l l
Pressure boundary welds were made in accordance with the design specifications using qualified weld procedures. Completed welds were inspected to established inspection criteria and were found acceptable.
Welders' qualifications were evaluated and found to be acceptable.
Therefore, the welds did not increase the probability of an accident evaluated previously in the SAR, and there was no increase in the consequences of an accident evaluated previously in the SAR. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased, and this activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The possibility of an accident of a different type than any evaluated previously in the SAR was l not created. An evaluation demonstrated that the welds are acceptable for 170
l continued use. The possibility of an equipment malfunction of a different type than any evaluated previously in the SAR was not created, and the margin of safety as defined in the basis for any Technical Specification was not reduced.
SE 98-28 Offgas System P&lD Revision Descrintion and Basis For Chance The Offgas System P&ID did not agree with the plant operation of the Control Air Supply Solenoid Valve for the 'A' Offgas Cooler Condenser.
Both Offgas Cooler Condensers are in service during normal operation.
The P&lD showed only one in service at a time. The positions of various valves on the OfTgas System P&lD were revised to reflect both Offgas Cooler Condensers as being in service. Other minor enhancements were also made to clarify coordinates and symbols.
Safety Evaluation Summary i
This activity did not increase the probability of occurrence of an accident previously evaluated in the SAR. Operating both Cooler Condensers does not degrade the operation of the Offgas System as described in the SAR, nor prevent the Offgas System from performing its design function. The Offgas System is not safety related. Because this change does not affect i the system performance in a manner that could lead to an accident, the probability of an accident previously evaluated in the SAR was not <
increased. This activity did not increase the consequences of an accident evaluated previously in the SAR. The Offgas System is not required for i the mitigation of any accidents defined in the NSOA and UFSAR. This ]
change did not prevent the OITgas System from performing its design basis l function as defined in the SAR. This change did not prevent nor degrade any essential safety function assumed by the NSOA to mitigate the consequences of design basis accidents. System reliability was not affected by this change. The probability of occurrence of a malfunction of l equipment important to safety evaluated previously in the SAR was not increased. The same equipment is used and it is operated in a manner for which it was designed. An operational failure or a component failure of j
the Offgas System does not result in a site boundary dose that is an appreciable fraction of 10 CFR 100 doses. This activity did not increase the consequences of a malfunction of equipment important to safety l evaluated previously in the SAR. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR. No new mechanical failure modes were introduced that affect equipment important to safety. The isolation logic was not affected. This modification did not create the possibility of a malfunction of equipment j 171 l
important to safety of a different type than any previously evaluated in the SAR. This modification did not change or affect any setpoints or surveillance requirements as defined in the Technical Specifications. This activhy did not reduce the margin of safety as defined in the basis for any Tnhnical Specification.
SE 98-38 (Ihvision 1) IIcating, Ventilation and Air Conditioning Drawing Changes Descriotion and Basis For Chance Bell & Gossett " Triple Duty Valves" and "Flo-Control Valves" are used in plant Heating, Ventilation and Air Conditioning (HVAC) Systems. The
" Triple Duty Valve" has three separate functions (check, isolation and throttle), while the "Flo-Control Valve" is designed to prevent gravity I flow and is also equipped with a handle to override this function and to allow for gravity circulation. Several HVAC P&lDs incorrectly identified ,
" Triple Duty Valves" as check valves and vice versa. Field walkdowns l were performed and the affected P& ids were revised as required to identify Bell & Gossett " Triple Duty Valves" with the correct symbol and to identify Bell & Gossett "Flo-Control Valves" with a normal check valve symbol. Other design drawings were revised to correctly identify valve l
types and to add / correct any equipment identification numbers. In addition to the document changes, it was assured that all "Flo-Control Valves" were placed in the non-override position, that Operations personnel were trained on the two valves, and that Operations Department Instruction," Tagging Practices" was revised to allow the use of" Triple Duty Valves" for isolation purposes. The equipment names in the equipment database were revised so that all " Triple Duty Valves" are now called " Triple Duty".
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Safety Evaluation Summary This activity made no physical changes to the plant. The valves are all part of plant HVAC heating and cooling water loops and are all non-safety related. The power generation objective of the HVAC Systems was not compromised by this activity. This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR.
There were no radiological consequences associated with this activity. No equipment important to safety was impacted. This activity did not increase the consequences of an accident evaluated previously in the SAR, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. The consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased. This activity did not create the possibility of an accident of a different type than any evaluated previously 172
! I l l l in the SAR. No new failure modes were created and no equipment important to safety was impacted. This activity did not create the possibility of a malfunction of equipment important to safety of a different l l
type than any evaluated previously in the SAR. The HVAC heating and cooling water components associated with this activity are not mentioned l in the DAEC Technical Specifications. This activity did not reduce the '
margm of safety as defined in the basis for any Technical Specification.
! I SE 98-39 Deletion Of Stability Power / Flow Map From UFSAR Description and Basis For Chance
< The DAEC Stability Power / Flow Map was deleted from the UFSAR and
! revised in order to provide a map in which every line is drawn from plant specific equations and/or data points and is easier to read. The new map is controlled on a cycle-to-cycle basis in the Core Operating Limits Report (COLR) and is still part of the DAEC Nuclear Licensing Basis. This i
revision to the Power / Flow Map was a result of the Natural Circulation Line (NCL) change. It was decided to draw and implement a new version of the Power / Flow Map using graphics software and the collected data.
Along with this action, an effort was made to make the map easier to read l by taking comments from various Control Room crews. When the revised map was completed, only the Natural Circulation and Cavitation Protection Lines were changed from the previous version. The NCL on l the revised Power / Flow Map was the major change that occurred. The new line was drawn from data points collected from DAEC 1974 start-up data. This was a change to the map that affected the analyzed initial j conditions of a dual Recirculation Pump trip, but it was not part of the normal operating regions for the plant. The NCL is a prediction of what the plant would do, with respect to power and flow, in a dual Recirculation Pump trip event. The revised NCL matches the results of a 1995 GE Stability Analysis for the DAEC that denotes reactor power will be approximately 48% of rated after a dual pump trip. This initial condition change resulting from the revised NCL is consistent with a GE analysis.
The changes te the NCL also shifted the left-hand side of the Buffer and Exclusion Zones, but the right-hand side which intersects the four loadlines remained constant. This ensures safe operation of the plant because the operating margin was not changed. The Cavitation Protection Line name was changed to the Low Feedwater (FW) Protection Line to l better describe its function. Along with this change, the line itself was moved to a constant value of 22% of rated power for all rated flows. The change was made to be consistent with DAEC start-up data. The cavitation line is used to provide a buffer to the Jet and Recirculation Pump Net Positive Suction Head (NPSH) Limits which have a maximum value of approximately 15% of rated power at 100% rated flow. No safety
! 173 l
margins were decreased by this change. The Jet and Recirculation Pump NPSH Limits provide the margin of safety needed to protect these pumps.
Safety Evaluation Summary l
The Power / Flow Map is used by the Control Room as a reference for stable regions in which the plant can safely operate. This is derived from the UFSAR that states the current Power / Flow Map is used as a reference for core flow and loadline to minimize the likelihood ofinstability events, i.e. reactor power oscillations. Only changes in the Buffer and Exclusion Zones (i.e. decay ratios) or the Low FW Protection Line, could lead to the probability of an instability event increasing. With respect to the Buffer and Exclusion Zones, the change would have to be one that places normal operating regions at potentially high decay ratios. In other words, the right-hand side of the Buffer and Exclusion Zone would have to be moved to the left, with decreasing flow. Because the right-hand side of the zones were not changed in this revision, the probability of an accident occurring has not been increased with respect to the Buffer and Exclusion Zones.
The Low FW Protection Line is used as a buffer to protect the Jet and Recirculation Pump NPSH Limits, and the Recirculation Pump runback to 20% speed is used to protect the pumps from cavitating at low flows. The margin between the Low FW Protection Line and the NPSH Limits has not changed, so the probability of an accident occurring because of the Low FW Protection Line change was not increased. The flow biased APRM SCRAM function limits the consequence of a power oscillation event by scramming the reactor at a given setpoint. The consequence of
, any accident was not increased. Because the Recirculation Pump runback i
has not been changed, the dose consequences of an accident has not increased. This revision to the Power / Flow Map did not change any design, ftmetion, or method of performing the function of any structure, l system, or component in the plant. Consequently, the increase in the probability of flow oscillations did not increase. All systems and equipment operate as before, with no additional loads or requirements, and l
all previous safety analyses are applicable. Therefore, the probability of equipment malfunction was not increased. The normal operating region is essentially the same and all safety margins are the same, so the consequences of a malfunction were not increased. Because the plant will not be operated in any new regions, all previous accident analyses are l bounding and no accidents of a different type than evaluated in the SAR are possible. Therefore, this revision did not create an accident of a i different type than any previously evaluated. No new malfunctions of equipment were created by this change. The margin of safety as defined in the basis for any Technical Specification was not reduced. The Buffer Zone was partially eliminated on the left-hand side, but the right-hand side remained constant. This is the side of the boundary that the plant would 174 i
encounter with decreasing flow along the loadlines, and because it remained constant, so did the margin of safety associated with it. The Low FW Protection Line change did not reduce the margin of safety associated with nonnal operations. The line does not represent a true limit, but rather a " buffer" region between the actual NPSH limits of the Jet and Recirculation Pumps, and the simulated runback to 20% speed the recirculation pumps would encounter with a loss of Feedwater. The line was changed so that no safety margin was lost and nonnal operating regions remained relatively constant.
SE 98-46 Revision To P&lD For IIVAC and Air Flow For The Standby Gas Treatment (SBGT) System Description and Basis For Chance It was discovered during tagouts that the 20 psi Air Supply Isolation Valves on the HVAC and Air Flow P&lD For the Standby Gas Treatment Systems did not exist. The valves at the locations on the P&lD, downstream of the 'A' and 'B' SBGT Instrument Air Supply Pressure Control Valves (PCVs), were actually relief valves. The relief valves are part of an assembly and were installed at the same time as the PCVs. The isolation valves have been deleted from the P&ID and relief valves (PSVs) have been added.
Safety Evaluation Summary These changes did not increase the probability of occurrence of an accident evaluated previously in the SAR. The affected systems, Control Building IIVAC, Primary Containment Isolation, and Standby Gas Treatment all function to mitigate the consequences of an accident by assuring the habitability of the Control Room and by assuring that radioactive material releases to the environs are below the values of 10 CFR 100. There were no increased radiological consequences associated with this activity. The changes did not increase the consequences of an accident evaluated previously in the SAR. The absence ofisolation valves is not detrimental to system operations since the instrument air supplies to the SBGT trains can be isolated using other isolation valves. The addition of pressure relief valves downstream of the PCVs provides additional assurance that the desired pressure of 20 psi to the applicable SBGT components will not be exceeded should the PCVs fail to maintain pressure. Total failure of a PSV is extremely remote. With the redundancy and the fail-safe design of the systems / components, it was concluded that the benefit, to assure overpressurization protection to sensitive instrumentation and to avoid a potential pipe rupture situation, outweighed the risk. The changes did not increase the probability of 175 f
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occurrence or the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new failure modes were created by this activity since the design, function and method of l
performing the function of the affected systems remained the same. The changes did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The ability of affected systems and their components to perform as designed to mitigate the consequences of an accident was not changed by this activity. The margin of safety as defined in the basis for any Technical Specification was not reduced.
SE 98-48 Special Test Procedure (SpTP) 199 - Cell Phone System Descrintion and Basis For Chance The Radiation Protection Department purchased the Ericsson Model DB600 (Cell Base Station), and the Model DCT1900 (Portable Cell Phone IIandset) Cell Phone System. A test of this Cell Phone System was conducted in accordance with SpTP-199. This is a commercially available Cell Phone System, designed to connect directly to commercial Private Business Exchange (PBX) systems. The base stations and phones are both Radio Frequency (RF) Transmitters. An intra-site Cell Phone System was chosen to supplement the existing wired phone system, in order to provide improved communications within the DAEC complex. The cell phone )
system was evaluated for installation in the following locations: Reactor i Building, Turbine Building, Radwaste Building, Low Level Radwaste Buildings, Control Building, Technical Support Center, and the Data Acquisition Center. The Drywell was evaluated for temporary installation during outages.
The testing was performed using a test rig to simulate the electro-magnetic / radio frequency interference (EMI/ RFI) of the future cellular phone equipment (CPE). The test rig was an Ericsson DCT 1800 Model System Demonstration Kit which conservatively simulated the system to l l be installed. The electronic test unit operated at a frequency of 1880 MIIz and at a power output of 250 milliwatt. The DCT 1900 produces a 1920 Mllz signal at 95 milliwatt. Rockwell(Collins Avionics &
i Communications Division) provided a portable field strength measuring l l
device. The field strength meter was intermittently used to verify the cell l phone was transmitting. The Special Test was performed at locations identified as having (EMI/RFI) sensitive equipment and monitored locally
, and in the Control Room. Certain equipment was required to be in operation during the tests. It was expected that the operation of this 4
equipment would be affected in the same way as during plant power
- operation. The test was conducted with the plant shutdown as part of a 176
l progressive testing plan to verify that the cell phones would not introduce l- any new malfunctions of equipment important to safety.when the phones are permanently installed. This test followed testing which was performed L in the plant's simulator. The simulator testing demonstrated that the cell I phones had no effect on instrumentation installed there. While this
- instrumentation is similar to that installed in the plant, there are
! differences which warranted additional in-plant testing performed by this L SpTP.
Safety Evaluation Summary This Special Test was performed during plant shutdown. The potentially )
l affected equipment was carefully monitored during the testing for any l effects on operation. Any effects noted on the equipment would have resulted in immediate cessation of the testing and return of the equipment to normal operation. The only credible effect of this test on plant equipment was a transient effect that would have stopped when the test l
was stopped. There were no credible effects of this testing which would have led to the initiation of an accident. Therefore, no increase in the probability of occurrence of an accident evaluated previously in the SAR resulted from this Special Test as it did not impact any event initiators. I L This activity did not increase the consequences of an accident evaluated l previously in the SAR. There was no increased dose consequence as a result of this test. Therefore, the normal actions required to mitigate the L effects of an accident were unaffected by this testing and this activity was bounded by the accident analyses contained in the SAR. This activity did
( not increase the probability of occurrence of a malfunction of equipment l important to safety evaluated previously in the SAR. There was the i unlikely possibility that this testing may cause a safety system to function as a result of a spurious signal; however, this would not have resulted in a system level failure of that equipment. High-powered radios were used at the DAEC in the vicinity of sensitive equipment for many years. There
[ were a few incidents which were attributed to the use of these radios which resulted in the suspension of the use of these devices in these areas.
L These radios were in the 1 to 4 Watt range. The cell phones have a maximum output of approximately 90 milliwatts. This history in conjunction with the simulator test data demonstrated that the cell phones would have no effect on plant equipment; however, in-plant testing was prudent to verify this with the installed equipment. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated p,reviously in the SAR. All required plant equipment was i available to respond to an event as required. The only credible effects on i
l- plant equipment would be the unlikely spurious actuation of a component
{ or system. This would not have increased the consequences of any malfunction. There were no credible effects of this testing which would 177 l
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have lead to the initiation of an accident of any type. Therefore, no accident event initiators were introduced which could create the possibility of an accident of a different type than any evaluated previously in the SAR for the given plant conditions. This activity did not change any existing plant systems, structures, or components. The possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. This Special Test did not change any parameter included in the basis for any Technical Specification.
EMI/RFI testing which could affect Average Power Range Monitoring (APRM) setpoints was done with the mode switch in shutdown, with all rods inserted and rod withdrawal motion blocked. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 98-50 UFSAR Change Eliminating Reference To Post Accident Sampling System (PASS) Gaseous Iodine Sample Description and Basis For Chanue This UFSAR change eliminated reference to a gaseous iodine sample with respect to the Post Accident Sampling System.
L Safety Evaluation Summarv j' The PASS is not an accident initiator. It provides sampling capabilities post accident. The PASS continues to be capable of providing liquid and gaseous samples for use in determining estimates of core damage.
Therefore, this change did not increase the probability of occurrence of an accident evaluated previously in the SAR. The PASS gaseous iodine samples from the Drywell and Torus are not relied upon as indicators of the degree of core damage following an accident. The Fuel Damage Assessment does not rely on the gaseous iodine PASS sample to determine core damage. Therefore, this change did not increase the consequences of a previously evaluated accident. No changes were made to the manner in which any safety related equipment is operated or j maintained. This change did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR This change did not create the possibility of an accident of a different type than l any evaluated previously in the SAR, and the possibility of a malfunction
- of equipment important to safety of a different type than any evaluated 1
previously in the SAR was not created. The Technical Specification l requirements for PASS are based on NUREG-0737. In the response to NUREG-0737, the DAEC did not credit a gaseous iodine sample to l
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l
determine core damage. This change did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 98-54 Training Program UFSAR Change Description and Basis For Chance Previously the UFSAR could be interpreted to require Instructors and the Mechanical, Electrical and Instrumentation and Control Maintenance Supervisors to attend all technical training scheduled for their respective shop personnel or trainee (s). This method did not provide the flexibility to accurately target the specific technical training needs of the instructor or maintenance supervisor positions. The selection criteria is based on position specific performance needs, technical expertise, current industry issues and management input. Refresher Training was changed to Continuing Training for all programs to reflect current training course descriptions. This activity was limited to the description of the DAEC training programs as listed in Chapter 13.2 of the USFAR. No systems or components were affected by this change. This change did not alter the design , function, or method of performing the function of any structure, system or component of the plant. Consequently, no Design Basis Documents were affected or applicable.
l Safety Evaluation Summary This change only affected the description of the training programs. It did not alter scope or reduce the effectiveness of the training programs. As a result, the probability of occurrence of" personnel errors" as initiating or contributing actions to the accidents evaluated in the UFSAR was not increased, and the consequences of an accident evaluated previously in the SAR were not increased. This change did not affect any system, structure or component important to cafety evaluated in the UFSAR. Consequently, this change did not increase the probability of occurrence of any safety related equipment malfunctions, and the consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased. The possibility of an accident of a different type than any evaluated previously in the SAR was not created. The possibility of a malfunction of equipment important to safety of a different type was not created. The margin of safety as defimed in the basis for any Technical Specification was not reduced.
i 179
I SE 98-55 Main Steam P&ID Revision i l
Description and Basis For Chance l
Discrepancies were identified on the P&lD for Main Steam, Iligh and Low Pressure Turbines, in showing the root isolation valves for seven pressure l
points (PPs) and six sample points (SXs) as normally open, while the l
Operating Instruction (OI), required them to be closed. A review of'his drawing also identified three additional SXs, the isolation valves for which were shown open, while the OI valve lineup required them to be closed.
The P&ID has been revised to show the root isolation vah'es as normally closed.
Safety Evaluation Summary This change involved showing a number ofisolation valves in the pressure point and sample point branch lines in the Main Steam System as normally closed, instead of normally open, in accordance with the Operating Instruction. These valves are designed for both normally open and normally closed positions. Therefore, the function of the valves, which is to maintain the system pressure boundary, was not impacted in any way.
This acu m., "i not increase the probability of occurrence of an accident evaluated previc usly in the SAR. Because the open/close position of these valves is not r.;evant to maintain the integrity of the system, and the valves are aesigned per the applicable codes, this activity did not increase the consequences of an accident evaluated previously in the SAR. The valves meet the safety standards for the system. Changing their position in accordance with the Operating Instruction did not increase the probability of occurrence of a malfunedon of equipment important to safety evaluated previously in the SAR. The consequence of a malfunction is the release of radioactive steam into the Turbine Building. Changing the valves from one position to the other has no affect on the system, because the open/close position of these valves is not relevant to maintain the integrity of the system. The system will function as before. Therefore, this activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and the possibility of an accident of a different type than any evaluated previously in the SAR was not created. This activity did not create the possibility of a malftmetion of equipment important to safety of a different type than any evaluated previously in the SAR. The subject valves are not included in the Technical Specifications. This change did not reduce the margin of safety as defined in the basis for any Technical Specification.
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SE 98-56 Containment Atmosphere Control System P&lD Revision Description and Basis For Chance The valve which isolates the nitrogen gas purge from the Containment Atmosphere Control System to the Traversing Incore Probe (TIP) indexing mechanisms was shown normally open on the Containment Atmosphere Control System P&ID, while the Operating Instruction (OI) required the valve to be normally closed. This valve is required to be opened for TIP system operation and closed at the end of operation. The P&lD has been revised to indicate the valve is normally closed.
Safety Evaluation Summary This activity changed the position of a non-safety related isolation valve from normally open to normally closed. This valve is manually operated to allow the nitrogen gas purge to the TIP indexing mechanisms, whenever the TlP system is operated. Therefore,it has no impact on any SSC. This activity met the applicable design codes, and is judged not to increase the likelihood of an accident. This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. Neither the TIP system nor the Nitrogen Gas Purge System is an accident initiator or mitigator. This change did not affect the radiological consequences of any accident, because the valve has no safety function, and it has no affect on any SSC important to safety. This activity did not increase the consequences of an accident evaluated previously in the SAR. This activity met the system design specifications, and it did not degrade any SSC reliability. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The purpose of the nitrogen purge is to control relative humidity, which could affect both driving properties and flux profiles. However, this is a manual valve, and it is required to be opened when the indexing mechanisms are operated, in accordance with the 01.
Any malfunction of the valve to open would be noticed from the pressure indication downstream and corrective action would be initiated. The TIP System functions as before, and there is no impact on the consequences of an accident. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, the possibility of an accident of a different type than any evaluated i previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. The TIP System and the Nitrogen Purge System that are asweiated with this valve are not included in the Technical Specifications. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
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i SE 98-57 Condensate Demineralizer P&ID Revision Description and Basis For Chance The spare dome valves for the Condensate Demineralizers were shown open on the Condensate Demineralizer P&ID and should have been shown closed. The tables for these valves, also located on the P&ID, were also updated to reflect this change. The pressure switches that were once supplied by these valves were removed under a previous revision which failed to show these valves as closed.
The Condensate Demineralizer System schematics were updated to show SV1732A-E, Condensate Demineralizers' Outlet Flow Indicating Controller Air Supply isolation, in parallel with SV1719A-E, Condensate Demineralizers' Effluent Valve Control Air Supply Isolation. The electrical connections for these solenoid valves are now shown between HS1719E, 'E' Condensate Demineralizer Effluent Valve Handswitch, and FS1733E, 'E' Condensate Demineralizer Flow Switch, on the P&lD.
Train 'E' on the P&lD is typical for all trains. The P&ID does not specifically show all components for the remaining trains. Further review i determined the P&ID also needed to be updated to better reflect the electrical connections. SV1732A shuts off the air supply to CV1719A, the
'A' Demineralizer Effluent Valve, when the bed is placed in hold or when CV1719A is closed. This causes CV1719A to close. The remaining four beds operate in the same manner.
Safety Evaluation Summary Adding the solenoid valves to these circuits did not affect the system performance in a manner that could lead to an accident. Failure of a solenoid valve or the associated circuit could shut down a Condensate Demineralizer bed, but as long as water quality is maintained, the design basis is met. This activity did not increase the probability of an accident previously evaluated in the SAR. The Condensate Demineralizer System is not required for the mitigation of any accidents defined in the NSOA and UFSAR. This change does not prevent the Condensate Demineralizer System from performing its design basis function as defined in the SAR.
l System flow and water quality can be maintained through the remaining ;
four Demineralizers in service or through the bypass, if one of the solenoid valves fail. Therefore, this change did not prevent nor degrade any essential safety function assumed by the NSOA to mitigate the consequences of design basis accidents. This activity did not increase the consequences of an accident evaluated previously in the SAR. Single I failure proof criteria does not apply to the Condensate Demineralizer System, as determined in the Design Safety Standards. The affected l i
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'I circuits do not perform a safety function, nor do they communicate with equipment important to safety. The probability of occurrence of a l malfunction of equipment important to safety evaluated previously in the SAR was not increased. This activity did not increase the consequences of -
a malfunction of equipment important to safety evaluated previously in the !
SAR. The possibility of an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR was not created. Technical Specifications identify the surveillance requirements and chemistry limits l
for the Reactor Coolant System. The monitored parameters include conductivity, chloride concentration and pH. The affected circuitry will !
not impact the monitored parameters. This revision did not change or affect any setpoints or surveillance requirements. This activity did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 98-58 Fuel Pool Cooling and Cleanup System P&lD Revision Description and Basis For Channe The P&lD for the Fuel Pool Cooling and Cleanup System was in error in l
that the root isolation valve for a level indicator was shown normally closed, instead of normally open. The Operating Instruction (OI) shows '
the valve normally open. A search of historical documents revealed that this valve was shown closed on the P&ID even at the time of the original plant startup. This valve needs to be in the open position for the level indicator to function. The P&lD has been revised to show the valve as normally open.
Safety Evaluation Summary Keeping the valve normally open does not affect the system in any way, because the line associated with it is only an instrument line having no fluid flow through it. Also, it is designed per applicable codes for both normally open or normally closed positions. This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. The level indicator is provided to monitor reactor well water level during refueling. The indicator is mounted on the fuel pool pump rack which controls flow to or from the reactor well during refueling. In order for the level indicator to function, its isolation valve needs to be open.
The valve is on a 3/4 inch instrument line, off the main line, and any postulated failure of this valve will not have any significant impact on the system. This activity did not increase the consequences of an accident evaluated previously in the SAR. The valve does not perform any safety 183
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function. Changing the valve position from normally closed to normally open does not lead to any malfunction. Any postulated malfunction of this valve does not impact any equipment important to safety, because it is not relied upon to do any automatic action or for any operator response during an accident. This change did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR, and it did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. This change did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created. There are no surveillance tests associated with the subject instrument. No Technical Specification requirements are given for the reactor well cavity level. This change did not reduce the margin of safety as defined in the basis for any Technical Specifications.
SE 98-59 Standby Gas Treatment System Air Flow Diagram P&ID Revision Description and Basis For Chance The power supplies for Differential Pressure Transmitters and Flow Transmitters were not shown on the Standby Gas Treatment System Air Flow P&lD. Identification num'oers for the SBGT Electric Coil Differential Temperature Indicating Controllers and Electric Coil Differential Temperature Transmitters were not properly depicted on the P&lD. The P&ID has been updated with this information.
Safety Evaluation Summary This activity did not impact single failure criteria and the SBGT system operation was not affected. Correcting equipment identification numbers did not impact the operation of any plant equipment. Because these activities did not affect the overall system performance in a manner that could lead to an accident, the probability of an accident previously evaluated in the SAR was not increased. The SBGT system still meets the operational requirements in the UFSAR. This activity did not increase the consequences of an accident evaluated in the SAR. No physical change to the plant was made. The changes performed were to maintain consistency between various plant documentation. The SBGT system is a single failure proof system. The document revisions did not affect the SBGT system and its ability to meet single failure criteria. Since the revised circuitry still meets the original design basis and electrical separation requirements, the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased.
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This activity did not increase the consequences of a malfunction of f
equipment important to safety evaluated previously in the SAR. The SBGT system is required to establish secondary containment integrity during a fuel handling accident, Loss-of-Coolant-Accident (LOCA) inside primary containment and LOCA outside primary containment. This change does not prevent the SBGT System from performing this design function. No failure can be postulated by this change that would create the possibility of an accident of a ditTerent type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a difTerent type than any previously evaluated in the SAR was ,
not created. The document revisions performed did not change or affect any setpoints or surveillance requirements as specified in the DAEC Technical Specification. The SBGT logic remains as it was previously.
l Therefore, this change did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 98-69 Drywell Cooling Water System P&ID Revision Description and Basis For Chance An inconsistency was identified in the depiction of the logic for the 1VCC007A/B Drywell Coolers on the Drywell Cooling Water System P&lD. It was also identified that these two coolers did not have high temperature alarms as referenced in the UFSAR. The UFSAR was revised to remove the references that indicated the coolers have high temperature alarms for air and water temperatures. The P&ID was revised to show the 7A cooler temperature elements going to the ' A' Drywell Cooling System Air / Water Loop Temperature Recorder, the 7A cooler temperature switches going to the loop 'A' controis, the 7B temperature elements going to the 'B' Drywell Cooling System Air / Water Loop Temperature Recorder, and the 7B cooler temperature switches going to the loop 'B' controls. A note was also added to the P&ID to indicate that the 7A/B coolers do not alum on high temperature. These changes reflect current plant configuration.
Safety Evaluation Summary The Drywell Coolers are required for normal power operation only. They are not required for the mitigation of an accident. The probability of an occurrence of a malfunction of equipment important to safety was not increased, nor were the consequences of a malfunction of equipment l important to safety increased. Removing the high temperature alarms from these Drywell Coolers and changing the logics supplied by the temperature switches did not adversely affect the Primary Containment System. The probability of an accident or the consequences of an accident l 185 l
l 1
were not increased. Adequate indication is available to monitor Drywell temperatures. The possibility of an accident of a different type than previously evaluated in the UFSAR was not created. Operator actions were not affected. The possibility of a malfunction of equipment important to safety of a different type than that previously evaluated in the UFSAR was not created. The requirements in the Technical Specifications were not impacted. The margin of safety as def'med in the basis for the Technical Specification was not reduced. Removing the associated alarms did not adversely affect safety.
SE 98-70 Drawing Changes For Removal OfIfigh Pressure Coolant Injection and Reactor Core Isolation Cooling Turbine Exhaust Vacuum !
Breakers !
I Descrintion and Basis For Chance '
The High Pressure Coolant Injection (HPCI) System and Reactor Core Isolation Cooling (RCIC) System turbine exhaust vacuum breakers in the Torus, depicted on the HPCI and RCIC System P& ids as well as other design drawings, are no longer present. The current installation consists of a pipe nipple, manual isolation valve, and a pipe cap. This configuration was field verified during Refuel Outage 15. The vacuum breakers have been removed from the drawings.
Although no record could be located for the removal of these valves, they were listed in the equipment database as abandoned in place and the associated Q200s stated that these vacuum breakers were modified during original plant construction and are no longer able to perform a vacuum breaking function. These valves are no longer installed in the plant and are no longer required to perform any safety function. The currently installed isolation valves are both Quality Level 1 and are capable of performing the required safety function, isolation / pressure boundary.
Added assurance is provided by the existence of pipe caps.
Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR, nor were the consequences of an accident increased. The HPCI and RCIC Systems are not accident initiators, but are required to mitigate the consequences of abnormal operational transients, accidents, and other events as described the NSOA.
The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased, nor were the consequences of such a failure. The current configuration inside the Torus (manual isolation valves and pipe caps) will maintain pressure integrity 186
l l
l l and prevent HPCI/RCIC turbine exhaust steam from entering the Torus air space through the abandoned 1 inch lines. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to ,
safety of a different type than any evaluated previously in the SAR was !
not created. The design, function, and method of performing the function !
of the affected systems remains the same as described in the UFSAR. The l
l ability of the HPCI and RCIC systems to perform their intended safety l actions was not impacted by this activity. There were no increased I 4
l radiological consequences associated with this activity, and no new failure
{
modes were introduced. The margin of safety as dermed in the basis for any Technical Specification was not reduced. !
l SE 98-71 Technical Requirements Manual Revision To Relax Requirement For Logic System Functional Test For Engineered Safeguards Compartment Cooling and Ventilation Systems Descrintion and Basis For Chance Technical Specification Amendment 223 to convert the DAEC Technical Specifications (CTS) to the Improved Technical Specifications (ITS) per NUREG-1433 included relocation of certain CTS requirements to the ;
Technical Requirements Manual (TRM). The TRM is a licensee-controlled document incorporated by reference into the UFSAR and subject to revision per 10 CFR 50.59.
This change revised the TRM to relax the surveillance frequency of the Logic System Functional Test (LSFT) for the Engineered Safeguards (ES)
Compartment Cooling and Ventilation Systems from 18 to 24 months.
The resulting requirements continue to ensure that process variables and SSCs are maintained consistent with the conditions and assumptions of the DBA and transient analyses. The change did not involve any design changes to the plant and is consistent with current plant operations.
Safety Evaluation Summary This change did not result in any hardware changes or permanent physical alteration of the plant and the change had no impact on the design or l
function of any SSC. The SSC process variables, characteristics, and functional performance were maintained consistent with the event initiator and initial condition assumptions for the DBA and transient analyses. The likelihood of event initiation remains as previously analyzed and there was
- no increase in the probability of occurrence of an accident evaluated previously in the SAR. These requirements did not satisfy the 10 CFR 50.36 screening criteria for inclusion in Technical Specifications and they 187
l do not impact, either directly or indirectly, any event analyzed in the SAR.
This activity did not change that conclusion as this change did not introduce any new modes of equipment operation. This change did not increase the consequences of an accident evaluated previously in the SAR.
The performance history of the LSFT for the ES Compartment Cooling l and Ventilation Systems was reviewed. While individual instruments were found outside of the recommended tolerances, no failures of the LSFT were identified. In addition, the coolers are functionally tested along with their associated pumps. A performance review of the annual Simulated Automatic Actuation tests did not identify any failures of the room coolers to start on demand. It was, therefore, concluded that this
! change would not adversely affect any plant equipment and there would be l no increase in the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The indication function of these pressure switches is not part of the primary success path of any analyzed event. Thus, there was no increase in the radiological consequences of an accident. This change did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. No new credible accidents were introduced. This change did not create the possibility of an accident of a different type than any evaluated previously in the SAR. The possibility of a malfunction of equipment important to safety of a different type than any evaluated l previously in the SAR was not created, and the margin of safety as defined in the basis for any Technical Specification was not reduced.
l SE 98-72 Technical Requirements Manual Revision To Delete Requirement For l Shiftly Channel Check Of Control Rod Position Description and Basis For Chance Technical Specification Amendment 223 to convert the DAEC Technical Specifications -(CTS) to the improved Technical Specifications (ITS) per NUREG-1433 included relocation of certain CTS requirements to the Technical Requirements Manual (TRM). The TRM is a licensee-controlled document incorporated by reference into the UFSAR and subject to revision per 10 CFR 50.59.
j This change revised the TRM to delete the surveillance requirement for
! the shiftly Channel Check of Control Rod Position Indication. The resulting requirements continue to ensure that process variables and SSCs are maintained consistent with the conditions and assumptions of the DBA and transient analyses. This change did not involve any design changes to the plant and is consistent with current plant operations.
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Safety Evaluation Summary This change did not result in any hardware changes or permanent physical alteration of the plant and the changes would have no impact on the design or function of any SSC, as multiple means are available for determining control rod position. Because the requirement for verifying Control Rod position did not change, the deletion of the shiftly channel check of the position indication did not have any impact on rod pattern control requirements. The SSC process variables, characteristics, and functional performance are maintained consistent with the event initiator and initial condition assumptions for the DBA and transient analyses. Therefore, the likelihood of event initiation remains as previously analyzed and there is no increase in the probability of occurrence of an accident evaluated previously in the SAR. Because these requirements did not satisfy the 10 CFR 50.36 screening criteria for inclusion in the Technical Specifications, they did not impact, either directly or indirectly, any event analyzed in the SAR. This activity did not change that conclusion as this change did not introduce any new modes of equipment operation. Specifically, control rod position is still required to be known. This change did not increase the consequences of an accident evaluated previously in the SAR, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR, and the possibility of an accident of a different type than any evaluated previously in the SAR was not created. This change did not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. Since the requirement for verifying Control Rod position has not been changed, the deletion of the shiftly channel check of the position indication did not have any impact on rod pattern control requirements. Therefore, this change did not reduce the margin of safety as defined in the basis for any Technical Specification.
SE 98-74 Change In Fire Protection Commitments Descrintion and Basis For Chance Changes were made which differ from commitments made in the NRC Fire Protection Safety Evaluation Report (SER) for the DAEC, dated June 1,1978. The following two items represent the scope of the Fire j Protection Program commitment changes. These changes did not affect j the UFSAR or the DAEC Fire Plan.
. Electrical supervision of a closed pair of 3-hour rated, swinging fire doors provide the required assurance that the door is closed in the 189 l
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,. event of a fire. The previous commitment was to provide an electro / thermal link actuated by smoke detectors in place'of a fusible link for a 3-hour rated, roll-up style fire door located in the fire barrier between the two divisional essential switchgear rooms.
e Mechanical ventilation (for exhaust only) is provided in the Standby _
Diesel Generator Day Tank Rooms rather than louvers at floor and ceiling as previously committed.
Safety Evaluation Summary Fire is not an entry condition, basis or an assumption for any accident previously evaluated in the UFSAR and NSOA. These changes did not i impact any of these analyses including their entry conditions, bases or
( assumptions. Changes to commitments with respect to fire door L supervision and Diesel Generator Day Tank Room ventilation do not change the probability of an accident, as fire is not an entry condition, basis or assumption for any of these events. The changes did not increase l the probability of any of the following events occurring: a firt. ;
inadvertent actuation of a suppression system, or loss of essentini HVAC I or other plant equipment credited in the DAEC safe shutdown analysis.
!~ These changes did not adversely affect the plant's ability to safely 1 l shutdown in the event of a fire. These changes did not increase the probability of occurrence of an accident evaluated previously in the SAR, i and they did not challenge the integrity of any fission product barriers. I The radiological consequences of any accident evaluated previously in the SAR were not increased. The ability of fire protection systems and features to perform their intended functions was not affected. These changes did not adversely affect procedures governing maintenance and modification activities for fire protection systems and features. The consequences of an accident evaluated previously in the SAR were not i increased, and the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR was not increased.
Since these commitment changes did not adversely affect the ability of either safety related equipment or fire protection systems and features to perform their functions, the consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased.
The possibility of an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a l malfunction of equipment important to safety of a different type than any l evaluated previously in the SAR was not created. Fire door supervision
- and Diesel Generator Day Tank Room ventilation are not addressed by Technical Specifications. These changes did not impact the margin of L safety as defined in the basis for any Technical Specification.
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SE 98-81 Residual Heat Removal (RHR) System P&ID Revision 1
Description and Basis For Chance '
The P&lD for the RHR System was revised to show the modification performed on the RHR Loop 'B' LPCI Inboard Injection Isolation Valve.
This change pertains to a modification of the valve to preclude the potential for valve inoperability caused by pressure locking and thermal l binding (PL/FB) as recommended in NUREG 1275, Volume 9. The l modification consisted ofinstalling a 1/2 inch pipe between the bottom l drain pipe on the valve and the side test connection pipe from the upstream l l side. Previously, an EMA for this modification was performed and the l
l isometric drawing was updated but the P&lD was not. !
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Safety Evaluation Summary l Similar modifications for the valves susceptible to pressure locking and
! thermal binding have been implemented throughout the industry. This l valve does not initiate any transient or accident. Therefore this activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. This change did not compromise the safety function of the valve in any way, and the system still functions as
, intended. This activity did not increase the consequences of an accident l evaluated previously in the SAR. By reducing the potential fbr valve l inoperability caused by pressure locking and thermal binding, this modification increases the overall system reliability and availability. The piping stresses and support loads were found to be within the code allowable limits. This activity did not increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the SAR. The containment isolation function of the valve to prevent the leakage of radionuclides from the containment is assured by the seal of the RHR-side (away from the reactor) seat ring. This change is an acceptable method to alleviate the PLffB concern and is widely used in the nuclear industry, per NUREG 1275, Volume 9. This activity did not increase the consequences of a malfunction of equipment important to safety evaluated previously in the SAR. The possibility of an accident of a different type than any evaluated previously in the SAR was not created, and this activity did not create the possibility of a malfunction of i equipment important to safety of a different type than any evaluated l previously in the SAR. No safety margins, safety settings or safety limits are defined in the Technical Specifications for the subject valve, so no
, margin of safety as defined in the basis for any Technical Specification was reduced.
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l SE 98-85 UFSAR Revisions Due To UFSAR Improvement Project Description and Basis For Chance The UFSAR Improvement Project identified changes to be made to the ,
UFSAR. These changes included updating the River Water Supply (RWS) pump minimum submergence value to 2 feet 7 inches from 3 feet 0 inches and correcting the description of the operation of the radwaste dilution valves to state that they receive an automatic closure signal from a "drywell high pressure or low reactor water level" and not from a " loss of offsite power". These changes to the UFSAR did not change plant equipment or procedures and did not alter the operation of the RWS system.
Safety Evaluation Summary The change to the submergence level did not change the operation of the RWS pumps. The operation of the pumps per DAEC Technical Specifications require that the level be maintained approximately 3 feet above the new minimum submergence level. Therefore the change to the UFSAR did not affect pump operations. The pumps' suctions still have the same height difference between the RWS pit floor and the pumps. The difference of approximately 12 inches has not changed and thus the silt that may be drawn through the system is the same as before this change to the UFSAR. With the increase in operating range for the RWS pumps, the probability and the consequence of an accident previously evaluated have been reduced. Also with the pumps having a greater range, the probability and consequences of an equipment failure have not been increased. The RWS System still supplies all of the required water flow to the safety loads and therefore this change did not create the possibility of a malfunction of plant equipment. The RWS System still operates svith a loss of offsite power. The radwaste dilution valves do not get a closure signal from a loss of offsite power, but may be closed due to non-safety related solenoids losing control power. With the loss of power to the non-safety related solenoids, the air is vented from the operators for the radwaste dilution valves and these valves fail closed on a loss of air.
Ilowever,if the non-safety related solenoid valves fail to vent the operators, the radwaste dilution valves obtain a close signal that deenergizes safety related solenoids on Drywell high pressure or low reactor water level or low level in the RIIRSW and RWS pits. The RWS System supplies water to the safety related loads under all previously evaluated accidents in the SAR. These corrected signals ensure that water is automatically supplied when required and therefore ensures that the probability and consequences of an accident as evaluated previously in the SAR are not increased. The corrections also ensure that equipment 192
l important to safety will operate and the safety related sub-systems will get cooling water to reduce the consequence of an equipment malfunction.
The correction shows the reader that the plant has three direct signals that
- control the radwaste dilution valves - Drywell high pressure, low reactor water level, and low level in the RHRSW and RWS pits. The possibility of an accident of a different type than any evaluated previously in the SAR l was not created, and the possibility of a malfunction of equipment j important to safety of a different type than any evaluated previously in the i SAR was not created. The margin of safety as defined in the basis for any
! Technical Specification was not reduced.
SE 98-87 UFSAR Clarifications and Corrections For Standby Gas Treatment l System l
l Description and Basis For Chance
! The UFS AR has been revised as follows to correspond to existing system conditions for the Standby Gas Treatment (SBGT) System. 1 The UFSAR described the SBGT charcoal filters as containing l i '
approximately 1500 lbs of charcoal Another UFSAR section stated that each SBGT train contains 1240 lbs of charcoal. The adsorbers actually contain 1224 lbs of net effective charcoal,2500 lbs. total. By changing the weight of the charcoal, the value of the specific loading is also ;
affected. The resultant specific loading based on a charcoal weight of ;
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1224 lbs. is 3.4 mg ofiodine per gram of charcoal.
The UFSAR described the HEPA filters as Cambridge Type 1E 1000 with galvanized steel frames. The filters are Cambridge Type 1E 1000, however, the frames are cadmium plated. Reference to the specific model l number was deleted.
l The UFSAR was also revised to describe the current testing requirements for new charcoal.
- Safety Evaluation Summary This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR. No physical changes to the plant were made. SBGT continues to meet the design, material, and construction standards applicable to the system. The SBGT System does not cause accidents as previously analyzed in the SAR. Changing the weight of the charcoal and specific loading does not adversely affect the efficiency of i
the charcoal. The charcoal adsorbers efficiency of 99% for removal of
- elemental iodine and methyl iodine will be achieved. The particulate filters efficiency of 99.97% for particles greater than 0.3 m will be met.
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The probability of a malfunction was not changed, and the consequences l of a malfunction of equipment important to safety previously evaluated in l-the SAR were not increased. This activity did not create the possibility of an accident of a different type than any evaluated previously in the SAR, and the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR was not created.
L, These changes did not affect the margin of safety as defined in the basis !
for any Technical Specification. The SBGT System does not operate differently as a result of these changes.
SE 98-92 UFSAR Revisions For Compressed Air Systems Description and Basis For Chance Changes to the UFSAR were needed to define the systems / components served by the' safety related air compressors, clarify some information pertaining to the use of accumulators on the Containment Purge Valves, and for overall enhancement to the compressed air section. Information ;
about the Safety Related Air System and the systems and components it supports during accident conditions was added to support a statement in the UFSAR which states that a loss of the normal plant Instrument Air System will not affect safety related equipment. Reference to the existence of air accumulators that serve as a reliabie back-up to the Containment Primary Isolation Vent and Purge Valves, the Reactor Building to Torus Vacuum Breaker Isolation Valves, and the Drywell Cooling Water Valves, has been removed. A loss of the normal Instrument Air System will not adversely affect safety related equipment -
and accumulators are not required for the Containment Isolation Valves to function because:
(1) of the existence of the safety related Air System which meets single failure proof criteria, (2) testing has been done to ensure that safety related components supplied exclusively by the normal Instrument Air System fail in the safe position and, (3) loads supplied by the Safety Related Air System have redundancies inherent to their design such that a failure of one train of the Safety Related Air System will not prevent them from performing their safety function.
l= Safety Evaluation Summary i
These changes were intended to enhance the information contained in the compressed air section of the UFSAR. The discrepancies that were Lc 194
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i identined in the UFSAR (i.e., the omission of accumulators from the
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instrument air lines which supply control valves) did not increase the probability of occurrence of an accident evaluated previously in the SAR.
The affected components, Containment Purge and Vent Isolation Valves,
- Reactor Building to Torus Vacuum Breaker Isolation Valves and the j ' Drywell Cooling Water Valves, are not accident initiators and therefore i
the removal of accumulators (or not taking credit for existing ones) did not increase the probability of occurrence of an accident. The Safety Related l Air System is a reliable single failure proof system which is fully capable i of supplying air to the Containment Purge and Vent Isolation Valves, ll Reactor Building to Torus Vacuum Breaker Isolation Valves and the 4
Drywell Cooling Water Valves. It is not necessary for the valves to have point of use accumulators for the actuators or T-seals, nor is it necessary to take credit for existing accumulators for a back-up supply of compressed
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air. These changes did not increase the consequences of an accident evaluated previously in the SAR. Since air quality and header pressure were not altered by this activity, the probability of occurrence of a I malfunction of equipment important to safety evaluated previously in the
- j. SAR was not increased. The consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not increased, j and the possibility of an accident of a different type than any previously
- evaluated in the SAR was not created. The only malfunctions which can l occur as a result of the instrument air supply to the'affected components
- are a result of poor air quality or insuf6cient air pressure. This activity,
- did not alter either aspect of the Instrument Air System. This activity did ret create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the SAR. The Safety Related Air System has been deemed capable of supporting major post-accident loads. Therefore, the performance of the affected valves will not deteriorate as a result of this activity. This activity did not reduce the margin of safety as denned in the basis for any Technical Specincation.
SE 98-% UFSAR Change Concerning Residual Heat Removal Service Water (RHRSW) System Descrintion and Basis For Chance A change to the UFSAR section for RHRSW involved removing the statement,"Thu extends the life of the RHR heat exchangers by providing a better demineralized water layup system." The sentence previous to this sentence in the UFS AR stated, " Piping and valving exist to supply 35 psig pressure to the demineralized water flush lines to the RHR heat exchangers," which is accurate. However, a Layup System is not used. ,
The statement that,"This extends the life of the RHR heat exchangers by 195
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providing a better demineralized water layup system" implied that the Layup System is used when, in fact, it is not.
Safety Evaluation Summary The probability of occurrence of an accident previously evaluated in the SAR was not increased because the RIIRSW System is not an accident initiator (nor could it affect any other system / structure / component which is l
an accident initiator). The consequences of an accident previously evaluated in the SAR were not increased because the R11RSW System still
! functions as assumed in the SAR due to the RilR Heat Exchanger testing L and inspections which are performed on a regular basis and the fact that the RHRSW System is included in the " Service Water Corrosion Monitoring Program" which insures intact RiiRSW piping and j components. The probability of occurrence of a malfunction of equipment l important to safety previously evaluated in the SAR was not increased.
l The RIIRSW System equipment was not adversely affected beyond any
! malfunctions evaluated previously in the SAR and the system maintains its ability to perform its safety functions. The consequences of a malfunction i
of equipment importa".t to safety previously evaluated in the SAR were not increased. The possibility of an accident of a different type than any evaluated previously in the SAR was not created, and the possibility of a
- malfunction of equipment important to safety of a different type than any l evaluated previously in the SAR was not created. The margin of safety was not reduced because, based on a review of the Technical Specifications, Technical Specification Bases, UFSAR and NSOA, there is i no defined basis for a margin of safety for any layup provision or l corrosion-induced failure of the RilRSW piping and components. ;
SE 98-104 UFSAR Revision Concerning Responsibilities L Descriotion and Basis For Chance l I Revisions were made to procedures that included descriptions in the SAR that define how an activity is performed. These changes involved revisions to responsibilities that were limited to the description of the l Duane Arnold Energy Center (DAEC) programs listed in Chapter 13.1 of the USFAR. No systems or components were affected by this change. !
The following changes were incorporated into the UFSAR:
- The Executive Vice-President Corporate Services reports to the
! President and Chief Executive Officer Interstate Energy Corporation.
i-j e The Executive Committee no longer exists. _
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l e The Corporate Supply Chain is no longer providing support to DAEC.
- Duties of the Scheduling Supervisor were not previously identified.
. Responsibilities for nuclear fuel reload design, previously the responsibility of the Business Unit, moved to Operations.
. Reporting chain of the Records and Microfilm Administrator was revised to reflect the current corporate organizational structure.
l e Civil / Facilities Support Group was not previously identified bi the
! composition of Maintenance Process Support Group.
- Environmental Supervisor position no longer exists.
- Chemistry Supervisor now has additional responsibilities.
. Radwaste Supervisor now has fewer responsibilities identified.
- Decontamination Supervisor position no longer exists.
. ALARA Supervisor position no longer exists.
. Technical Specification Section referenced as the Plant Manager's Standard of Qualification no longer exists.
- The position of" Training Supervisor-Simulator" no longer exists.
Safety Evaluation Summary l
The changes involved responsibilities performed by groups or individuals of equal or higher position. As a result the changes in responsibility did not increase the probability of occurrence of" personnel errors" as initiating or contributing actions to the accidents evaluated in the UFSAR.
This activity did not increase the probability of occurrence of an accident evaluated previously in the SAR, and the consequences of an accident evaluated previously in the SAR were not increased. These changes did not alter the scope or reduce effectiveness of the programs. Accidents in
- the UFSAR were not affected and the resulting consequences of the
- accidents were not changed. These changes did not affect any system, structure or component important to safety evaluated in the UFSAR. The L changes did not increase the probability of occurrence of any related equipment malfunctions, and the consequences of a malfunction of equipment important to safety evaluated previously in the SAR were not L increased. The changes did not create the possibility of an accident of a
! different type, and the changes did not create the possibility of a
._ malfunction of equipment important to safety of a different type. The 197 l
l l margin of safety denned in the basis for any Technical Specincation was .:
not reduced. l i
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1 0 t Section C - Experiments -
l This section has been prepared in accordance with the requirements of 10 CFR Section 50.59(b). No experiments were conducted during the period beginning March 1,1997 4
and ending September 30,1998.
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l Section D - Fire Plan Changes The information contained in this section identifies, briefly describes and provides assurance that changes made to the DAEC Fire Plan during the period beginning March 1,1997 ending September 30,1998 did not alter our commitment to the NRC guidelines l contained in " Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance."
Volume I- Revision 33 Changes were required to the'DAEC Fire Plan as a result of the 10 CFR 50 Appendix R
- re-baseline analysis. The majority of the changes to the Fire Plan were needed to make the Plan better reflect the compliance strategies to meet 10 CFR 50 Appendix R requirements. The following changes were made to Volume I of the DAEC Fire Plan resulting in Revision 33 i
. Changed the scope of the Fire Plan operability and surveillance requirements to
- address only those Fire Protection Systems and features required for compliance 3 with 10 CFR 50 Appendix R.' Therefore, the operability and surveillance j requirements pertaining to various detection, suppression and hose station systems j that are not required for 10 CFR 50 Appendix R compliance have been removed, !
Fire Protection Systems required for 10 CFR 50 Appendix R compliance have ;
been added, and the Bases section of the Fire Plan Operability Requirements have i been revised accordingly.
. Changed the plant's carbon dioxide tank volume from a 90% minimum level to an 80% minimum level.
. Removed the surveillance requirement to demonstrate redundant fire pump operability on a daily basis when one fire pump is considered inoperable.
. Included fire dampers in the list of fire rated assemblies required for Appendix R that have operability and surveillance requirements.
. Revised the Firewatch requirements to allow the Fire Marshal to evaluate whether a fire watch is required or other measures may be used if the fire watch is required to enter a high-radiation area, a contaminated area or an area of the DAEC where ;
the fire watchers presence could endanger the safe operation of the DAEC or where a personnel safety hazard is likely to exist.
. In the Operability Requirements, included liose Station 36 as a compensatory measure for inoperable Cable Spreading Room suppression. This liose Station, as well as Station 35, support compliance with the NRC Safety Evaluation Report '
dated June 1,1978. 4 1
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- Included Administrative Control Procedure," Addressing Design and 1 Modification Controls", among the administrative control procedures which play l
a role in fire prevention at DAEC.
1 e Documented the plant's common practice of not isolating the Cable Spreading Room CO2Fire Suppression System when security door checks in this area are made. Such door checks are very brief and require minimal entry into the Cable l Spreading Room, thereby posing no significant safety concern.
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e included semi-annual operability checks on the CO2 thermal detectors in the Cable Spreading Room among verifications to be performed to demonstrate operability of the room's CO 2System. This was added to reflect the automatic initiation capability of this suppression system.
i e Added to the qualifications listed for the DAEC Fire Marshal and Fire Brigade Instructor positions. The additional criteria referencing specific NFPA codes simply describes how the DAEC has chosen to meet the regulatory requirements and does not represent a commitment to the NRC.
- Relocated fire brigade and security related infor. nation to Volume 11 of the DAEC Fire Plan and titled Volume 11 " Fire Brigade Organization." Volume I of the Fire Plan has been titled " Program" to distinguish it from Volume II. Volume I still ;
includes general requirements for the brigade.
- Used the terms " safe shutdown" rather than " safety related" throughout the document when referring to systems, equipment and features that are credited for reaching safe plant shutdown conditions. Both safety related and non-safety related systems and equipment are credited in the fire protection safe shutdown analysis.
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- Corrected the Fire Plan's wording under the surveillance requirements section for fire doors to match that contained in the UFSAR. The UFSAR change involved an exception for locked fire doors accessing cable and pipe chases and the steam tunnel from the semi-annual surveillance requirement to verify operation of auto-closing devices and latch mechanisms.
- Made various editorial changes to further clarify the information contained in the Fire Plan, including a new section entitled " Operability Considerations" which addresses fire protection equipment installed to meet non-regulatory as well as regulatory requirements or commitments.
- Revised the " Operability Requirements" section to focus on Appendix R required Fire Protection Systems and features. The basis section was also reorganized.
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. . Changed titles ofindividuals and department names and made organizational L chart changes to reflect the current organization at the DAEC. The primary ,
change is that the Supervisor Fire Protection position was clinkated and the bulk l of the program responsibilities are assigned to the Long Terrr. Program Team Leader in engineering.
l e General Fire Plan document layout changes an J reference additions.
r l l . Included clarification ofinsurance reporting requirements.
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Volume I - Revision 34 L e Revised " Quality Assurance" section, to focus on those systems and features ;
L listed in the " Operability Requirements" section.
i e Revised " Operability Requirements" section as follows:
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- a. Changed requirements and associated basis for raceway fire rated wrap inspection to 100% every five years. This was accomplished by deleting the 35% every cycle language that is included for other barrier items such as penetration seals. The change was made because there is only one such ,
installation left in the plant.
- b. 'Added semi-annual detection test for Deluge 18. This was added to i surveillance requirements for completeness. The test was already being performed.
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- c. Other minor corrections were incorporated and references added.
I' Volume II - Revision 31
- Moved specific brigade information into this volume, j e . Revised process for revising Area Fire Plans (pre-fire plans) that allows individual revision / approval for added flexibility. Area Fire Plans are now controlled by the Fire Marshal.
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.- Miscellaneous minor Fire Protection Program updates.
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l Section E - Commitment Changes The information contained in this section identifies and briefly describes various commitment changes which were made during the period beginning March 1,1997 and ending September 30,1998. The changes described were evaluated and are being i
reported per the Nuclear Energy Institute's " Guideline For Managing NRC
! Commitments", dated December 19,1995.
0 l AR 971012.00 In Licensee Event Report 88-016," Engineered Safety Feature Actuation and Loss of an Essential Bus Due to Inadequate Degraded Voltage Detection Logic Design", dated November 29, 1988, a corrective action is listed to identify all affected meters
, associated with the DC Distribution Systere to ensure that all were l included in the Preventive Maintenance Program for calibration to t
ensure proper indication of the monitored parameter. This l
commitment has been revised to delete the requirement for l periodic calibration of 48 VDC meters. The 48 VDC System has l l two battery chargers operating in parallel and one source of power is diesel generator backed to ensure reliability. The 48 VDC l l System supplies power to the non-safety related plant telephone and Cardox System ammeters. These meters provide indication of equipment status only. Any identified meter inaccuracy or failure is addressed by a maintenance request and subsequent maintenance activities.
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AR 971020.00 In NG-93-4446," Reply to a Notice of Violation Transmitted with Inspection Report 93014", dated October 29,1993, a corrective l action is listed to enhance the dress-out portion of General l Employee Training (GET) to enhance the awareness of High
!- Radiation Areas. This commitment has been deleted. The use of other instructional settings (classroom and computer based instruction) have replaced the practical dress out evaluation.
Employee awareness training of high radiation entry requirements effectively occurs through these other instructional settings.
- Instructional material will continue to emphasize awareness of
( High Radiation Areas, only the practical dressout evaluation has
[ been deleted.
l AR 971798.00 As per letter," Modification Of Feedwater And Control Rod Drive Return Line Nozzles Inspection Intervals", dated December 16, 1986, from D. Muller, NRC, to L. Liu, Iowa Electric Light and Power Company, DAEC committed to performing an ultrasonic (UT) examination once every cycle of a 2.5 foot stainless steel section of the Control Rod Drive Return Line (CRDRL) which 203 i
contains stagnant water. NUREG 0619 requires that during each refueling outage, that portion of CRDRL containing stagnant water must be inspected in accordance with recommendations of NUREG-0313 Revision 1. NUREG-0313 Revision I contains provisions to inspect less frequently if no indications are found.
This examination will be performed in accordance with the recommendations of NUREG-0313 Revision 1.
AR 972361.00 NG-85-3307," Response to NRC Inspection Report 85-12", dated July 7,1985, lists a corrective action to institute a dual lock system, under operations and health physics control, for all areas having potential dose rates in excess of 10 R/hr. Improved Locked Ifigh Radiation Area key controls,"look ahead" Health Physics planning, and improved personnel dosimetry have eliminated the need for dual lock control of areas in excess of 10 R/hr. Therefore, the dual lock system has been eliminated.
AR 972362.00 In NG-85-3307," Response to NRC Inspection Report 85-12",
dated July 7,1985, as the result of an identified weakness, a commitment was made to institute administrative controls which required the Shill Supervisor to be formally notified of the completion of the Health Physics technician's briefing to workers prior to the Shift Supervisor issuing a key to a Locked High Radiation Area. Keys to Locked Iligh Radiation Areas are now controlled by the Health Physics Department and are not issued to workers until the llealth Physics technician's briefing is completed. There is no longer a need to notify the Shill Supervisor of completion of the briefing.
AR 972471.00 In the " Station Blackout Rule Conformance Evaluation" letter (
from C. Shiraki NRC, to L. Liu, Iowa Electric Light and Power Company, dated June 15,1992, DAEC was committed to NUMARC initiative Sa, which includes accelerated testing of the Emergency Diesel Generators (EDGs). A change was made to remove accelerated testing requirements from the EDG Reliability Program as allowed per Generic Letter 94-01," Removal Of Accelerated Testing And Special Reporting Requirements For Emergency Diesel Generators". The EDG Reliability Program has been shifled to the Maintenance Rule Program (10 CFR 50.65).
The Maintenance Rule performance criteria were developed utilizing the work completed for the Station Blackout Rule. The Maintenance Rule Program maintains the connection between the EDG Reliability Program and Regulatory Guide 1.155.
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AR 982371.00 In LDR-82-256, from L. Root, Iowa Electric Light and Power Company, to J. Keppler, NRC, dated September 9,1982,
" Response to Inspection Report 32-07", DAEC committed to perform Main Steam Isolation Valve (MSIV) closure time testing during startup from shutdowns that involve closing the MSIVs.
This commitment was made to detect future MSIV actuator problems. The actuators have since been replaced with larger actuators. Closure time testing of the MSIVs is now performed in accordance with the Inservice Testing Program and Technical Specifications.
. AR 982563.00 LER 92-002, dated January 26.1992, described an event in which the Turbine Building Sump Room fire door was blocked open without the required Technical Specification compensatory action of a fire watch. In this event a surveillance was being performed which required the Turbine Building Sump Room ventilation fans to be secured. The control panel containing the handswitches for the ventilation fans cautioned the operator to block open the -
Turbine Building Sump Room door to preclude a hydrogen buildup with the ventilation secured. Corrective action required that applicable surveillance procedures, system operating instructions, annunciator response procedures, and the Turbine .
Building Sump Room ventilation fans' control panel caution placard be changed to reflect the requirement fbr a fire watch when the fans are secured and the Turbine Building Sump Room door is blocked open. Since this incident, the condition causing hydrogen 4 buildup when the ventilation fans are secured no longer exists, and the DAEC Fire Plan has been revised such that this Turbine t Building Sump Room door is no longer considered a fire door. A
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fire watch is no longer required when this door is blocked open.
The procedural requirements, operating instructions, and fan '
control panel placard requiring a Firewatch when the fans are secured and the Turbine Building Sump Room door is blocked open are no longer necessary, c
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