ML20207K976
| ML20207K976 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/31/1998 |
| From: | WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20207K970 | List: |
| References | |
| 14, NUDOCS 9903170326 | |
| Download: ML20207K976 (160) | |
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j Attachment I to ET 99-0003
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WOLF CREEK NUCLEAR OPERATING CORPORATION !
Wolf Creek !' T,erating Station !
Docket No.: 50-484 ,
Facility Operating License No.: NPF-42 l t
ANNUAL SAFETY EVALUATION REPORT i
Report No.: 14 r
i Reporting Petiod: Januar/ 1, 1998 through December 31, 1998 !
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Attachment I to ET 99-0003 Page 2 of 2
SUMMARY
This report provides a brief description of changes, tests, and experiments performed at Wolf Creek Generating Station pursuant to 10 CFR 50. 59 (a) (1) .
This report includes summaries of the associated safety evaluations that were reviewed and found to be acceptable by the Plant Safety Review Committee for the period beginning January 1, 1998 and ending December 31, 1998. This report is submitted in accordance with the requirements of 10 CFR 50.59(b) (2) .
A significant number of the safety evaluations summarized in this report are a result of a review of the Updated Safety Analysis Report as described in letter WM 97-0009, dat.ed February 9, 1997, from O. L. Maynard, WCNOC, to USNRC.
On the basis of these evaluations of changes:
- There is no increase in the probability of occurrence or the consequences of an accident or malfunction of aquipment important to safety previously evaluated in the Updated Safety Analysis Report (USAR).
- There is no possibility that an accident or malfunction of equipment important to safety of a different type than any evaluated previously in the USAR may be created.
- The margin of safety as defined in the basis for any Technical Specification is not reduced.
Therefore, all items reported herein are determined not to involve an unreviewed safety question.
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i Attachment II to ET 99-0003 Page 1 of 217 I
Safety Evaluations- 59 93-0226 Revision: 2 I Positive Displacement Charging Pump Replacement '
Revision 0 of this Unresolved Safety Question Deternination (USQD) l evaluated the replacement of the Positive Displacement Pump (PDP) in the '
chemical and volume Control System (CVCS) with a centrifugal charging !
pump, called the Normal Charging Pump (NCP) . In_that revision to this ,
USQD, three accidents were identified which were associated with the l modification and were evaluated. It has now been identified that there ;
are three plant events other than these accidents which should have been. l evaluated: Inadvertent Actuation of Safety Injection (IASI) , Normal Letdown Isolation with pressurizer overfill ar2d Low Temperature Overpressure (LTOP). PIR 98-0645 has been initiated to determine the cause of missing these three events in the original USQD. Since the time of the original revision of DCP 04590, USOD 95-0151 evaluated the effect of the Normal ;
Charging Pump (NCP) on the inadvertent actuation of safety injection for !
Design Change Package (DCP) 06181, and found the effect to be acceptable.
This revision to this USQD evaluates the effects of the NCP on the normal '
letdown isolation and LTOP events. In the disposition section of Revision ;
3 to Plant Modification Request (PMR) 4590, the IASI and LTOP events were [
discussed, but the USQD was not revise 9 to include an evaluation of these events. j f
Three valves were located in a radioaccive pipeway to facilitate the installation of DCP 04590. The location of these three valves appears to be in conflict with a statement in Updated Safety Analysis Report (USAR) l Section 12.3.1.1.2, which states that valves and instruments are not i located in radio. ',ive pipeways. USAR section 12.3.1.1.2 is being revised to indicate that "whenever practical", valves and inctruments are not located in radioactive pipewayo. !
In addition, there are several sections, figures and tables that were not screened and evaluated for DCP 04590 in previous revisions. This revision to USQD 93-0226 evaluates the changes to the USAR that were not evaluated !
in previous revisions to this USQD.
Evaluation of LTOP analysis:
The analysis for mass input pressure transient was performed assuming one charging pump operating in a configuration producing its maximum delivery
- rate plus 100 gpm. This is the basis for determining the maximum allowable Power Operated Relief Valve (PORV) setpoints for the Cold j overpressure Mitigation System (COMS). This allowed the Positive I
Displacement Pump-(PDP) to be run at low reactor temperatures with one charging pump operable, since the PDP could not produce more than 100 GPM flow rate at any reactor pressure. This is documented in PMR 04835, !
Updated Safety Analysis Report Change Request ('USARCR)94-137, and Technical Specification Amendment 71. j i
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Attachment II to ET 99-0003 Page 2 of 217 The NCP can produce a wide range of flow depending on reactor pressure.
At low reactor pressures, that flow can exceed 100 gpm. This flow would not be bounded by the analysis for mass input pressure transient.
Therefore, in Revision 3 of PMR 04590, a statement was made that operating procedures would have to be revised to make the NCP non-functional when the plant is in the LTOP range and COMS is armed. USARCR 97-75 was issued to reflect the installation of the NCP, and the fact that it must be made non-functional whenever the plant is operating at low temperatures. There is no change to the mass input pressure transient analysis, nor to its input assumptions. Since the NCP will be made non-functional whenever the Reactor Coolant System (RCS) cold leg temperature is below the COMS arming temperature, there is no change to the margins to safety, no new accidents which are created, and no new malfunctions of equipment which are created.
As indicated in Revision 3 of the PMR, an administrative control on the operation of the NCP is necessary to maintain compliance with the LTOP analysis. This administrative control is documented in operations procedures. These precedures were revised as a result of Revision 3 of the PMR and have been in compliance since that time. Section 5.2.2.10.2 of the USAR has been revised to indicate that the NCP must be made non-functional.
This USQD revision reviews the effect of the NCP on the mass input pressure analysis for the COMS setpoint analysis for LTOP. This event is not considered a Design Basis Accident and is discussed in USAR Chapter
- 5. This modification requires the NCP to be made non-functional, whenever the RCS cold leg temperature is below the COMS arming temperature. With this requirement incorporated in the operational procedures, no new
- accident scenarios are identified. This modification requires the NCP to be made non-functional whenever the RCS cold leg temperature is less that the COMS arming temperature. Since the pump will be non-functional, the assumptions in the mass input pressure analysis remain the same. The COMS will function as it is designed. There are no credible malfunctions of equipment important to safety affected by this modif.ication.
The mass input pressure transient analysis provides the basis for establishing the setpoint of the PORVs for COMS. The transient analysis is documented in calculatinn AN-93-031. The assumptions used for this calculation include a tota mass input to the reactor which is the sum of the maximum flow capable from one Centrifugal Charging Pump plus 100 gpm margin. If the actual flow could be more than the assumed flow, administrative action needs to be taken, as indicated in section 5.2.2.10.2 of the USAR. This modification requires that administrative action to be taken. Therefore, the acceptance limit is not affected.
Evaluation of Normal Letdown Isolation with Pressurizer Overfill:
USAR section 5.2.2.10.4.2 discusses the results of the analysis of Normal Letdown Isolation with Pressurizer Overfill. It states that the operators will have to take manual action to mitigate the event and will have 18 minutes to take that action.
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l Attachment II to ET 99-0003 Page 3 of 217 A review of the history of the USAR statement has determined that it came j from FSAR question 440.106. The question postulates an event where power is lost to one of the DC safety busses at low temperature operation, j causing an isolation of the Chemical and Volume Control System (CVCS) l l letdown line and inoperability of one Pressurizer Power Operated Relief i
! Valve (PORV) . This would disable one/ half of the COMS. A single failure of the second PORV would completely disable COMS while the plant is in a '
potential overprersure event. The original answer to the question was that the PDP is normally in service. In the postulated event, the maximam ,
flow mismatch would be the maximum flow possible from the PDP, 100 gpm. '
l With that flow, the operator would have more that 10 minutes to terminate the event before the pressurizer was filled to a water solid condition.
Later, the answer was incorporated into the USAR in Section 5.2.2.10.4.2.
At a still later date, the statement was changed to show the real t calculated time of 18 minutes, which is more that the original 10 minutes stated.
The original answer to Question 440.106 was in error in that it assumed 100 gpm flow from the charging system (PDP running). This error was carried over into the calculation when 18 minutes were calculated. The actual operating conditions at low temperature operations originally required that the PDP be placed in Pull-to-Lock and a CCP be used for normal charging. Therefore, the CCP flow should have been originally i considered. The CCP can deliver more flow than the NCP, so if the CCP is analyzed, it will bound the use of the NCP for charging.
Calculation SA-92-095 has been superseded by AN-98-080. The new calculation assumes that the CCP is running whenever the COMS is armed.
With this new assumption, the crerator has less that the 18 minutes stated l in the USAR today, but more thar.- the 10 minutes originally indicated in the answer to FSAR question 440. LOG. Since there are still more than the 10 minutes originally reviewed and accepted by the NRC, the plant design bases are still met. USAR section 5.2.2.10.4.2 is being revised to reflect the new calculated time.
No procedures or plant activities are affected as a result of the new letdown isolation analysis. No tests or experiments are required to support this analysis. This USQD revision reviews the effect of the NCP on the analysis for normal letdown isolation with pressurizer overfill.
This event is not considered a Design Basis Accident and is discussed in USAR Chapter 5. In the event of a normal letdown isolation with potential overpressure, the operators will have greater than 10 minutes to take action to mitigate the event. No plant component is affected. No procedures are affected. No new accident scenarios are identified.
Operations has greater than 10 minutes to take mitigating actions to prevent pressurizer overfill in the case of isolation of normal letdown.
No equipment is affected differently than previously analyzed. There is no change to the configuration of the plant nor to the operation of the l plant. This change will not affect any previously analyzed malfunction
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Attachment II to ET 99-0003 Page 4 of 217 i
scenarios.
The analysis performed for normal letdown isolation with pressurizer
- overfill used an acceptance criteria of no operator action for 10 minutes. This time was used by Westinghouse for evaluating many plant '
transients, including the main steam line break, as shown in USAR section ,
15.0.3. ANSI Standard 58.8 was issued in 1994 and stated that 10 minutes ,
is an acceptable time to assume for no operator action. The original response to FSAR question 440.106 used 10 minutes as the acceptance l
criteria. Therefore, the acceptance criteria is that pressurizer overfill does not occur within 10 minutes.
Evaluation of Valves in Radioactive Pipeways:
USAR section 12.3.1.1.2 states, under the paragraph entitled ' PIPING',
" Valves and instruments are not placed in the radioactive pipeways". DCP 04590 and DCP 04782 added new valves and located them in a radioactive pipeway.
USAR section 12.3.1.1.1, under ' VALVES', states "To minimize personnel exposure from valve operations, motor operated, diaphragm, or other remotely actuated valves are used to the maximum extent practicable." It also states that "Whenever practicable, valves for clean, nonradioactive systems are separated from radioactive sources and are located in readily accessible areas." The design basis for separating valves from radioactive sources is, therefore, to maintain doses to the radworker as l low as reasonably achievable. Throughout USAR Section 12.3.1.1.2 the term i "Whenever practicable" is used in describing the methods of achieving l doses ALARA, including the paragraph under ' PIPING', where the statement is made that " valves and instruments are not placed in the radioactive pipeways". This is the only statement in the section which does not use the term 'whenever practicable'. In fact, according to Table 3-11(B)-3, several valves were originally located in pipeway 1204, which is the pipeway under discussion. Among these valves are EJHV8811A, EJHV0021, EJHV0023, EJHV0025, and ENHV0001. Manual valves listed in the environmental notebook as being installed in pipeway 1204 include BMV0124, l EFV0177, EMV0175, KAV0135, and KAV0138. I It is apparent that the statement that valves and instruments are not j placed in radioactive pipeways was not meant to be absolute. Valves l BGV0630, BGV0806, and BGV0807, the valves installed by DCP 04590 and DCP 04782, are valves which are not meant to be operated frequently. They are either diaphragm valves or a check valve with expected low maintenance recuirements. It was practical to locate these valves in an area where minimum time and material would be used for installation. In addition, l these valves have been qualified to the radiation levels in the pipeway.
Therefore, there is no concern other than a minimal ALARA concern with the l location of these valves, and the USAR statement will be revised to j include the term 'whenever practicable'. This is an ALARA concern only, and not a safety or off site dose concern. These valves do not have to be manipulated in the event of an accident or safe shutdown.
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Page 5 of 217 No procedures or plant activities are affected as a result of the f permission to locate valves or instruments in radioactive pipeways if
, needed. No tests or experiments are needed to support this change. This USQD revision reviews the effect of placing valves in radioactive pipeways
. on the exposures'of personnel inside and outside the plant as discussed in l USAR Chapter 12. It also reviews the effect on the analysis on radioactive release from a subsystem or component as discussed in USAR Chapter 15. Locating new valves in a radioactive pipeway does not create any new accident scenarios. The potential breach of the CVCS pressure I
boundary has already been analyzed and is documented in USAR section 15.6.2. These valves are qualified for the system pressure snd temperature and for the environmental conditions in the pipeway. The ne9 valves located in the radioactive pipeway are qualified for the environmental conditions found in the pipeway during normal and accident conditions. The only credible malfunction.that could be affected would be breach of the pressure boundary. Since the pipe and valves are qualified for the internal pressures, they are installed under a qualified QA program, and they are qualified for the external environmental conditions, this malfunction is not affected by this change. The acceptance criteria for locating valves in the plant is that they will be qualified for the internal pressure, the internal flowrate, the fluid chemistry, and the external environmental conditions. The new valves were purchased to a safety related specification which requires that they meet the conditions for which they are being used and for the environn.snt in which they are located.
Evaluation of Miscellaneous USAR changos:
The USAR changes being made are to Sections 9.4.3.2.1, 14.2.12, Tables 3B-1, 3.11(B)-1, 3.11 (B) -2, Figures 3B-1, 3.6-1, 9.5.1-2-01, and figure 12.3.2. In all cases, the changes are editorial in nature. The editorial changes are to indicate that the pump name changed from the Positive Displacement Pump to the Normal Charging Pump. No technical or operational information is changed or affected. None of these editorial changes have any effect on the physical plant, the operation of the plant or any design bases analyses.
No procedures or plant activities are affected as a result of the permission to locate valves or instruments in radioactive pipeways if needed. No tests or experiments are needed to support this change. Since theses miscellaneous changes are only editorial in nature, there are no design basis accidents that need to be reviewed potential impact.
Making. editorial changes in the USAR tC change title of the pump from the Positive Displacement Pump (PDP) to the Normal Charging Pump (NCP) does not create any new type of credic7.e accident. Editorially changing the title of the pump from PDP to NCP does not affect any previously analyzed credible nalfunction of equipment important to safety. The only acceptance criteria for editorial corrections is that the USAR must accurately reflect the plant conditions.
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Safety Evaluation: 59 96-0095 Revision: 1 i Clarification of Updated Safety Analysis Report Limits for Post Accioent sampling system f Revision 0 of Unreviewed Safety Question Determination (USQD) 96-0095 evaluated the following: r This change to the Updated Safety Analysis Report (USAR) involves the ,
clarification of limits within the USAR to coincide with the regulatory
- requirements of Regulatory Guide 1.97, Revision 2 " Instrumentation for .
Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs '
Conditions During and Following an Accident," and NUREG-0737, "
" Clarification of TMI Action Plan Requirements." The analyses to be .
performed by the Post Accident Sampling System were clarified as to the '
location for the analysis (i.e. on-site or off-site depending on dose !
rates). The change states that the containment hydrogen monitor is not a !
part of the Post Accident Sampling System. 4 Revision 0 of USQr 96-0095 determined that this change does not involve an j unreviewed safety question. '
i Revision 1 to USQD 96-0095 evaluates the-following:
This change to the USAR involves the clarification of limits within the ;
USAR to coiteide with the regulatory requirements of Regulatory Guide 1497, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident". Specifically, the limits identified in USAR section 18.2.3.2, Table II.B.3-1 (i.e. the boron, pH, hydrogen and chloride limits) . These limits were revised so that they correspond with Regulatory Guide 1.97, Revision 2. The limits stated in the USAR were not the limits used by Chemistry in the performance of these analyses In addition, it will ,
clarify the times when and where a chloride sample analysis will be l performed. The samples will be performed on site if the dose rates are -
low enough, otherwise the sample analysis will be performed off-site.
NUREG-0737 states that chloride sample times of 4 days (96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />) are allowed for plants which do not use seawater as a cooling medium and that ;
on-site analysis of the sample is not required. This change also clarifieu !
that the containment hydrogen monitor is not part of the Post Accident Sampling System cabinet (SJ145). There are no issues or concerns about -
the changes for the chloride analysis .ime limit and location, the analyses limits for pH, boron or chloride nor the clarification about the i location of the containment hydrogen analyzer. -
There are no other activities performed in association with the Post l Accident Sampling System that would make the information contained within the USAR incorrect. The operation of the system is based on NUREG-0737 '
and Regularary Guide 1.97, Revision 2.
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Attachment II to ET 99-0003 Page 7 of 217 L review of the accident scenarios in the USAR (contained within chapters 2, 3, 6, 9 and 15) identified no scenarios in which the proposed revision ,
w1]l have any impact. The Post Accident Sampling System is used in post !
accident scenarios to help evaluate the extent of the accident. The use of the system after an accident has happened dictates that the system will have no activities which might cause an accident to occur.
There are no credible accident scenarios that the proposed revision to the USAR could create. This is based on the documentation and pr0 cess only nature of the change. The process changed per this revision to the USAR is the allowance of a chloride sample to be analyzed either on-site or off-site and the time limit for the sample analysis to be completed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
Due to the location of the PASS panel and its support equipment, there are no credible malfunctions of equipment important to safety due to the proposed change. The basis for modification of the various sample limits and clarification of the chloride analysis location / time is based upon a review of NUREG-0737 and Regulatory Guide 1.97, Revision 2.
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Safety Evaluation: '59 96-0208 Revision 0 1
Effect of Higher Fuel Pellet Density On Spent Fuel Pool Criticality
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Analyses This evaluation addresses changes to the Updated Safety Analysis Report (USAR) which account for differences in the pellet theoretical densities <
used in the Spent Fuel Pool (SPF) criticality analyses and the fuel vondor pellet manufacturing tolerances.
Reactor Engineering performed a self assessment to evaluate the accuracy !
of the USAR with respect to Core Design data and methodologies employed at Wolf Creek Generating Station (WCGS) (Reference l' One of the items ,
identified as part of the self assessment was a difference in the range of fuel pellet densities used in the Spent Fuel Pool (SFP) criticality analysis described in USAR Appendix 9.1A and the range of pellet densities contained in the fuel vendor testing and acceptance specifications.
The SFP criticality analysis is contained in USAR Appendix 9.1A. The analysis'was produced by Pickard, Lowe, and Garrick, Inc. (PL&G) in 1987.
As described in the USAR, this analysis assumes a nominal fuel pellet density of 94.5 percent of theoretical density with an associated j uncertainty of 11 percent.
i The testing and acceptance criteria on fuel pellet density for all fuel used at WCGS is contained in the Wertinghouse Nuclear Fuel Division ;
Product Specification on Uranium Dioxide Pellet NFP 31029. This specification requires pellet testing for density such that at a 95 1 percent confidence level, at least 95 percent of the pellets will have a density of 95 percent of theoretical density with an associated )
uncertainty of 11.5 percent.
An analysis has been performed to access the impact on the SFP criticality j analysis of using the higher theoretical density used by the fuel vendor, i These calculations have determined that the increase in density results in i small increases in predicted k-infinity for SFP Regions 1 and 2. However,.
the analysis has also shown that these small increases in k-infinity can be accommodated in existing margins and the change has no impact on the l conclusions documented in the USAR. l The conclusion of the existing analysis contained in the USAR is that an accidental criticality event cannot occur in the SFP because under all conditions the multiplication factor (i.e., k-infinity) of Region 1 and ,
Region 2 will remain below 0.95. The change in theoretical density does l not change the analysis conclusions and the multiplication factor for Regions 1 and 2 continue to be less than 0.95. i The change to the USAR to account for the difference in fuel pellet !
theoretical density used in the SFP criticality analysis and the fuel l r
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1 Attachment II to ET 99-0003 Page 9 of 217 l vendor manufacturing tolerance has no impact on analyses which establishes spent fuel fission product inventories for use in determination of l radiological consequences of postulated events. The above changes will improve the accuracy and clarity of the USAR.
The accident analysis was reviewed for potential impacts. The changes have potential impact on the SFP criticality analysis and on the analysis of radiological consequences of postulated events involving spent fuel fission product inventories. i As discussed above, factoring the small increase in predicted k-infinity for SFP Regions 1 and 2 associated does not change the original conclusion stated in the USAR that a criticality accident in the SFP will not occur '
because the multiplication factor will be less than 0.95 under all postulated accident conditions. For Region 1, the effect of increasing
- the assumed theoretical pellet density from 94.5 percent il to 95 percent 11.5 percent is an increase in basic cell k-infinity from 0.9329 to 0.9335. The conservatively calculated multiplication factor for Region 1, after adjusting for all biases and uncertainties, is 0.9480 which is less ,
than the 0.95 design limit as described in the USAR. For Region 2, the effect of increasing the assumed theoretical pellet density is an increase ;
in basic cell k-infinity from 0.8998 to 0.9008. The conservatively calculated multiplication factor for Region 2, after adjust.ing for all biases and uncertainties, is then 0.9276 which is less than the 0.95 design limit.
i The minimum burnup requirements versus enrichment for storage of spent fuel in SFP Region 2 are shown in USAR Figure 9.1A-26 and corresponding Technical Specification Figure 3.9-1. These figures are not impacted by the change since they include substantial margin with respect to the requirement that kinf be less than 0.95.
The change to address the differences in theoretical fuel pellet density does not have any impact on the separate Westinghouse analysis by which tables of spent fuel fission product inventories are produced. Fuel fission product inventories are used to evaluate the radiological consequences of postulated events such as the fuel handling accident.
These inventories are described in USAR Table 15A-3. Source terms for these values are obtained from the re-rate analysis of the Wolf Creek Generating Station. Pellet densities were correctly accounted for in this analysis and thus, there is no impact to spent fuel fission product inventories as listed in the USAR associated with this change request.
The proposed USAR changes will not create any credible accidents which have not already been addressed by the original Final Safety Analysis Report (FSAR), USAR, or revisions already addressed in the USAR. All design and performance criteria continue to be met and no new failure modes have been introduced for any system, component, or piece of equipment as a result of the changes. The changes will not impact either the normal plant operation or the response to accident conditions. The
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l Attachment II to ET 99-0003 page 10 of 217 change in pellet densities does not involve a physical change to any physical structure, component, or equipment.
The changes do not result in a different response of safety related
, systems and components to accident scenarios than those described in the l
USAR. No new equipment malfunctions have been introduced that will affect fission product barrier integrity. There are no new credible accidents j
associated with the USAR changes. The changes do not affect the safety function of safety related systems and components which are related to accident mitigation. Therefore, the changes will not create the
- possibility of a malfunction of equipment important to safety different i than those already described in the USAR. '
The proposed USAR changes do not affect any acceptance limits which are contained in the bases for the Technical Specifications or license bases I
- documents. This evaluation shows that all design and safety analysis l limits continue to be met and that these limits are supported by the applicable Technical Specifications. The margin of safety as defined in i l tiae Bases is not reduced for any USAR accident, no new accidents are l created, and no new malfunctions of equipment important to safety are created and therefore, the margin of cafety has not been reduced. There is no impact on spent fuel fission product inventories as described in the USAR and thus, there is no impact on the radiological consequences to any postulated event. All acceptance limits continue to be met.
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Safety Evaluation: 59 97-0057 Revision: 0 Makeup Domineraliser System Upgrades i Design Change Package (DCP) 06586 replaces much of the instrumentation in~
the Demineralized Water Rakeup System (WM) Control Panel, 1PLO12J, located j in the Shop Building. Three Permutrol SSP Programmers, 1WMP1, 1WMP2 and i IWMP3, presently installed in the control panel, operate the WM System in l the automatic mode. One of the programmers is presently inoperative which requires the system to be operated only in the manual mode, The SSP .
programmers are obsolete and can no-longer be repaired. l t
The Permutrol SSP Programmers will be replaced with an Allen-Bradley PLC-5/30 programmable logic controller (PLC). The PLC will perform all j functions previously performed by the SSP programmers. [
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A computer will be installed near control panel IPLA12J. Data from !
various analyzers, transmittars and equipment in the demineralizer system !
will be acquired through the PLC and trended and displayed by the computer. ,
The_ changes to the WM system are required to replace obsolete equipment l which in some cases is inoperable and can no longer be repaired. The !
replacement equipment will increase the system operability, reliability ;
and accuracy. Functions performed by the Demineralized Water Makeup l System remain unchanged. l Most of the instrumentation which will be replaced is located in the Water !
Treatment Control Panel, IPLO12J, located in the Shop Building. The flow l transmitters to be replaced are located in the demineralizer trains inside the Shop Building. The level and temperature transmitters are installeo in the water storage tanks located outside on the west side of the shop j Building. !
The Demineralized Water Makeup System serves no safety functions and has no safety design basis. There are no design basis accidents discussed or ,
referenced in USAR chapters 2, 3, 6, 9 or 15 which are impacted by the l proposed equipment modifications. ;
Credible accidents that could occur in regard to the Demineralized Water l Makeup System include leakage of softened, filtered and/or demineralized ;
water, chemical leaks, failure to produce tha required water quality and ,
failure to' operate. None of the credible accidents affect plant safety functions. The probability of accident occurrences are not increased by i the proposed modifications. There are no new accidents created by the I proposed changes to be implemented by DCP 06586. ?
i None of the equipment to be modified under DCP 06586 performs any safety-related functions. The equipment has no failure modes which can affect '
plant safety. There are no special scope functions performed by the j
Attachment II to ET 99-0003 Page 12 of 217 j equipment.
USAR section 9.2.3.1.2, " Power Generation Design Basis One", and USAR Table 9.2-16 both show the Plani +ter Chemistry Specifications for Demineralized Water. Conductivi.y, silica and pH requirements are presented in both places. DCP 06586 replaces equipment which is used to measure, display and/or record these parameters. However, DCP 06586 does not alter the chemistry specifications or WM system functions.
Based on this evalustion, this modification will not increase the i probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated '
previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification.
Therefore, this modification does not involve any unreviewed safety ,
question.
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. Safety Rvaluation: 59 97-0073 Revision 0 Updated Safety Analysis Report Changes to Update to Cycle 9 Values This evaluation addressee changes to the' Updated Safety Analysis Report [
(USAR) which account for differences between ' current and historical core !
design parameters. (
l In response to Performance Improvement Request (PIR) 96-2601, Recctor ;
Engineering performed a self assessment to evaluate the accuracy of the i Updated Stfety Analysis Report (USAR) with respect to Core Design data and methodologies employed at Wolf Creek Generating Station (WCGS) . The .
- following USAR inaccuracies were identified. l l
Item 1: Current revision of USAR Section 4.3.2.4.6 contains this statement: _
" Excess reactivity of approximately 10 percent delta p (hor) is installed
- at the beginning of each cycle to provide sufficient reactivity to '
compensate for fuel depletion and fission product buildup throughout the cycle." This exact statement also appears in Section 4.3.2.4.6 of the ;
WCGS Final Safety Analysis Report (FSAR). It was true for the 12-month cycle initial core. The beginning of life (BOL) hot excess reactivity for an'18-month cycle reload core at WCGS is close to 20 percent, not 10 l percent. This fact is documented in Calculation AN-96-128.
The statement in Section 4.3.2.4.6 is part of a general description of )
reactivity control. As stated in Section 4.3.1.1, "A limitation on I
initial installed excess reactivity or average discharge burnup is not required other than as is quantified in terms of other design bases, such as core negative reactivity feedback and shutdown margin...."
USAR Change Request (CR)97-101 revises Section 4.3.2.4.6. This revision explains that the core is composed of fuel assemblies, and that these assemblies are composed of fuel rods. The revision explains that the uranium fuel in these rods is enriched in the U-235 content. The revision explains that the "approximately 10 percent" initial excess reactivity
- statement in USAR Section 4.3.2.4.6 was appropriate for the low-enrichment 12-month cycle of the first core. The revision explains that Wolf Creek is no longer operating in Cycle 1. The revision explains that Wolf Creek has operated on 18-month cycles for several years, starting with cycle 4. The .
revision explains that initial excess reactivity for an 18-month cycle must be greater than the initial excess reactivity for a 12-month cycle. I The revision explains that the initial excess reactivity for Wolf Creek 18-month cycles is approximately 20 percent.
Item 2: Table 4.3-2 contains a list of reactivity coefficients, including Moderator Temperature Coefficient (MTC), which are described as values "used as design limits in the transient analysis" in Section 4.3.2.3.5.
. Table 4.3-2 shows the limit of MTC values as +6 to -41 pcm/ degrees 1
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I Attachment II to ET 99-0003 Page 14 of 217 Fahrenheit. The lower limit on MTC has been changed to -50 pcm/ degrees Fahrenheit, as evaluated in USQD 59 96-162. The limit was changed for Cycle 9 operation.
Item 3: Table 4.3-2 contains a list of reactivity coefficients, including Rodded Moderator Density Coefficient (MDC), which are described as values "used as design limits in the transient analysis" in Section 4.3.2.3.5.
Table 4.3-2 shows the limit on MDC as < 43,000 pcm/gm/ce. Associated with the change in MTC range (Item 2) is a change in this MDC limit. The new limit is less than 47,000 pcm/gm/cc. The-limit was changed for Cycle 9 l operation and is included as a safety analysis limit in the Reload Safety Analysis checklist (RSAC).
- Item 4: Table 4.3-2 contains a list of reactivity coefficients,-including l Doppler Temperature Coefficient (DTC), which are described as values "used as design limits in the transient analysis" in Section 4.3.2.3.5. Table -
4.3-2 shows the limit of DTC values as -2.9 to -1.4 pcm/ degrees F. This range has been expanded to be -3.5 to -1.0 pcm/ degrees F. This is l documented in AN-95-064 " Cycle 9 Rod Ejection Analysis" and is included as a safety analysis limit in the RSAC. I Item 5: Table 4.3-2, Sheet 2 contains the following values for Prompt Neutron Lifetime: 19.4 microseconds (BOL), 18.1 microseconds (EOL).
These values are in the corresponding WCGS FSAR table. They were calculated for Cycle 1. Prompt neutron lifetime is different for each reload core. There is no design limit associated with prompt neutron lifetime; however, values calculated for each reload core are used in transient analyses. These values in Table 4.3-2 will be updated to Cycle 9 values. :
Item 6: Section 4.3.2.4 states, " Boron concentrations for several core conditions, are listed in Table 4.3-2." Table 4.3-2 does contain seven RCS boron concentration values (ppm) for seven sets of conditions. These identical values also appear in the corresponding WCGS FSAR table, Table 4.3-2A for Core A, Cycle 1; therefore, these values have been out-of-date since the beginning of Cycle 2.
The changes to boron concentrations contained in USAR CR 97-101 is an update of information; this change will remove values which were valid only for Cycle 1 and replace them with values which are valid for Cycle 9. '
Item 7: Section 4.3.2.5 states, " Typical control bank worths are shown in Table 4.3-2." Table 4.3-2 contains bank worth (pcm) values for Bank D, Bank C, Bank B, and Bank A. The only condition listed is "HZP no overlap". These exact values occur in WCCS FSAR Table 4.3-2A for Core A, Cycle 1; therefore, these values have been out-of-date since the beginning of Cycle 2.
This change to Rod Control Cluster Assembly (RCCA) bank worth values !
contained in USAR CR 97-101 is an update of information; this change will r
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Attachment II to ET 99-0003 Page 15 of 217 remove values which were valid only for Cycle 1 and replace them with values which are valid for Cycle 9.
Item 8: Section 4.3.2.3.5 states that Table 4.3-3, " Reactivity Requirements for Rod Cluster Control Assemblies", contains values for RCCA worths for a " hypothetical equilibrium cycle". It also states that these values are "provided for information only since refueling specifications for subsequent cycles have not yet been established." Values currently in Table 4.3-3 for the " hypothetical equilibrium cycle" are the same as those in FSAR Table 4.3.3A for Core A, Cycle 1. These old "for information only" " hypothetical equilibrium cycle" values will be replaced with RCCA worths based on Cycle 9.
Details of origin of the RCCA worth values in Table 4.3.3 for the
" hypothetical equilibrium cycle" are unknown. They were provided "for information only". Changes to Table 4.3.3, contained in USAR CR 97-101, replace these old values with values based on calculations for the Cycle 9.
The above changes will improve the accuracy and clarity of the USAR and have previously been evaluated.
USARCR 97-101 has three categories of changes:
- 1) Changes evaluated by previous USQDs.
- 2) Changes to safety analysis limits under the control of the Reload Safety Evaluation /RSAC process.
- 3) Information only updates of parameters which are part of the core design process.
Therefore, appropriate review has been previously performed.
Attachnent II to ET 99-0003 Page 16 of 217 Safety Evaluation: 59 97-0086 Revision: 3 Fiber Optic Containment Penetration The scope of DCP # 07065 is being expanded to include communications cabling for remote radiation monitoring / dosimetry and for non-outage video monitoring. This expanded scope requires the routing of fiber optic cables to locations not previously identified in DCP # 07065. DCP Revision B is being issued for this purpose. This revision to the DCP relocates some of the fiber optic cables previously designated for free-air routing and also adds new fiber cables. These cables are now being scheduled for installation in plant separation Group 5 or 6 raceway. This will minimize routing concerns involving safe shutdown separation criteria brought about by free-air installation of the cable. The fire loading identified in Updated Safety Analysis Report (USAR) Section 9.5B will be affected in two ways: There will be an increase in the combustibles due to new cables and different fire zones will be affected due to re-routing of cables. Cables remaining in the design as " free-aired" (from Revision 6 of the DCP) were previously evaluated on Revision 2 of USQD # 59 97-0086.
For those cables, Revision 2 USQD remains valid.
The electrical separation criteria for permanent plant raceway at Wolf Creek was developed from IEEE 384-1974 and NRC Regulatory Guide 1.75, Revision 1. The criteria is applied to WCGS in documents E-0, E-11013 and E-1R8900. The separation criteria established in these documents is for maintaining the redundancy of Class 1E power systems and the redundancy of I Safety-related protection systems. The regulatory commitments are identified in USAR Section 8.3.1.4.1. These requirements are being met by l the cable installed under this revision since it will be placed in i existing raceway. All cabling identified under this modification is non- .
Safety related and is not connected to any plant process systems. Cable l jacketing is IEEE-383 qualified for flame retardance as required by l electrical design criteria. l l
All cables added or affected by Revision B of DCP # 07065 is being i installed in existing separation group 5 or 6 .eway. None of the design l basis accidents identified in the subject chapters involve non-safety related cable faults in these raceways. Therefore, the additions have no impact on accident discussions in the USAR.
Due to the restrictions placed on their installation, the non-safety related cables can not cause any credible accidents. The only credible l failure mode is the failure to perform their own non-safety functions.
These functions are typically communications for non-process computers, outage video communications, outage microcell telephone system interconnection, remote radiation monitoring and dosimetry and communications for other outage activities such as eddy current testing and sludge lancing. Cable failures in these applications do not affect nuclear safety related systems.
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The cable additions identified on Revision 8 of DCP # 07065 will be installed in existing plant raceway. The installation will adhere to design criteria specified in applicable design documents. Based on this, credible malfunctions of equipment important to safety are not affected.
The acceptance limits for cable separatio" will not be exceeded since the applicable design requirements are beine followed for this installation.
The combustible loading additions hav' been evaluated, found to be acceptable and the information conesrning fire loading in the USAR is being revised accordingly. The "echnical Specifications do not address this issue and are not affectea by the USAR change.
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Attachment II to ET 99-0003 Page 18 of 217 Safety Evaluation: 59 97-0089 Revision: 1 Clarification to Updated Safety Analysis Report The pipe routing of the boric acid system was reviewed in detail to address a question regarding the minimum temperature allowable for assuring that the 4 WT percent boric acid solution remains above its solubility temperature-limit. This will be clarified in the Updated Safety Analysis Report (USAR) since it may have an affect decisions regarding the minimum temperature that would be acceptable for several rooms in the Auxiliary Building. For 4 WT percent boric acid solution USAR Section 9.3.6.2.1 suggests that the solubility limit is 55 degrees Fahrenheit (F). Those areas of the ALniliary Building that can contain tanks, piping and components containing 4 WT percent boric acid solution may experience boron precipitation at solution temperatures below 55 degrees F.
The precipitation of boron would not cause the boric acid system to become immediately inoperable due to pipe plugging. The precipitation phenomena would instead be a long term degradation concern. Over time, if a stagnant borie acid pipe experiences multiple events of precipitation of boron, layers of boron could build up to a point where the pipe could become plugged. The short term occurrence of room or solution temperature being below the solubility limit is not an operability concern but operation in this range should be limited to avoid possible long term degradation of the flow path.
One of the boric acid flow paths that would be affected is used for immediate or emergency boration as called for in Off Normal Operating Procedure OFN BG-9. The other affected flow paths include the path used during normal boric acid batching operations as well as during normal operating boration via the Centrifugal Charging Pumps or Normal Charging Pump. Although some of these affected flow paths are safety related, they are not the only flow path available for boration of the reactor.
T'eo emergency procedures primarily rely upon the Safety Injection (SI) flow path for safe shutdown of the plant following a design basis accident. The Safety Injection pump takes suction from the Refueling Water Storage Tank (RWST). A 55 degree F temperature will not adversely affect SI because the solution concentration in the RWST is 2400 to 2500 PPM or less than 2.5 WT percent which has a solubility limit of 37 degrees Fahrenheit (F). Although an affected flow path is called upon in Off Normal Procedures to provide immediate boration of the reactor, there are other flow paths that would be available to align suction to the RWST that are described in the Off Normal Procedure as long as RWST level is sufficient. This option could be used by operators to borate the reactor if immediate horation via BGHV8104 was not available. Thus, the primary SI safe shutdown boration flow path is not impacted. However, the possibility of losing an affected flow path is a concern because it would reduce the number of redundant flow paths available to the operators.
i Attachment II to ET 99-0003
- Page 19 of 217 A USAR change request has been initiated to revise Table 3.11(B)-1 " Plant Environmental Normal Conditions", to acknowledge the concern of boron '
precipitation in certain rooms within the Auxiliary building. Note 8 shall be revised as follows "The minimum normal operating room temperatures do not represent operability or equipment qualification limits. They do not represent the minimum temperature that may occur during abnormal conditions. However, -
rooms containing tanks, piping or components with 4 WT percent boric acid .
solution (rooms 1115, 1114, 1107, 1113, 1108, 1111, 1109, 1407) or higher concentrations (rooms 1306, 1302, 1117, 1116, 1203, 1204, 1307 and 1308),
may experience boron precipitation from the boric acid solution if the solution temperature drops below the minimum normal room operating temperature stated in the table." ,
i Note 8 of USAR Table 3.11(B)-1 indicates that the minimum stated temperatures are not operability limits or equipment qualification limits. However, the minimum temperatures for those rooms with high concentration boric acid solution may experience boron precipitation if solution temperatures fall below the stated minimum room temperatures.
The WCGS design / licensing basis regarding the winter design temperatures
, and assumptions were reviewed. USAR Table 9.4-1 "Outside Environment Design Conditions" and USAR Table 1.2-1 " Design Envelope" identifies
" normal design conditions" and " extreme design conditions" for the WCGS.
After reviewing the USAR, the following is concluded:
- 1. Originally, some WCGS systems, structures and components may have been designed for the original SNUPPS normal outdoor winter temperature of -25 degrees F. Later, the Wolf Creek site specific winter outdoor dry bulb design temperature of +7 degrees F was established and used for HVAC system design.
- 2. Originally, some of the WCGS structures, systems and components may have been designed for the original SNUPPS " extreme" winter temperature of -60 degrees F. Later, the Wolf Creek site specific " extreme" winter conditions was defined as -30 degrees F.
Safety Analysis accident scenarios are all based upon the worst case temperatures for the given accident which is almost always the maximum temperature that might occur. The only accident that was found to use a minimum temperature value was the inadvertent containment spray scenario.
In this scenario it is assumed that the lowest outdoor temperature for the Refueling Water Storage Tank (RWST) is 37 degrees F but the indoor 4
containment temperature is conservatively assumed 100 degrees F. A lower containment temperature would only lessen the pressure change resulting i from an inadvertent containment spray event.
WCGS relies upon non-safety related space heaters during normal winter ;
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Attachment II to ET 99-0003 Page 20 of 217 operation to maintain many plant areas at acceptable temperatures. This heating is not required post accident. This is an acceptable use of non-safety related equipment.
During performance of emergency (EMG) procedures that require a Safety Injection, there is no need for the boration flow path through the pump rooms which was questioned. Also, alternate boration flow paths, such as alignment to the RWST, would be available to achieve horation if sufficient level is available in the RWST.
In general, the design basis assumption has been that the average room temperature in the Auxiliary building is above 60 degrees F and above 50 degrees F in the containment.
As discussed in above, or rooms containing 4 WT percent or higher concentration boric acid solution, precipitation of boron from the solution may occur at temperatures less than the minimum stated in Table 3.11(B)-1 for pipes that do not have operable heat tracing.
l This USAR change will recognize the boron precipitation concern and will l make the USAR consistent with the design and safety analysis assumptions ;
as stated in the USAR. Thus, the change does not impact any of the design basis accidents. 1 There are no new accidents being created by recognition in the USAR that the precipitation of boron can occur at temperatures lower than those i stated in Table 3.11(B)-1.
There are no malfunctions of equipment which could be affected by this I consistency change to the USAR.
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There are no Technical Specification limits based upon minimum temperature that would be affected. The Limiting Conditien for Operation (LCO) on Borated Water Source Operating / Shutdown that is shown in USAR sections 16.1.2.5 and 16.1.2.6 address the minimum boric acid system storage requirements. This USAR change would not cause a change in the LCO.
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attachment II to ET 99-0003 Page 21 of 217 Safety Evaluation: 59 97-0170 Revision 0 Revision to Updated Safety Analysis Report to Correct Inaccuracies The Updated Safety Analysis Report (USAR) Table 3.11 (B) -2, Note 16, states that in rooms served by Engineered Safety Features (ESP) coolers or fans outside containment, the temperature and relative humidity will not exceed 122 degrees Fahrenheit (F) and 95 percent relative humidity, respectively, following a loss of normal ventilation with the major components (e.g.,
pumps) in the room operating for extended periods. This note applies to room 1115. However, the Mechanical / Nuclear Design Criteria (M-000), Table 2-3 does not reference the same note applicable to room 1115.
The USAR Table 3.11(B)-1 lists the plant environmental normal conditions.
For ESF pump roort it shows the normal temperature of 104 degrees F. The USAR and the WCG. .echanical/ Nuclear Design Criteria (M-000) states that the conditions (the pressure, temperature, and humidity ) provided are for the limiting Design Basis Analysis (DBA). However, it does not apply to the pressure, temperature, and humidity conditions for other than the containment and the rooms numbers 1411, 1412, 1508 and 1509 as specified in the note. Therefore, Note 13 needs to be tagged only with the above room numbers / building.
Note 16, in the USAR Table 3.11(B)-2 was added by FSAR Change Request 86-018 for room numbers 1107 through 1315. However, during updating the same information in the Mechanical / Nuclear Design Criteria, M-000, Table 2-3, was added only for room numbers 1107 through 1114 and skipped for room number 1115.
The existing Note 13 of USAR Table 3.11(B)-2 and that of the Mechanical / Nuclear Design Criteria, M-000, Table 2-3, is applicable to the entire tables. The note is not tagged only to containment and auxiliary building room numbers 1411, 1412, 1508 and 1509. However, according to piping stress calculations P-001, P-001A, P-002, P-026 and P-27BY this note is applicable only to containment and auxiliary building room numbers 1411, 1412, 1508 and 1509. Therefore, this note is added for the containment and auxiliary building room numbers 1411, 1412, 1508 and 1509 in the above referenced tables.
Note 6, of USAR Table 3.11(B)-2, plant environment qualification parameters, is not consistent with that note in Mechanical / Nuclear Design Criteria, M-000, Table 2-3. The USAR states that following a postulated main steam line break the containment vapor could become superheated, and the temperature of the vapor could exceed the containment design value of 120 degrees F. The corresponding note in Mechanical / Nuclear Design Criteria, M-000 states, the corresponding containment design temperature is 320 degrees F. Similarly, the USAR states that equipment design considers containment conditions: superheated vapor temperature 386.5 degrees F, saturated temperature 279.4 degrees F and duration of
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superheated conditions 150 seconds. Where as the Mechanical / Nuclear Design Criteria, M-000 states the corresponding superheated vapor temperature 384.9 degrees F, saturated temperature is 250 degrees F and t duration of superheated conditions is 120 seconds. Per USAR Change Request 92-090 (PMR 03478) the corresponding correct values are as follows:
i Superheated vapor temperature = 386.5 degrees F ,
Saturated vapor temperature = 279.4 degrees F Duration of superheated conditions = 150 seconds During close out of the above change package, while incorporating the ;
above revised information in the USAR, the design temperature of 320 degrees F was changed to 120 degrees F by drafting error and was never noticed before. Containment design calculations are performed assuming ,
temperature of 320 degrees F. i Note 7 of USAR Table 3.11(B)-2 was deleted per FSAR CR 86-018. However, j it appears that it was overlooked at that time and could not be deleted from the Mechanical / Nuclear Design Criteria M-000. There is no need for this note; therefore, it has been deleted from the Mechanical / Nuclear Design Criteria M-000, t In order for the information in the US,AR and in the Mechanical / Nuclear Design Criteria, M-000 to be consistent, both documents are revised with the correct information. The changes are administrative, incorporated to ,
captured the design basis information.
This revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.
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i Safety Evaluations- 59 97-0175 Revision: 0 Correct System Descriptions for Residual Heat Removal and Containment Spray These descrepancies were identified _during the Updated Safety Analysis Report (USAR) Fidelity Review effort. This Configuration Change Package (CCP) revises system descriptions for Residual Heat Removal (RHR) and .
Containment Spray systems to correct the following sections: '
RHR System Description Section 3.1.10 states, "The only motor-operated }
valves in the RHR System which are subject to flooding are the suction isolation valves." The elevations of these valves are above the maximum i containment flood level'following a LOCA/MSLB. Therefore, these valves !
will not be subject to flooding. The system description has been revised to read, " The motor-operated suction isolation valves are not subject to j flooding." ^
Section 3.1.9 of the RHR system description states, "A leaktight seal is :
provided so that neither the pressure vessel nor the guard pipe is connected directly to the sump or containment atmosphere." This statement is not correct as the guard pipe is connected (welded) to the inside liner plate of the sump. This statement has been revised to read, "A leak tight seal is prorided such that the ambient inside the pressure vessel and :
outside the process line and enclosed within the guard pipe is not l directly connected with the' containment sump or containment atmosphere " ,
i A similar statement is provided in Updated Safety Analysis Report (USAR)
Section 3.1.11 of the Containment Spray system description which reads, "
A seal is provided so that tt" encapsulation is not connected directly to ,
the containment sump or containment atmosphere." This statement is not correct as the encapsulation arrangement is connected (welded) to the inside liner plate of the containment sump. This statement has been t revised to read, " A seal is provided so that the ambient inside the encapsulation is not directly connected with the containment sump or containment atmosphere."
The following sections of the USAR are revised for clarification:
USAR Section 5.4.7.2.3 (Page 5.4-37) for Encapsulation has a statement which reads, " A leaktight seal is provided so that neither the compartment nor the guard pipe is connected directly to the sump or ;
containment atmosphere." This statement has been revised to read, "A leak ;
tight seal is provided such that the ambient inside the pressure vessel and outside the process line and enclosed within the guard pipe is not directly connected with the containment sump or containment atmosphere."
USAR Section 6.2.2.1.2.2 (page 6.2-46) for Encapsulation has a statement: ;
"A seal is provided so that the encapsulation is not connected directly to ;
the containment sump or containment atmosphere." This statement is l l
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l revised to read, "A seal is provided so that the ambient inside the '
encapsulation is not directly connected with the containment sump or l containment atmosphere."
Because no design basis accidents are identified the probability of l occurrence of an accident is not affected. Because no design basis I accidents are identified, the consequences of accidents are not affected.
Because no malfunctions are identified the probability of occurrence of a i malfunction is not affected. Because no malfunctions are identified the consequences of a malfunction are not affected. Because no credible accidents that could be created are identified no accidents of a different type can be created. Because no malfunctions are identified no malfunctions of a different type can be created. Because no acceptance ;
limits are identified that could be affected, the margin of safety is not j affected.
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Attachment II to ET 99-0003 Page 25 of 217 Safety Evaluation: 59 97-0176 Revision: O Spent Fuel Pool 35 Degree Allowance The Updated Safety Analysis Report (USAR) is being revised to reflect the acceptability of allowing Spent Fuel Pool (SFP) water temperature to go as low as 35 degrees Fahrenheit (F). USAR Sections 9.1A.2.2 and 9.1A.3.3 currently address in the reactivity analysis the acceptability of water temperatures as low as 60 degrees F. Subsequent review of the reactivity analysis, as follows, has found that a water temperature as low as 35 degrees F is acceptable, therefore allowing the USAR to be revised to reflect the lower temperature. Review of structural calculations indicates that temperature differentials of 160 degrees (normal maximum differential) and 220 degrees (maximum off-normal differential) were used. These bound any credible differential temperatures from a structural standpoint.
Spent Fuel Pool SFP) Region 1:
USAR Section 9.1A.2.2, paragraph 5, indicates a minimum SFP temperature of 60 degrees F. The minimum can be reduced to 35 degrees F with no adverse effect on SFP criticality. Because the SFP is over-moderated at the temperature range in question, as evidenced by Figure 9.1A-4, reduction in temperature from 60 to 35 degrees F adds a negative reactivity. The reactivity added, based on Figure 9.1A-4, is as follows: A large negative reactivity (approximately -0.0015 delta K) from 60 degrees F to the maximum water density point, 39.2 degrees F, plus a very small positive reactivity (arnroximately 0.0002 delta K) from 39.2 to 35 degrees F.
Spent Fuel Pool Region 2:
USAR Section 9.1A.3.3, paragraph 2, does not indicate a minimum SFP temperature of 60 degrees F. It is inferred to be 60 degrees F, as discussed for Region 1. The minimum temperature can be reduced to 35 degrees F with no adverse effect on SFP criticality. Because the SFP is over-moderated at the temperature range in question, as evidenced by Figure 9.1A-23, reduction in temperature from 60 to 35 degrees F adds a negative reactivity. The reactivity added, based on Figure 9.1A-23, is as follows: A large negative reactivity (approximately -0.0010 delta K) from 60 degrees F to the maximum water density point, 39.2 degrees F, plus a very small positive reactivity (approximately 0.0002 delta K) from 39.2 to 35 degrees F.
There is no affect on the SFP or other equipment based on the review of the reactivity analyses. There is no cpecific discussion of the lowest acceptable temperature allowed in the SFP with the exception of the discussion of the reactivity analysis in USAR Sections 9.1A.2.2 and 9.1A.3.3. These sections are being revised and the acceptability of operating with 35 degrees F water in the SFP is being added to the discussion of the Spent Fuel. Pool Cooling and Cleanup System is being added to USAR Section 9.1.3.
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The only design bases accident pertinent to this change is related to the discussions of the reactivity analysis in USAR Sections 9.1A.2.2 and l 9.1A.3.3, and those sections are being revised by this package. The reactivity analysis was previously completed using a minimum water temperature of only 60 degrees F , Fowever, subsequent review and revision of the analysis showed 35 degrees r SFP water is acceptable.
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Based on the increase in negative reactivity being added to the SFP, there 1 l are no credible accidents created by this change. In fact, the I possibility of a criticality concern being created in the.SFP is reduced j
, as the temperature goes down, thus enhancing safety, i
This change is correcting a concern identified by Performance Improvement Request (PIR) 97-4602 that Plant Modification Request (PMR) 04380 i l overlooked reviewing the impact of reducing the temperature of the Component Cooling Water to 35 degrees F on the SFP reactivity analysis.
Therefore, based on the previous PMR and on the review and revision of the l reactivity analysis noted in this prckage, there are no credible malfunctions of equipment important to safety that may be directly or )
indirectly affected by this change.
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.There are no acceptance limits contained in the bases for the Technical j specifications affected by this change. The conditions required to meet j
, the Tech Spec have been reviewed aad they are also not affected by this l change. Before this change the K-effective is less than 0.95 for l unborated water, including all tolerances and uncertainties. Because the i l
! SFP is overmoderated the effect of going from 60 degrees F to 35 degrees F :
is a net negative reactivity addition. !
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l Page 27 of 217 Safety Evaluation: 59 98-0001 Revision: 0 l
l Addition of Security Buildingn The scope of modifications to be implemented under Design Change Package (DCP) 07393 is as follows:
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- 1. Install 3 new G foot by 12 foot security buildings and modify the one existing security building. The new buildings are to be generally located near each corner of the Turbine Building. Each building is to be equipped with electric power, telephone, Plant Gai-Tronics and a P.C. connected to the LAN.
- 2. Modify and slightly relocate portions of the Radiologically Controlled Area (RCA) fence to convert it to a " Floppy Fence" to allow the RCA fence to serve as a delaying barrier to potential intruders.
- 3. Install concertina wire on portions of the Protected Area Boundary (PAB) fence generally on the east-west runs which are due north of and due south of the Power Block.
- 4. Faoricate and install Door Safety Barriers (DSB) for 3 Turbine Building Doors. Door Safety Barriers consist of steel channel welded in a honeycomb pattern which is 8 feet high by 4 feet wide by 8 feet deep. The !
DSBs are placed in position by a fork truck and must be removed by fork truck. The DSBs serve to prevent intruders from placing explosive charges on the doors to gain entrance.
- 5. Remove the north-south portion of the masonry wall which runs south from the Walter P. Chrysler Support Bldg and remove 30 feet of the east end of the wall which runs east-west from the Turbine Bldg.
- 5. Install new security chain-link fence running north-south generally from the Health Physics Dosimetry Bldg. to the northwest corner of the RCA fence. This fence is to be installed just east of the north-south asphalt road which runs parallel with the west PAB fence.
This modification has been evaluated in accordance with 10 CFR 50,54(p).
As a result of this evaluation, it has been determined that this modification does not reduce the effectiveness of the Security Plan.
Because no desigt basis accidents are identified the probability of occurrence of an accident is not affected. Because no design basis accidents are identified, the consequences of accidents are not affected.
Because no malfunctions are identified the probability of occurrence of a malfunction is not affected. Because no malfunctions are identified the consequences of a malfunction are not affected. Because no credible accidents that could be created are identified no accidents of a different l l type can be created. Because no malfunctionc are identified no malfunctions of a different type can be created. Because no acceptance l
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Page 28 of 217 l t
limits are identified that could be affected, the margin of safety is not affected.
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Safety Evaluation 59 98-0003~ Revisiont0 Updated Safety Analysis Report Clarification The Updated Safety Analysis Report (USAR) Change request will correct USAR Figure 5.4-8 to show that Residual Heat Removal (RHR) Pump number 1 takes suction from Reactor Coolant System (RCS) hot 'eg loop 1 not 4. [
i Additionally, it will show on data sheet 6.1 of USAR Table 7A-3 that the range of the containment pressure inoicators is 0-69 psig not 0-60 psig.
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The subject changes are minor corrections to the USAR descriptions. The ;
USAR document is currently in error in how it reflects where the RHR number 1 pump takes suction f rom the RCS, and what the range of the -
containment pressure gauges are. Correcting these USAR document errors j will enable a true reflection of the plant. These minor changes to the }
A USAR do not increase the probability of an accident, malfunction or create the possibility of a new type of accident or malfunction because these minor changes don't affect the safety design features of the plant. i This revision will.nc. .. crease the probability of occurrence or the k consequences of an accident or malfunction of equipment important to i safety previously evaluated in the safety analysis report. This revision i does not create a possibility for an accident or malfunction of a l
[ different type than any evaluated previously in the safety analysis i report. The margin of safety, as defined in technical specifications, is !
not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question. i
Attachment II to ET 99-0003 Page 30 of 217 Safety Evaluation: 59 98-0004 Revision 0 Approved Fire Suppression Uses This Updated Safety Analysis Reoort (USAR) change modifies USAR section 9.5.1.2.3, Table 9.5A-1, and Table 9.5E-1 to identify that allowed uses of the fire suppression system include a water source for fire brigade training, and a back up raw water source for plant safe shut down for design basis accidents other than fire.
USAR Table 9.5.1-1 references NFPA 24 for the design and installation of the fire suppression system pumps and m.derground supply fire suppression piping. NFPA 24 (page 8, paragraphs b-7) states that "the use of hydrants and hose for purposes other than fire-related services shall be prohibited." USAR Table 9.5.1-1 also references NFPA 803 for fire protection of Nuclear Power Plants. NFPA 803 (page 20 paragraph 12-3) states that "the fire main system piping shall not serve service water system functions." The purpose of both the NFPA 24 and 803 requirements are to ensure that there is a dedicated automatic fire suppression system capable of meeting full demand flow.
The Wolf Creek Generating Station (WCGS) fire suppression system is a dedicated system which is designed and installed for fire suppression functions only. This change to the USAR documents that acceptable uses of the fire suppression system include a water source for fire brigade training, and a back up raw water source for plant safe shut down for design basis accidents other than fire.
Appendix R section III G requires that a single train of equipment required for safe shut down (SSD) be maintained free from fire damage in the event of a fire in any fire area. In some safety related areas of the plant, automatic water suppression systems are credited to maintain SSD equipment of circuits free from fire damage. To meet the Appendix R Section III.G requirements, automatic suppression capability for safety rclated areas must be maintained.
The fire suppression system is designed to supply 2800 gpm at a pressure of 80 psig at the furthest Sergeant & Lundy(S&L)/ Bechtel interface point. STN FP-204 establishes system operability at 2800 gpm at 80 psig at the furthest interfac point. The design demand is based on the largest system demand flow plus 500 gpm allowance for additional hose streams. The largest suppression systems are located in the Turbine Building and have a design demand of 2300 gpm. The largest suppression system demand in safety related areas is 1035 gpm, and it is applicable for several of the systems.
For the purposes of NFPA 24, the fire brigade training is being considered
" fire related service." As part of the fare brigade training cycle, fire brigade members are required to perform live fire training which includes
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i Attachment II to ET 99-0003 l Pr.ge'31 of 217 the use of a charged hose line. During the fire training evolutions, the l control room is notified and the electric fire pump is started. A maximum l of two 1-1/2 inch hose lines will be in use at any one time. The training i involves the use of two charged hoses with a hose in back up. The hose ;
lines are connected to hydrant 1FP617 by way of two 2-1/2 inch hoses then connected to gated wye's. Additionally, a single 2-1/2 inch nozzle may be f used. For training with foam, a foam maker which consists of a !
specialized eductor nozzle connected to a single 1-1/2 inch hose line is used. l The value for discharge flows from the actual fog nozzles being used at {
WCGS for fire fighting will be used. Hose nozzles are rated by their j
' discharge flow at 100 psi nozzle pressure. .A fog nozzle with a variable !
4 angle nozzle may be rated for either a steady flow rate or a variable flow '
rate based on discharge angle. The nozzles used at WCGS are identified j below. For the purposes of.this evaluation the maximum nozzle flows are: j 1-1/2 inch Akron Style 1716P Turbojet 125 gpm. f
- 2-1/2 inch Akron Style 2730 Turbojet 250 gpm. ;
1-1/2 inch Akron Style 4616 Marauder 125 gpm. l 2-1/2 inch Elkhart Style 4000-27 Chief 250 gpm. !
Based on the nozzle ratings, the flow from either the two, 1-1/2 inch !
nozzles or the single 2-1/2 inch nozzle will not exceed 250 gpm. The flow l through the foam eductor will be less than 250 gpm. !
The actual pump performance is measured annually by conducting procedure !
STN FP-209. Based on the pump curve, it is assumed that in the flow range !
of 2000 to 4000 gpm, the curve'is basically a straight line. In that flow l region, the flow / pressure relationship (slope of the line) in ;
approximately 25 gpm of flow per 1 psi of change in pump head. The margin of 15 psi pressure identified at the SL/Bechtel interface point would l
equate to a margin of approximately 375 gpm of flow capability. The estimated flow produced by the diesel driven fire pump at an furthest SL/Bechtel interface point can be estimated as 2922 + 375 - 3297 gpm. The 3300 gpm flow value is consistent with previous years STN FP+204 test j performance for the diesel driven fire pump where flows of 3300 gpm or greater at 80 psi were recorded.
Based on the above set of assumptions and engineering judgments, the fire i brigade training hose discharges will not result in the fire suppression j being inoperable, and the suppression system will be capable of providing an' automatic flow of 2000 gpm at 80 psi at the furthest SL/Bechtel ,
interface point in the event of a fire. Because the fire s2ppression l system demands in safety related areas are less than the fire suppression !
system demands used to establish operability, fire safe shut down can be .
achieved. !
The fire suppression system is utilized as a source of raw water for cooling and make up water in va 2ous "Off Normal" and Emergency procedures j i
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Attachment II to ET 99-C 003 ,
Page 32 of 217 I to support reactor safe shut down and to maintsin shut down conditions. A fire coincident with any other design basis accident is considered out l side the plant design basis. As used in the "Off Normal" and Emergency l procedures, the fire suppression system is used as a secondary source of !
water when other safety related equipment is not available. l
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Currently the USAR does not describe any of the fire suppression system l uses other than fire protection.
i This USAR change revises USAR section 9.5.1.2.3, Table 9.5A-1, and Table l
9.5E-1 to identify that allowed uses of the fire suppression system l include a water saurce for fire brigade training, and a back up raw water ,
source for plant safe shut down for design basis accidents other than fire. !
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Because no design basis accidents are identified the probability of l occurrence of an accident is not affected. Because no design basis i accidents are idizatified, the consequences of accidents are not affected. -
Because no malfunctions are identified the probability of occurrence of a malfunction is not affected. Because no malfunctions are identified the consequences of a malfunction are not affected. Because no credible :
accidents that could be created are identified no accidents of a different !
i type can be created. Because no malfunctions are identified no j malfunctions of a different type can be created. Because no acceptance l
limits are identified that could be affected, the margin of safety is not affected. ;
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Page 33 of 217 I l
t Safety Evaluation: 59 98-0005 Revision 0 l Computer Setpoint Change for Essential Service Water Room Temperature .
Alarms .l The current computer alarms for Essential Service Water (ESW) pump house l room temperature are set at the following values: Low Low 40 degrees Fahrenheit (F), Low 65 degrees F, High 119 degrees F, and High High 149
{
r degrees F. USAR Table 3.11(B)-1, identifies the minimum and maximum i temperatures for these rooms as 50/122 degrees F. l Unreviewed Safety Question Determination (USQD) 59 96-0116 evaluated the f equipment in these rooms remaining operabla for a temperature of 45 i degrees F or above. The Low Low alarm will be changed to 45 degrees F l based on the evaluation performed in Performance Improvement Request (PIR) l 96-0205. This alarm will require actions be taken to increase the i temperature in the room since the equipment has not been evaluated for an i environment below 45 degrees F.
l The Low alarm will be changed to 50 degrees F. This will inform the control room that the ventilation system is not working properly and it needs to be monitored. The basis for 50 degrees F is the minimum design !
temperature of the room. [
t The High alarm will conservatively remain at 119 degrees F. This is based i on the design temperature of 122 degrees F minus 3 degrees for !
conservatism. The USAR requires the room temperature to be lowered when ;
it reaches 119 degrees F. ;
i The High High alarm will be deleted since 149 degrees F has no real basis. Actions will be taken at the High alarm and no additional actions ,
- will be required after the High alarm is reached. j The USAR incorrectly states that the heaters in the pumprooms come on at +
50 degrees F and turn off at 55 degrees F. The heaters are set at 52 + 2 l degrees F and will turn off 1 degree above where they are set. This [
statement will be deleted from the USAR since it adds no value to the i required function of the heaters. The only requirement of the heaters is to ensure the pumproom stays 50 degrees F or above. ,
i This computer'setpoint change will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
This revision does not create a possibility for an accident or malfunction ;
of a different type than any evaluated previously in the safety analysis :
report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.
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Attachment II to ET 99-0003 Page 34 of 217 Safety Evaluation: 59 98-0008 Revision 0 Security Camera Rotation This change consists of rotating the head of Security Closed Circuit Television (CCTV) Camera 54 to give the camera a new field of view. This modification is accomplished to enhance Wolf Creek Security and implement the new physical security philosophy.
This change will not increase the probability of an accident previously evaluated in the Updated Safety Analysis Report (USAR). There are no accident scenarios contained in the USAR which include Security CCTV cameras.
This change will not increase the consequences of a radiological accident previously evaluated in the USAR. There are no accident scenarios contained in the USAR which include Security CCTV cameras.
This change w1A1 not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR. The plant Protected Area Boundary (PAB) Security CCTV cameras do not interface with any safety related equipment or equipment important to safety.
This change will not increase the radiological consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The plant PAB Security CCTV cameras do not interface with any safety-related equipment or equipment important to safety.
This change will not create the possibility of an accident different from that which was previously evaluated in the USAR. The plant PAB Security CCTV cameras do not interface with or have any common functions which could impact on or create an accident different than was previously evaluated in the USAR.
This change will not create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the USAR, The plant PAB Security CCTV cameras do not interface with equipment important to safety. Rotation of CCTV camera 54 will not impact any plant equipment which is safety-related or important to safety.
This change does not reduce the margin of safety as defined in the basis for any Technical Specifications. Rotation of CCTV camera does not affect any equipment which could reduce the margin of safety or to impact the plant Technical Specifications.
Attachmer.t II to ET 99-0003 Page 35 of 217 Safety Evaluation: 59 98-0009 Revision 0 Drip Pan Installation This change package approves the prior installation of a drip pan mounted to the Process Sampling S'/ stem (RM) Constant Head Chamber assembly and field rerouting and supporting the three fourths inch drain line that runs from the drip pan to Turbine Building oily waste floor drain (LE-592),
located along the North Wall of the Secondary Chemistry Labratory Building. The Constant Head Chamber assembly is located above the roof of the Secondary Chemistry Labratory, in the Turbine Building, some 25 feet above the sampling station. The initiating issue was that the drain line was inadequately supported.
Contrary to the intent of the original design, the constant chamber vents overflow the sample fluid frequently. Estimates of some flows from the drain, based on intermittent walkdowns of the drain line, appear to be as much as 1 gpm. The flow from the vents has been directed to LE-592 since startup. If provisions were not made to collect and transport to a nearby drain, the fluid would flow onto the top of the roof of the Secondary Chemistry Lab Building, where the RM System Constant Head Chamber is located.
Approximately midway up each of the Constant Head Chambers is an overflow. When fluid is at the overflow level, constant pressure is ;
assured at the sampling station. The sample pressure so provided, assures i a constant sample flow rate for the analysis equipment. ;
I There are fifteen Constant Head Chambers. The overflow lines are routed to one of two drain headers or Turbine Building oily waste floor drain (LE-592). One drain header, 002-HCD-2, is designed to receive the overflow from the four Steam Generator Blowdown Constant Head Chambers and direct it to the Steam Generator Blowdown Sample Recovery Tank (TRM01) .
Radiation levels of the Steam Generator Blowdown samples are monitored continuously at the Nuclear Sampling Station and are automatically isolated on high radiation signal. The overflow from TRM01 and the drain from the sample sink, drain to the Turbine Building Oily Waste System.
The overflow from the circulating water constant head chamber is routed directly to LE-592. The other overflows (Condenser hotwells, Condensate pump discharge header, Heater Drain pump discharge header, Steam Generator Feedwater, Main Steam lines, Condensate Demineralizer outlet, L.P.
Feedwater heater outlet, Demineralized water degasifier inlet and outlet, and the Moisture Separator reheater drains) are routed to the Miscellaneous Condensate Drain Tank (TLE02).
The Constant Head Chambers are not designed for fluid flow through the vent lines. Routing the vent line flow to the common drain LE-592 is contrary to the flow path of the Constant Head Chamber overflows. The difference, summarized from above, is that all vent flows are directed to
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f Attachment II to ET 99-0003 Page 36 of 217 i LE-592, unlike the overflows that are directed to TRM01 and TLE02. The bases for.the acceptance is that both the overflow from TRM01 and the tank i TLE02 are Secondary Liquid waste (LKW), and consistent with 9.3.2.2.3 of l
the USAR that describes that the sample line and sample sink drains in the ;
Process Sampling System are collected in the secondary liquid waste system !
where they are processed for reuse. In addition, since the sample sink [
drains to the LE system, it is acceptable to route the vents to the LE '
system. l The drip pan and drain line are not original design by the vendor and ,
there is no documentation of the installation. The Constant Head Chamber portion of the RM System is non safety-related. 'The drip pan and drain line have no impact on the operation of the system. The drip pan and I drain line,are open to the atmosphere, and there is no potential for j pressurizing. The drip pan and drain line were installed only as a house :
keeping measure. l l
The change described above requires that USAR Figures 9.3-4-01 and 9.3 !
02 be revised to reflect the drip pan and drain line. In addition, the [
Identification of the drain line from the Steam Generator Blowdown l Constant Head Chambers to the Steam Generator blowdown recovery tank shown !
on USAR Figure 9.3-4-01 will '1xn corrected. l Because no design basis accidents are impacted, the probability of occurrence of an accident, and the consequences of any accident are not affected. Because no malfunctions are impacted, the probability of !
occurrence and the consequences of a malfunction are not affected. l' Because there are no credible accidents that could be created, no accidents of a different type can be created. Because no malfunctions are identified, no malfunctions of a different type can be created. Because l no acceptance limits affected, the margin of safety is not affected.
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Attachment II to ET 99-0003 Page 37 of 217 f
Safety Evaluation: 59 98-0011 Revision 0 Addition of Combustible Katerial in the Auxiliary Building.
This modification adds combustible material in the Auxiliary Building. The combustible material and the fixed combustible fire load identified in USAR Appendix 9.5B for fire area A-1, room 1102 is being updated to reflect the addition of the combustible material which accounts for 95 Btu's/Sq.Ft. of additional combustible load. This modification added a Security feature in the Auxiliary Building. Details of this modification j are considered " Safeguards Information," and are available for review at i Wolf Creek Generating Station. !
i A fire in any portion of Fire Area A-1, including 1102 will not prevent i safe shutdown of the plant. Additionally, the combustible loading in room ;
1102 will be maintained well below that required to challenge the three hour rated fire barriers surrounding room 1102. The probability of a fire in room 1102 in unchanged as the combustible material does not introduce l any new significant fire hazard or ignition source. i The inputs, assumptions, and analysis of design basis accidents other than fire are not affected by this change. This modification will not -
increase the probability of occurrence or the consequences of an accident I or malfunction of equipment important to safety previously evaluated in (
the safety analysis report. This modification does not create a !
possibility for an accident or malfunction of a different type than any y evaluated previously in the safety analysis report. The margin of safety, l as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed ,
safety question. j I
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Attachment II to ET 99-0003 Page 38 of 217 1
Safety Evaluation: 59 98-0012 Revision: 0 4
Corrision Product Monitoring Modification This modification is implemented because there are not adequate sampling paths from the condensate and feed water systems to monitor iron corrosion 4 products. Measurement of iron corrosion products will help identify the sources and processes which promote iron transport. The operational goal is to minimize iron transport to the steam generators thereby minimizing the amount of corrosion products deposited.
Design Change Package (DCP) 07417 replaces an existing capped nipple on the common condensate discharge line to the low pressure heaters with an isolation valve. This is the extent of work requiring plant shutdown for installation.
No detrimental effect on the function and operation of the condensate system is expected from the addition of this isolation valve. When it is closed, it will maintain the integrity of the system pressure boundary, in the same way as any other normally closed vent or drain valve. When it is open, it will allow a small amount of condensate to flow to the corrosion monitoring cabinet, similar to other sample taps from the condensate and feed water system. Sampling the condensate and feed water is not a new activity. The systems have been and are being sampled routinely, but under limiting conditions.
The condensate discharge line ta the low pressure heaters is a quality class 'D' line, constructed to ANSI B31.1 standard. It is a non nuclear
' safety related piping. The design pressure of this line is 650 psig, the design temperature is 175 degrees F. The normal pressure is 436 psig and the normal temperature is 127 F. -The discharge line is non seismic, and it is not environmentally qualified. The new sample line will be constructed to the same standards.
The new isolation valve complies with existing standards. Existing specification M-234 is designated for this application. Existing drawing M-234-00162 and referenced data sheet M-234-SD-001 describe the valve.
This type of valve is commonly used in other similar applications. The new valve is a Vogt, class 800, forged carbon steel, globe valve. Its assigned configuration identifier is ADV0488 The 3/4 inch pipe nipple upstream of the valve is ASTM A106 grade B, carbon steel, schedule 80 piping. Immediately downstream of the isolation valve, a transition is made to stainless steel tubing.
This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfunction
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Attachment II to ET 99-0003 Page 39 of 217 of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is ,
not reduced by this modification. Therefore, this modification does not t 4
involve any unreviewed safety question.
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Attachment II to ET 99-0003 Page 40 of 217 Safety Evaluation: 59 98-0013 Revision: 0 Turbine Building Process Sampling Room Modification Design Change Package (DCP) 06543, Revision 1, supersedes DCP 06543, Revision 0, in its entirety DCP 06543, Revision 1, performs several modifications in the Turbine Building Process Sampling Room and is evaluated by this Unresolved Safety Question Determination.
The air pumps are removed from the Hydrazine Analyzer (RMAIT0164) and the Sodium Analyzers (RMAIT0019, RMAIT0112, RMAIT0130, RMAIT0137, RMAIT0155, RMAIT0477 and RMAIT0567) on panel RM172. The life of an air pump is approximately one year. Instrument air is routed to the analyzers to perform the functions previously performed by the air pumps.
The Beckman/Rosemount cation filters FRM645 through FRM653 on panel RM172 are replaced with Sentry Equipment Co. cation filters. This change was requested by Chemistry to better utilize the Sentry resin refills previously approved on CCP 07337.
A 120 VAC electrical power circuit, 12 process sample lines and instrument air are routed to the west end of panel RM172 to support the use of an ion chromatograph (IC) by Chemistry. The 120 VAC power will be routed from circuit 20 in the pInnt lighting panel QA01. The additional process sample and instrument air lines are merely extensions of existing lines already in panel RM172. A bottle rack must also be installed near the panel to hold helium and/or nitrogen bottles for the ion chromatograph.
Revision 1 to DCP 06543 will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfunction
, of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.
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Attachment II to ET 99-0003
- Page 41 of. 217 Safety Evaluation: 59 98-0014 Revision 0 Updated Safety Analysis Report. Changes to Section 4.4 This revision to Updated Safety Analysis Report (USAR) corrects discrepancies identified in Self Assessment 96-040. The changes delete references in USAR Section 4.4.2.1, 4.4.3.5, 4.4.4.4, and Table 4.4-2 to 3-
- loop operation at power since Wolf Creek Generating Station (WCGS) is not
- licensed to operate in this mode. Section 4.4.2 5 is revised to indicate that enthalpy distributions shown in Figures 4.4-5 through 4,4-7 are for first core only. The Table 4.4-1 limiting value of kW/ft for centerline s fuel melt is revised to reflect the current analyzed limit.
Section 4.4.4.3.2 is revised to reflect the use of relaxed axial offset
, control (RAOC) versus constant axial offset control (CAOC) at WCGS. A number of references and Figures in USAR section 4.4 are revised to l reflect the current licensed core thermal-hydraulic analysis methodologies
- employed by the reload design.
! Because no design basis accidents are impacted, the probability of occurrence of an accident, and the consequences of any accident are not
. affected. Because no malfunctions are impacted, the probability of occurrence and the consequences of a malfunction are not affected.
Because there are no credible accidents that could be created, no accidents of a different type can be created. Because no malfunctions are
' identified, no malfunctions of a different type can be created. Because
- no acceptance limits affected, the-margin of safety is not affected.
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Attachment II to ET 99-0003 Page 42 of 217 1
I Safety Evaluation: 59 98-0015 Revision 3 0 Power Spike Factor This revision to the Updated Safety Analysis Report (USAR) provides changes in the following categories: l
- 1. Editorial changes
- 2. Changes reviewed and approved by the NRC as license amendments
- 3. Changes to core design / safety analysis parameters made under the l control of the Reload Safety Evaluation (RSE) reload design process.
Core neutronic design analyses are performed with the Westinghouse code set- Application of the cross section generators, core nodal simulators, and auxiliary codes has been fully documented. Use of the code by WCNOC, and the associated modeling assumptions, has been reviewed and approved by the NRC.
This USAR change addresses changes to the USAR to reflect the current core design methodology. These changes to the core design methodology were incorporated under the control of the Reload Safety Evaluation (RSE) reload design process. Specifically, this USAR change reflects the elimination ;f the " power spike factor" in pin peaking analysis. Modern fuel manufacturing techniques have eliminated the need for a power spike factor and the USAR is being updated to reflect this change to Wolf Creek analytic methods.
Reactor Engineering performed a self assessment to evaluate the accuracy of the USAR with respect to Core Design data and methodologies employed at WCGS. One of the items identified was a difference in the current pin peaking methodology with respect to the applicability of a power spike factor. Wolf Creek employs the Westinghouse core design methodologies and in March, 1995, via WCAP 13589-A, the Westinghouse methodology was modified to eliminate the peaking penalty resulting from the power spike factor.
The removal of the power spike factor is based on data and analysis contained in WCAP-13589-A, " Assessment of Clad Flattening and Densification Power Spike Factor Elimination in Westinghouse Nuclear Fuel". The analysis was produced by Westinghouse in 1995.
Prior to WCAP 13589-A the power spike factor was used to account for the increase in pin peaking due to fuel densification. A brief explanation of the fuel densification effect on pin peaking follows. Fuel densification which causes fuel pellets to shrink both axially and radially, combined with random hang up of fuel pellets can result in gaps in the fuel column. The gaps arise due to settling of fuel pellets below the hung up pellet. Due to decreased neutron absorption in the vicinity of the gaps in the fuel column, increased power peaking results in the adjacent fuel reds. This increased peaking resultant from fuel densification was l
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Attachment II to ET 99-0003 Page 43 of 217 accounted for in core design analysis methodologies through t he use of a ,
power spike factor.
The fuel densification effect (and power spike factor) as described in the USAR was applicable to Westinghouse fuel manufactured under older generation fabrication techniques. However, WCAP 13589-A concludes that current generation Westinghouse fuel is highly stable with respect to fuel densification. This has been verified by measurement of axial gaps which form in the fuel column. In-core flux traces have also been examined to ,
detect' flux spikes which could occur if larger axial gaps were present in the fuel column. Measured data confirm that axial fuel column gaps due to fuel densificaticn are very infrequent and extremely small. In fact the ,
gaps are small enough that local power peaking is not significantly impacted. Therefore, WCAP 13589 concludes that a densification spiko factor of 1.0 is appropriate for current generation, stable Westinghouse fuel.
After WCAP 13589-A was approved to reflect the elimination of the power spike factor consistent with current generation Westinghouse fuel manufacture and performance Wolf Creek incorporated this approved core design methodology. The USAR is being changed to reflect the current densification power spike factor methodology.
The change to the USAR to reflect the current pin peaking methodology with '
respect to the fuel densification spike factor has no impact on analyses which establishes spent fuel fission product inventories for use in determination of radiological consequences of postulated events. The above changes will improve the accuracy and clarity of the USAR and have ;
been previously determined to be acceptable.
The accident analysis was reviewed for potential impacts. The changes )
impact the maximum FQ calculated in the cycle specific evaluation of the over power peaking transients. The Chapter 15 overpower transients I discuss maximum FQ in terms of being less than an FQ which results in peak kw/ft limits less than the fuel melt limit. The elimination of the spike factor does not increase the peak FQ. Therefore, the results of the bounding overpower transients analyzed in Chapter 15 are not affected.
The proposed USAR changes will not create any credible accidents which have not already been addressed by the original FSAR, USAR, or revisions already addressed in the USAR. All design and performance critoria continue to be met and no new failure modes have been introduced for any system, component, or piece of equipment as a result of the changes. The changes will not impact either the normal plant operation or the response -
to accident conditions. The change in core design methodology does not involve a physical change to any physical structure, component, or ;
equipment. )
I The changes do not result in a different response of safety related l systems and components to cecident scenarios than those described in the !
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i Page 44 of 217 l USAR. No new equipment malfunctions have been introduced that will affect !
fission product barrier integrity. There are no new credible accidents t associated with the USAR changes. The changes do not affect the safety ;
function of safety related systems and components which are related to .
accident mitigation. Therefore, the changes will not create the !
4 possibility of a malfunction of equipment important to safety different l than those already described in the USAR. ;
l The proposed USAR changes do not affect any acceptance limits which are }
contained in the bases for the Technical Specifications or license bases documents. This evaluation shows that all design and safety analysis '
limits continue to be met and that these limits are supported by the ,
applicable Technical Specifications. The margin of safety as defined in !
the BASES is not reduced for any USAR accident, no new accidents are created, and no new malfunctions of equipment important to safety are -j created and therefore, the margin of safety has not been reduced. There ,
is no impact on spent fuel fission product inventories as described in the )
USAR and thus, there is no impact on the radiological consequences to any :
postulated event. All acceptance limits continue to be met. l 1
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Safety Evaluation: 59 98-0016 Revision 0 Spent Fuel Pool Boron Concentration j l
This change revises sections of the U pdated Safety Analysis Report (USAR) j to apply consistency between sections concerning Spent Fuel Pool (SFP) !
Boron concentration and makeup to the spent fuel pool. The SFP Boron l
concentration is required to be no less than 2000 ppm by design. Numerous >
sections have +/- 50 or specify 2000 PPM. Also, make up to the SFP is j through the reactor water makeup which is not Borated. One USAR section ,
states that Borated Demineralized water is used to makeup SFP water inventory. This conflicts with another section saying that normal makeup i of the SFP.is through the reactor makeup water system (which is demineralized water).
l These changes are not related to any changes in operating procedures or l requirements for regulating boron concentration in the SFP water of SFP l
makeup water. These changes will improve the accuracy and clarity of the ,
USAR. The accident analysis was reviewed for potential impact. The !
original assumptions for accident conditions are not being changed. The l consistent use of a minimum boron concentration of 2000 parts per million ]
(ppm) will not effect the accidents discussed in any of the USAR sections.
1 Because no design basis accidents are impacted, the probability of occurrence of an accident, and the consequences of any accident are not affected. Because no malfunctions are impacted, the probability of {
occurrence and the consequences of a malfunction are not affected. <
Because there are no credible accidents that could be created, no l accidents of a different type can be created. Because no malfunctions are identified, no malfunctions of a different type can be created. Because ]
no acceptance limits affected, the margin of safety is not affected.
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Safety Evaluation 59 98-0017 Revision: 0 i i
Fuel Handling Tool :
This revision to the Updated Safety Analysis Report (USAR) adds wording to reflect that the new fuel handling tool is used in conjunction with the i monorail hoist to move new fuel. The USAR states that new fuel is moved I from storage to the new fuel elevator with the monorail hoist. In ;
addition, words are being added to reflect that new fuel can be delivered straight to the Spent Fuel Pool (SFP) , which is consistent with other USAR ;
sections and Technical Specifications. This change also clarifies !
statements in USAR section 9.1 to provide more complete descriptions. l These changes improve the accuracy and clarity of the USAR. No fuel l handling procedures are affected by this change. ;
I The accident analysis was reviewed for potential impacts. The Fuel l Handling Accident is the only design basis accident that could be impacted i by this change. The Fuel Handling Accident is not impacted by this change [
because: 'the method by which the fuel is moved, the frequency of the I movement, and the number of fuel assemblies moved in conjunction with one }
another. No other fuel handling features are changed by this revision. l These changes to the USAR will not create any credible accidents which have not been addressed by the original Final Safety Analysis Report, ;
USAR, or revisions already addressed in the USAR. All design and ;
performance criteria continue to be met and no new failure modes have been '
introduced for any system, component, or piece of equipment as a result of l these changes. Normal plant operation and accident response are not !
affected. [
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Attachment II to ET 99-0003 !
Page 47 of 217 Safety Evaluation: 59 98-0018 Revision: 0 '
Clarification to' Updated Safety Analysis Report Section 9.2 i
This change will revise Updated Safety Analysis Report (USAR) Section
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9.2.2.2.2 to clarify a statement found in the second to last paragraph of this section that currently reads as follows: The normally closed (with power lockout) parallel sets of containment isolation valves will allow the operator to establish cooling water to the Reactor Coolant Pumps and l the excess letdown heat exchanger under emergency conditions, with a j single' failure. !
l This USAR sentence has been interpreted by some, including the AE NRC ,
inspection team, to imply that the above valves can be opened under a LOCA ;
or MSLB accident if the operator desires to maintain cooling water to the ;
Reactor Coolant Pumps (RCPs) and the excess letdown heat exchanger. This interpretation is -in direct contradiction with USAR Table 3.11(B)-3 that reflects that no credit is taken for the function of these valves during the Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB) !
accidents. USAR Table 3.11(B)-3 is a listing of safety related equipment l and specific components required for a design basis accident (DBA) and/or l safe shutdown of the plant. This Table does reflect that these valves can !
be used to bring the plant to a hot and cold shutdown condition. i The.following changes and additions to USAR Section 9.2.2.2.2 have been !
proposed by this USAR Change Request to clarify this USAR Section. The expected affects of these changes is that it will clear up any confusion- l that the old statement can generate especially when it is compared to USAR l Table 3.11(B)-3. The old sentence has been nodified and clarified by ;
explaining that the normally closed (with power lockout) parallel sets of i containment isolation valves in an emergency, where its normal valve can ,
not be recovered from its closure position (single failure), can be opened by the operator to establish cooling water to the Reacto'r Coolant Pumps
~(RCPs) and the excess letdown heat exchanger. The bypass valves are not ;
credited for operation under the LOCA or MSLB conditions as reflected in USAR Table 3.11(B)-3 because; (1) no single failure can disable both means I of cooling the RCP seals (Component Cooling Water cooling to the thermal i barrier or seal injection), (2) the excess letdown flow path is not required post LOCA or post MSLB and (3) the valves are closed with power ;
lockout preserving the integrity of their containment penetration. The '
bypass valves are assumed to be closed during all accidents, reference i Technical Specification basis 3/4.6.3. However, these valves may be !
utilized when recovering from a LOCA or MSLB even though they are not required, because they have the same environmental qualifications as the normal valves. The bypass valves are installed to provide the operator flexibility to maintain Component Cooling Water (CCW) cooling to I containment if desired should its normal valve not recover from its closure.
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Attachment II to ET 99-0003 Page 48 of 217 This USAR clarificattion only affects the USAR document. Clarifying the USAR document will enable a better interpretation of the USAR. This clarification to the USAR does not increase the probability of an accident, malfunction or create the possibility of a new type of accident or malfunction because the clarification does not affect the safety design features of the plant.
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Attachment II to ET 99-0003 i
Page 49 of 217 Safety Evaluation: 59 98-0019' Revision: O Process Sampling System Alarm Setpoints This change to the Process Sampling System (pSS) alarm setpoints in Procedure ALR 412 will reflect current operating conditions. The PSS is a non safety-related subsystem of the Plant Sampling System which is i described in Up dated Safety Analysis Report (USAR) Section 9.3.2. The PSS is designed to continuously monitor water samples from the turbine cycle ;
and the circulating water system. Measurement of PH, Conductivity Levels, ;
Dissolved 02, Residual Hydrazine and Sodium Concentration are obtained from water quality analysis and used to control water chemistry and to permit appropriate corrective action by the operating staff. The proposed ,
changes to alarm setpoints will enhance Operation's ability to meet the design basis function of the PSS. Various alarm setpoints used as action ,
levels in Procedure ALR 412, "RM Panel Alarm Response," will be different
!- than alarm setpoints listed in USAR Table 9.3-5, " Process Sampling System Sample Point design Data." 1 The PSS alarm point changes will not impact any of the accidents discussed in the above chapter. These alarm setpoints for the PSS cannot cause or are not used to mitigate any of the accidents described in the USAR.
The proposed alarm setpoint changes for the PSS are not accident
- initiators and will not create any credible accidents. The proposed alarm setpoint changes for the PSS can not create a malfunction of equipment 3 important to safety. These alarms provide the operations staff indication-of declining turbine cycle water or circulating water quality so that ,
appropriate corrective action can be taken. ;
The acceptance limits contained in the bases for the technical specifications or other licensing documents are not affected by the ,
proposed procedure change. Alarm set points are changed for the non- ,
safety related Process Sampling System which monitors water samples from the turbine cycle and the circulating water system. l I
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Attachment II to ET 99-0003 Page 50 of 217 Safety Evaluation: 59 98-0020 Revision: 0 Evaluation of Thermal Relief Valve During a chemical and Volume Control (BG) system walkdown it was noted that the 1" piping downstream of thermal relief valve BGV0020 (Letdown Heat Exchanger EBG01 Outlet Relief Valve) was essentially at the same temperature as the outlet Component Cooling Water temperature (approximately 140 degrees Fahrenheit). Considering the length of this piping, it appears that this valve is leaking through continuously.
Tha thermal relief valve, (BGV0020) on the Letdown Heat Exchanger, (EBG01) is deleted by this change package. Thermal relief valves are provided to protect components from excessive pressure due to trapped fluid.
Therefore, when the component is in service a thermal relief valve is not required. In addition, the isolation of the heat exchanger is controlled by strict administrative procedures. Therefore, the thermal relief valve does not perform a safety function during operable modes and is not required by the ASME Section III, Class 3 Code. Anytime the heat exchanger is isolated, it is recommended that it should be either drained or vented to atmosphere to prevent the possibility of over pressure from a thermal transient event. Updated Safety Analysis Report (USAR) Fig. 9.3 02 (P&ID M-12BG02) is revised to incorporate the above change.
The thermal relief valve will be replaced with blind flanges, piping spool pieces and elbow. Other than these changes, there are no procedures, activities, administrative controls, or sequences of plant operations; or any plant structures, systems, components or equipment; or any requirements outlined, summarized or describeu in the USAR, which would make information in the USAR no longer true or accurate or would violate a requirement stated in the USAR. There are no tests or experiments not described in the USAR which may adversely affect the adequacy of SSCs to prevent accident or mitigate the consequences of an accident.
The thermal relief valve performs no safety function during operable modes and only provided pressure boundary integrity of the component cooling water. The blind flanges configuration will provide the same level of design margin as the relief valve body. Anytime the heat exchanger is isolated it is recommended that the heat exchanger should be either drained or vented to atmosphere to prevent the possibility of over pressure from a thermal transient event. The proposed activity can not create potential impact to design bases accidents as identified, discussed or referenced in USAR chapter 2, 3, 6, 9 or 15.
Replacement of the thermal relief valve by the blind flange assembly could not create any type of credible accidents because it does not perform any safety function during operable modes. Anytime the heat exchanger is isolated it-is recommended that the heat exchanger should be either drained or vented to atmosphere to prevent the possibility of over
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Attachment II to ET 99-0003 Page 51 of 217 pressure from a thermal transient event.
Replacement of the thermal relief valve by the blind flange assembly could not directly or indirectly affect any type of credible malfunctions of equipment important to safety because it does not perform any safety function during operable modes. Anytime the heat exchanger is isolated it is recommended that the heat exchanger should be either drained or vented to atmosphere to prevent the possibility of over , pressure from a thermal transient event.
The proposed USAR change will not affect the acceptable limits of any technical specifications.
Attachment II to ET 99-0003 Page 52 of 217 Safety Evaluation: 59 98-0021 Revision 0 Updated Safety Analysis Report Revision to Relative Humidity Range This change revises Updated Safety Analysis Report (USAR) Section 9.4 and Table 3.11 (B) -1. Section 9.4 is being revised to remove the specific range for relative humidity. The specific range for relative humidity is already listed in Table 3.11(B)-1. Table 3.11(B)-1 is also being revised to decrease the lower range of normal relative humidity to encompass a range accepted in NUREG.0700, as reflected in Section 6.1.5, Control Room Workspace, Environment. The existing low end in the USAR Table is 30 percent for the control re?m and associated rooms. NUREG 0700 reviews plants with acceptance of the relative humidity for control room operator comfort as low as 20 percent. Therefore, the USAR changes are within published NRC guidance. Wolf Creek Generating Station (WCGS) is not committed to NUREG 0700; however, it has been used in the past as a guide to Human Factors considerations in the Control Room.
The issue of operator comfort is addressed in NUREG 0737, lessons learned from TMI-2. There are no expected negative affects of this change as the normal relative humidity is a subjective standard that will be limited to the control room and associated rooms. Any set of standards of this type are potentially subject to disagreement by individuals as to " personal Comfort."
The normal upper bound of 70 percent relative humidity is not under consideration for change as part of this change to the USAR. That normal upper limit was part of the originally accepted license for WCGS.
Lowering the acceptance bounds for normal relative humidity will have no affect on procedures, activities, administrative controls, sequences of plant operations: any plant structures, systems, components, or equipment; or any requirements outlined, summarized or described in the USAR Section
- 9. 4 and Table 3.11 (B) -1.
This change will make no information in the USAR untrue or inaccurate nor will it violate a requirement stated in the USAR. There are no tests or experiments associated with the proposed change. There are no design bases accidents discussed or referenced in the USAR chapters 2, 3, 6, 9, or 15 potentially impacted by the proposed activity. Altering the acceptance range for the normal minimum relative humidity in the control room while remaining within guideline limits as published by the NRC will have no affect on any other components, safety related or otherwise.
No credible accidents will be created by the proposed activity. As part of the concerns with operator comfort, the possibility of experiencing minor shocks from static electric discharges was considered. Maintaining acceptance limits within the published guidelines should minimize any static electric shocks. This change has not significantly changed the conditions that produce static electricity. Therefore, by minimizing or
I Attachment II to ET 99-0003 l Page 53 of 217 eliminating static shocks, there are no concerns with spurious signals being generated from the control room or associated rooms.
There are no credible malfunctions of equipment important to safety which may be directly or indirectly affected by the proposed activity. The proposed changes are to streamline the USAR and to widen the range of acceptability for normal relative humidity in the control room and adjacent rooms. To supplement the humidification process in the central Alarm Station area, humidity is maintained with the aid of portable humidifiers. There are no acceptance limits contained in the bases for the Technical Specifications for licensing basis documents that could be affected by the proposed activity. These changes will have a minimal effect on control room operator comfort and are within established guidelines.
l Attachment II to ET 99-0003 Page 54 of 217 Safety Evaluation: 59 98-0022 Revicion 0 !
Establish a New Flow Path for the Steam Generator Blow Down Process The steam generator blow down mixed bed demineralizer (SGBDMBD) resins l must be disposed of as radioactive waste. This is because no regenerative I process is available for these demineralizers. After this modification is complete, the Steam Generator Blowdown System (BM) process fluid will by- t pass the SGBDMBD's and be sent to the Condensate Demineralizer (AK) l vessels.
l This modification will establish a new flow path for the steam generator ,
blow down process fluid. This new flow path will uce the AK System ;
demineralizer vessels for water clean up instead of using the steam ;
generator blow down mixed bed demineralizers. By using this new flow c path, radioactive waste disposal will be reduced. These resins retain !
trace amounts of radioactive nuclides. The present Stiam Generator Blow i Down (BM) system configuration has no way of regenerating these resins. -l Therefore, they become radioactive solid waste.
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this change will not affect radiological barriers. The D-augmented section of isolation valves is unchanged. BM process fluid is still -
monitored and controlled by radiation monitor RE-25 which shuts down BM i flow if a primary to secondary tube leak occurs.
This modification will not increase the probability of occurrence or the l consequences of an accident or malfunction of equipment important to. l l safety previously evaluated in the safety analysis report. This (
modification doer. not create a possibility for an accident or malfunction !
of a different t)pe than any evaluated previously in the safety analysis l report. The marcin of safety, as defined in technical specifications, is !
not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question. !
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y Attachment II to ET 9S-0003 Page 55 of 217 Safety Evaluation: 59 98-0026 Revision 0 Temporary Modification to Fire Protection Valves Valves 1FPO49 and 1FPO43 are fire protection system yard loop isolation valves which are electrically supervised by position switches. The ,
circuit (or zone) alarms to the control room KC08 panel in the event the !
normally open valves are not in the full open position. !
1 The zone reports to local panel KC324 then on to KCOB. At KC08, an intermittent trouble signal "KC324 Earth Ground" was being received. This {
Temporary Modification order (TMo) is initiated to remove the affected ,
zone from the KC324 and KCOB panels until the final resolution is i implemented. The intermittent ground signal is a nuisance alarm to !
control room operators and the ground if allowed to continue could cause a i failure of the power supply in the KC324 panel. During the period of time l the TMo is in place, the control room will not receive alarms in the l control room in the event the valves are not in the full open position.
l USA *. Table 9.5A-1 sheet 51 (Appendix A to Branch Technical Position APCSB t 9.5-1) requires that "All valves in the fire water systems should be j electronically supervised. The electrical supervision signal should ,
indicate in the control room. When electrical supervision is not l practicable, an adequate management supervision program should be provided. Such a program should include locking the valves open with strict key control, tamper proof seals, and periodic vicual check of all -
valves." ,
Valve 1FPO43 is located in the security isolation zone, and access to this area is highly restricted. Additionally, the valve's position is verified at least once every 30 days by Procedure STN FP-201, " Water Supply Fire Protection Valve Position Verification." Valve 1FPO49 is located in the yard on the east side of the Turbine Building. The valve will have a tamper proof lock installed as part of this TMO, and the valve's position is verif2ed at least once every 30 days by STN FP-201. Additionally, there are three yard loops to turbine building risers. When one riser is [
isolated (valve closed) and an adequate water supply is still available to the turbine building suppression systems.
The above actions will provide an equivalent level of protection to ensure that the two valves remain in the open position and the intent of USAR Table 9.5A-1 sheet 51 and Appendix A to Branch Technical Position APCSB 9.5-1 is met.
This temporary modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
. This temporary modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the
Attachment II to ET 99-0003 Page 56 of 217 i
safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this temporary modification does not involve any unreviewed safety question.
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Attachment II to ET 99-0003 Page 57 of 217 Safety Evaluation: 59 98-0027 Revision: O service Water and Essential Service Water Cross-tie Valve Replacement change package 07416 provides for the replacement of the cross-tie valves.
between the Service Water System and the Essential Service Water System (EFHV0023, 24, 25, 26, 39, 40, 41, 42) with tight shut-off stainless steel butterfly valves procured to the requirements of Specification M-261(Q) .
The existing Limitorque motor operators will be reused on the new valves.
The valves shall match the existing piping and flange configuration and will be drop in replacements, eliminating the need for piping alterations. The change package will also provide all of the appropriate documentation updates.
Updated Safety Analysis Report (USAR) Table 3.11(B)-3, " Identification of Safety Related Equipment and Components Environmental Qualification,"
Sheets 55 and 56, identify these valves along with their design specification number. This design specification is changed from M-235 to M-261. Essential Service Water System valves EFHV0025, EFHV0026, EFHV0039 and EFHV0040 are scheduled for r*9lacement during Refuel 10. EFHV0023, EFHV0024, EFHV0041 & EFHV0042 are scheduled for replacement during Refuel 11.
The valves are ASME Section III, class 3 motor-operated tight shut-off butterfly valves. These valves are classified as active valves per USAR Table 9.3 (B)-16 for system pressure boundary. The valves are designed seismic category 1. They are normally open and receive an " SIS CLOSE" signal. Per USAR Table 3.11(B)-3, Sheets 55 and 56, all eight of these valves are located in a Category D mild environment (room 3101). Thus, environmental qualification for a harsh environment is not required. In accordance with specification M-261(Q), the replacement valve assembly is being analyzed to confirm that the natural frequency of vibration is greater than 3 Hertz. A structural integrity and weak link analysis is also performed on pressure boundary parts to verify stresses are within ASME Code allowable under applicable loading.
A reliable isolation between the Service Water and Essential Service Water systems is required for operability of the Essential Service Water system I following a Safety Injection, Loss of Off-site Power or Low Auxiliary ,
Feedwater Suction Pressure conditions. There is currently an available ,
margin of 140 gpm for Ultimate Heat Sink (UHS) leakage. The new valves )
are designed to be Class VI shut-off (per ANSI /FCI 70-2) or as close as j possible to zero leakage so that the leakage margin can be maintained i available for containment cooler leakage, underground piping leakage and l other unmonitored leakage that might occur. j i
The new valves have equivalent design requirements as the original valves l they replace thus there are no credible accidents created by the change. !
The new valves have equivalent design requirements as the original valves j i
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Attachment II to ET 99-0003 .
Page 58 of 217 they replace; therefore, there are no credible malfunctions of equipment created by the change. There are no acceptance limits within Technical i specifications that would be affected by the change because this is an equivalent replacement.
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Attachment II to ET 99-0003 ,
Page 59 of 217 Safety Evaluation 8 59 98-0030 Revision: 0 Revision to Containment Leakrate Testing Program Procedure AP 29E-001, " Containment Leakrate Testing Program," is being revised to incorporate program interfaces with the Inservice Testing (IST)
Program and the Surveillance Test Master Cross-reference and Review Requirements. The revised procedure also provides methodology for administrative leakage limits. Sections of the Updated Safety Analysis Report (USAR) are being revised to carrectly reflect the test methodology, the reference to 10CFR50, Appendix J, the test frequency for Type A test and to correct the penetration identification for the containment pressure sensing monitors. These changes to the USAR do not change any commitments or requirements previously stated. USAR Section 6.2.6.3 is being revised to remove the reference to equipment that is not used during the Containment Isolation Valve Leakage Rate Tests. A holding tank and absolute pressure gauge are not used during testing. The reference to recording of the air flow rate and pressure readings at specified data sheet intervals was removed. USAR Section 9.4.6.4 is being revised to correct the reference to 10 CFR 50, Appendix J. Table 18.2-2 is being revised to correct the penetration numbers for the containment pressure !
sensing monitors. USAR Section 16.6.1.1.1 is being revised to reflect the l correct test frequency extension allowance for the Integrated Leakage Rate Test (ILRT) .
The Containment Leak Testing Program as described in procedure AP 29E-001 will require updating. This procedure provides the methodology and values for the administrative controls for component leakage limits. The changes do not impact any other information in the USAR.
The changes to procedure AP 29E-001 and the USAR do not affect any of the design basis accidents discussed or referenced in USAR. No plant st ructures , systems, or components are affected by this change.
These changes correct the test methodology, the reference to 10 CFR 50 Appendix J and the test frequency for the Type A test in the USAR, and do not affect the operation of any SSC as described in the USAR, no credible accidents are identified. Since the changes do not affect the operation or function of any SSC as described in the USAR, no credible malfunction of equipment important to safety are identified.
Since no design basis accidents were identified, the probability of occurrence of an accident is not affected by this change. Since no design basis accidents were identified the probability of occurrence of an accident is not affected by this change. Since no design basis accidents were identified, consequences of accidents are not affected by this change. Since no malfunctions were identified the probability of occurrence of a malfunction is not affected by this change. Since no malfunctions were identified the consequences of a malfunction are not
I Attachment II to ET 99-0003 Page 60 of 217 affected by this change. Since no credible accidents that could be created are identified no accidents of a different type could be created.
Since no malfunctions were identified no malfunctions of a different type could be created. Since no acceptance limits were identified that could ,
be affected , the margin of safety is not affected by this change.
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Attachment II to ET 99-0003 Page 61 of 217 Safety Evaluation: 59 98-0031 Revision: 0
' Fire Detection Replacement Revision 13 to Plant Modification Request (PMR) 04519 is issued to install a new. stand alone fire detection / door closer device in the control room pantry. The single detector is a Simplex fixed temperature (135 degrees Fahrenheit) heat detector which, when alarmed, will activate a new 120V powered LCN model 4040SE series door closer. The detector will also actuate a Simplex strobe being installed in the control room next to the pantry door. The detector will not alarm to the control room KC008' panel.
Updated Safety Analysis Report (USAR) Appendix 9.5B, the fire hazards analysis, describes the fire protection features in the pantry which is located in Fire Area C-28. The current description states that an ionization detector is installed which alarms to panel KC008. There is
. currently no discussion of the door closer device. The design basis accidents'of USAR Chapters 2,3,9 and 15 were reviewed. No design basis accident scenarios are impacted by this USAR change. The change only affects the inputs used for a design basis fire. All inputs and assumptions used for evaluation and analysis of other design basis accidents identified in USAR chapter 15 or other USAR sections other than 9,5.1 are unchanged.
Since this area contains no safe shutdown equipment and is separated from adjacent areas by three-hour-rated barriers, a fire in this area will not prevent safe shutdown A fire in this area will be detected and alarmed by the automatic detection system. The three hour rated fire barriers will limit the spread of fire to adjoining fire areas. The fire can be extinguished manually with the portable extinguishers and/or hose station located in the adjacent area. Continuous occupancy of the adjacent control room ensures a rapid response to a fire.in this area. In the remote event of failure of the door closer, control room personnel can close the door to ensure the fire barrier is maintained until manual fire suppression activities are initiated.
No new failure modes are created and the design basis fire has been previously evaluated. The fire door will limit the spread of fire between the two adjacent fire areas.
There are no safety related circuits,1 ceway or equipment located in room 3602. The fire door will limit the spread of fire between the two adjacent fire areas and ensure a fire in room 3602 will not spread to the control room.
No significant new fire hazard is being introduced. The consequences of a design basis fire in room 3602 or Fire Area C-28 is not affected by the
i Attachment II to ET 99-0003
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Page 62 of 217 change. The door closer will actuate to ensure the fire barrier is maintained.
The impact of a design basis fire in room 3602 is unchanged. There are no radiological consequences from a fire in Fire Area C-28; therefore, the consequences of a fire in this area are unchanged.
No new fire hazard or ignition source is being added. The impact of a design basis fire in room 3602 is unchanged. There are no circuits or equipment required for fire safe shut down located in Fire Area C-28.
I The impact of a design basis fire in room 3602 is unchanged. There are no radiological consequences from a fire in Fire Area C-28 therefore the consequences of a fire in this area is unchanged. There are no circuits or equipment required for fire safe shut down located in Fire Area C-28.
No new fire hazard or ignition source is being added. The impact of a design basis fire in room 3602 is unchanged. No new failure modes or ;
possible equipment failures are created. The probability of a fire in !
room 3602 is not changed, r
No new fire hazard or ignition source is being added. The impact of a [
design basis fire in room 3602 is unchanged. No new failure modes or ;
equipment failures are created. The probability of a fire in room 3602 or fire area C-28 is not changed.
This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specitications, is not reduced by this modification. Therefore, this modification does not !
involve any unreviewed safety question. !
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I Attachment II to ET 99-0003 Page 63 of 217 Safety Evaluatior.: 59 98-0032 Revision 0 Post Accident Sampling System Change This revision to the Updated Safety Analysis Report (USAR) involves the elimination of the requirement to perform a Reactor Coolant System (RCS) dissolved hydrogen or RCS total gas analysis in accordance with NUREG 0737, " Clarification of TMI Action Plan Requirements." The analysis performed for the dissolved hydrogen or total gas samples are not used as part of any Core Damage Assessment Methodology at Wolf Creek Generating Station (WCGS). The samples provide no information which is used to mitigate or reduce the severity of an accident response.
There are no other activities performed in association with the Post Accident Sampling System that would make the information contained within the USAR incorrect. The operation of the system is based on NUREG 0737 and Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."
A review of the accident scenarios in the USAR identified no scenarios in ,
which the proposed revision will have any impact. The Post Accident Sampling System is used in post accident scenarios to help evaluate the extent af the accident.
There are no credible accident scenarios that the proposed revision to the USAR could create. This is based on the documentation only nature of the change since there will be no operational or procedural changes to the methodology employed at WCGS.
Due to the location of the PASS panel and its support equipment, there are no credible malfunctions of equipment importat.t to safety due co the proposed change. There is no affect on the Technical Specifications or their bases as a result of this change to the USAR.
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t Safety Evaluation: 59 98-0033 Revision 0 !
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' Cooldown Re-analysis l
J Re-analysis of the time required to cooldown the plant resulted in minor ;
changes in the times to cooldown based on two train and one train Residual f
Heat Removal- (RHR) operation. 'The cooldown times are still within limits j specified.in the Technical Specifications. The cooldown rates also j maintain a maximum 120*F Component Cooling Water to the shell side of the RHR Heat Exchangers and limit the Reactor Coolant System (RCS) cooldown l rate to a maximum of 50 degrees. Fahrenheit / Hour. i The cooldown re-analysis was performed as a result of a determination that l
the actual normal Service Water flow rate is 8,800 gpm. Also, more ;
realistic auxiliary heat loads were used to more accurately reflect the actual heat loads expected during a normal cooldown. A major change in ;
the auxiliary heat loads acknowledges that the waste and recycle 'l evaporators are unlikely to be operating and do not need to be retained in !
the analysis. !
Though'there are many' incidental changes to the Updated Safety Analysis Report (USAR) resulting from this re-analysis, the changes are primarily l related to addressing the effects of th. mutually. offsetting conditions of l reducing heat loads and reducing the. normal flow of Scr".ce Water. l l
The information provided in the USAR was found to be more i definitive / descriptive than the Technical Specifications, creating the -
l potential for, if not the appearance of, contradictory information being presented for the same scenario. Therefore, the USAR is being revised to make it consistent with Technical Specification requirements. Analysis confirmed Technical Specification requirements can be met during cooldown. ,
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None of the design basis accidents discussed or referenced in the USAR j l
will be affected by this change as the re-analysis encompasses the most
.likely and up-to-date collection of actual operating systems and components. Even if there were an additional heat load that is being .
overlooked, there would be no safety related concerns. This is because l l' . the system may take longer than expected to cooldown, but it will l cooldown. Therefore, the only real concern is to get cooled down within l the time frame (s) specified in the Technical Specifications, not a true i I
concern with whether plant safety would be put at risk.
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l No new credible accidents will be created by making the requested l changes. The only issues really being discussed here are of-timing and j what heat loads will actually be present. We have reviewed the list of !
i potentially. operating components / systems and found that previous analyses l l were too conservative with respect to estimated heat loads when compared l to actual operating heat. loads. The fact that this analysis did not use l 5
the 8800 gpm normal Service Water flow earlier was a concern that is being j l )
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Attachment II to ET 99-0003 )
l Page 65 of 217 '
l addressed by the corrective action program. However, the reductions in i
. heat loads are more than enough to offset the reductirn in normal Service 1 Water flow. l The mutually off-setting affects of reducing expected heat loads and reducing normal Service Water flows result in no significant changes that could directly or indirectly affect the credible malfunctions of any ,
equipment important to safety.
The re-analysis of the mutually off-setting affects of reducing expected heat loads and reducing normal Service Water flows resulted in the Technical Specification requirements being unaffected. I t
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Attachment II to ET 99-0003 !
l Page 66 of 217 i l
Safety Evaluation: 59 98-0034 Revision 0 l
Organisation Changes l
.The change affects the reporting structure and resumes in the Updated Safety Analysis Report (USAR). This is a change to organizational l structure and personnel. These changes are being made to better align the )
organization to meet the need of changing conditions and expectations. '
The changes include the addition of the position of Assistant to the Chief Administrative Officer and the personnel assigned to that position. This '
change also includes reassignment of personnel to the positions of ;
Superintendent Operations, and Manager Integrated Plant Scheduling. The 1 position.of Superintendent Operations Support will be temporarily vacant until an individual is selected. In addition, the Superintendent -l Operations is assuining a position as shift Supervisor.
This is a change in personnel and reporting structure. The change wilt not delete any processes or administrative controls. With the exception .;
of the Shift Supervisor, there are no additional qualifications for these l positions other than the minimum identified in USAR Section 13.1.13 and Table 13.1. The personnel identified for all penitions meet the required qualifications and experience. All of the funct2cns of the position that l will be temporarily vacant will continue. .The personnel reporting to the !
vacant position will have management ovessight.
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Design basis accidente are not affected by this change in personnel because they do not take credit for these positions. All qualifications remain the satae. No administrative or organizational functions have been deleted. !
No administrative or organizational functions have been deleted.
Personnel meet the required qualifications. Therefore, no new accidents will be created.
No administrative or organizational functions have been deleted.
Personnel meet the required qualifications. Therefore, the change will not affect credible malfunctions of equipment important to safety. ;
The criteria for personnel established in the USAR continues to be met.
There is no change in qualifications with these personnel changes.
- Therefore, no change is being made in acceptance limits.
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Safety Evaluation: 59 98-0035 Revision: O solid waste storage Area Grating This Design Change Package (DCP) was created to replace the grating over the trenches in the Solid Waste Storage Area with checkered (diamond) plate steel. There are a total of 115 linear feet of open trenches in the Radwaste Solid Waste Handling Area. The grating currently installed over the trenches at elevation 2000'-0" allows debris to fall into the !
trenches. The debris in the trenches drain to the Radwaste Drumming Area !
Sump, which contains potentially contaminated waste. This not only ,
increases the volume of contaminated waste, but it also introduces new :
types of chemicals into the waste volume. Changing the grating to checkered plate will keep most of the debris out of the trenches.
The grating will be removed from the open trenches in the subject area and 3/8 inch thick checkered plate will be installed over the trenches.
Because of the difference in thickness between the grating and the ,
checkered plate, the new plate must be shimmed.
The new checkered plate will p, event debris from entering the trenches.
The original design of Lh: t enches was to provide a more direct and larger drainage method for '.he washing down of the floors. Washing down ,
of the floors.is no longer performed as it was in the past. Therefore, ;
the grating is no longer required for drainage purposes. In addition, the l checkered plate will provide a smoother surface for rolling equipment over.
Along with C-OS7212, C-1C7221, C-1C7241, and 10466-A-1348, drawing M-1G011 will be revised to reflect the change from grating to checkered plate.
Because this drawing is shown in Figure 1.2-3 of the Updated Safety Analysis Report (USAR), a USAR change is required.
Because no design basis accidents are identified the probability of )
occurrence of an accident is not affected. Because no design basis accidents are identified, the consequences of accidents are not affected.
Because no malfunctions are identified the probability of occurrence of a 1 malfunction is not affected. Because no malfunctions are identified the l consequences of a malfunction are not affected. Because no credible accidents that could be created are identified no accidents of a different type can be created. Ba:L no malfunctions are identified no malfunctions of a diffe;/ c t Pe can be created. Because no acceptance limits are identified ti,- could be affected, the margin of safety is not affected. ,
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Attachment ~ II to ET 99-0003 i Page 68 of 217 Safety Evaluations in 98-0036 Revision: 0 i
Radwaste Building Jib CI me Modification The proposed activitl installs a cantilever-type jib crane on the wall of !
the Radwaste Building to :he south of the wall opening at the west end of .!
the floor at EL. 20318-fd. The jib crane will provide a means to !
raise / lower equipment through the wall opening. The 2 wall brackets for ,
the jib crane will be welded or bolted to baseplates, which will be -
anchored into the end of the 2-foot thick wall (pour OC741W19) separating !
the Seconday Liquid Waste System (SLWS) Charcoal Adsorber (FHF01) from the
'SLWS Demin (FHF02). .
The jib crane and its supporting mechanisms have been qualified to take a rated load of 700 lb for the full range of its swing, in addition to the >
load from its own weight. USAR Figures will be revised to reflect the addition of the jib crane. No other USAR descriptions of conclusions would change or be untrue because of this change.
l Because this change is in the Radwaste Building, which contains no safety related equipment, none of the design basis accidents in USAR chapters 2, i 3, 6, 9 or 15 woulu be affected. 5 The Jib Crane has been evaluated to be adequate for its design loads. l However, even a complete failure of the Jib Crane would only result in the pullout of the anchors from the walls, and a collapse of the Jib Crane onto either the 2022' or 2031'-6" elevation. Because there will be no' ]
safety related items below the Jib Crane or adjacent to its supporting {
walls, no new credible accidents would be created by this change.
l There is no equipment important to safety in the proposed area of the Jib Crane. Therefore, ':here can be no credible malfunctions of equipment important to safety.
There is no equipment important to safety in the proposed area of the Jib Crane. Therefore, no acceptance limits contained in the bases for the technical specifications could be affected.
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Attachment II to ET 99-0003 Page 69 of 217 4
Safety Evaluation: 59 98-0037 Revision: 0 s.
j Plant Cooldown and Shutdown Changes I This Updated Safety Analysis Report (USAR) change is a result of an issue i from the Residual Heat Removal (RHR) and Component Cooling Water Systems l j (CCW), NRC Architect Engineer Inspection. . This USAR change will revise the Plant Cooldown and Shutdown paragraph in USAR Section 9.2.2.2.3 to clarify that the CCW flow to the Spant Fuel Pool (SFP) Heat Exchanger will e
- not always have to be reduced or terminated four hours after the plant is
, shutdown when a 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> cocidown of the plant is desired. Currently, this USAR Section can be interpreted to mean that the CCW flow to the SFP Heat Exchanger will always be reduced or terminated four hours after this type e of shutdown. When reviewing this interpretation with plant procedures it >
has been found that not all of the associated plant cooldown procedures comply with this interpretation of always reducing or terminating the CCW flow to the SFP Heat Exchanger. The current statements in the USAR are based upon the bounding conditions of removing the evaluated largest duty heat loads from the plant four hours after shutdown with the highest expected cooling lake water temperature. If the heat loads are less, colder cooling water is available or the cooldown occurs later in time no :
reduction in CCW flow to the SFP Heat Exchanger may be necessary.
This USAR clarification only affects the USAR document. Clarifying the USAR document will enable a better interpretation of the plant by the USAR. Making the subject clarification to the USAR does not change any plant equipment, setpoints or emergency procedures. This clarification is an explanation of the currently existing USAR statements. The subject clarification to the USAR does not increase the probability of an ,
accident, malfunction or create the possibility of a new type of accident or malfunction because the clarification does not affect the safety design i features of the plant.
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Attachment II to ET 99-0003 Page 70 of 217 Safety Evaluation: 59 98-0038 Revision: 0 Emergency Diesel Generator Load Increase This Design Change Package (DCP) provides analysis of cold weather operation of the Component Cooling Water System (CCW). Due to increased friction of the lubricating oil at 32 degrees Fahrenheit, additional power requirements will be imposed on the Emergency Diesel Generators (EDGs).
The analysis shows an additional 12 kW of load to be imposed for each of the Safety Injection Pumps (DPEM01A,B) and 12 kW of additional load for each of the Centrifugal Charging Pumps (DPBG05A,B). The scope of this Unreviewed Safety Question Determination (USQD) is to evaluate the acceptability of adding 24 kW of additional load per Train to KKJ01A,B.
The EDG loads are tabulated on design drawing E-11005 (USAR Figure 8.3-2).
The design continuous rating for each EDG is 6201 kW. The maximum continuous load for each EDG is 5845 kW. The maximum load of 5845 kW includes safety-related and non-safety related loads and occurs during a Station Black-Out event while in the cold shutdown mode. There remains a margin of 356 kW after the 24 kW is included in the EDG load cLlculation.
Overload of the EDG is not an accident scenario previously evaluated in the USAR, nor will the addition of 24 kW of load to the EDG increase the probability of any accident. Both EDGs have a load margin of 356 kW and will continue to be capable of performing the intended design basis function. The addition of 24 kW load to the EDG will not increase the radiological consequences of any accident. Sufficient load margin exists for both EDG's.
The proposed change will not increase the probability of occurrence of a malfunction of equipment important to safety what has been previously evaluated in the USAR. The addition of 24 kW load to the EDGs will not impact the probability of a me.lfunction of equipment important to safety.
The proposed change will not increase the radiological consequences of a malfunction of equipment important to safety. The EDG's continue to have significant load margin. Consequently the additional load of 24 kW will not impact the ability of the EDG's to perform the required design basis function.
The proposed change will not create the possibility of a new or different type of accident from which has been previously analyzed in the USAR. The change consists of increasing the load on each EDG by 24 kW, which is well within the capability of the EDG's.
The proposed change will not create the possibility of a different type of malfunction of equipment important to safety which has been previously analyzed in the USAR. The addition of 24 kW of load to each EDG 1 eaves a margin of 356 kW. Tb? "nG's will continue to perform the intended design
Attachment II to ET 99-0003 Page 71 of 217 f
basis function.
The proposed change does not reduce the margin of safety as defined in the Technical Specifications. The EDG load margin is not defined in the Technical Specifications. l J
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Attachment' II to ET 99-0003 1 Page 72 of 217 1
1 Safety Evaluation: 59 98-0039 Revision: O Clarifiction To Residual Heat Removal System Leakage j Updated Safety Analysis Report (USAR) Table 5.4A-1 indicates that WCNOC is in compliance with the following statement from Regulatory Guide 1.139:
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"The leakage limits at which an RHR train is to be declared inoperable and ,
isolated should be stated in the Plant Technical Specifications." l The USAR Table 5.4A-1 response to this item is being revised to state that '
Residual Heat Removal (RHR) leakage is addressed in the Reactor Coolant ,
Sources outside Containment program as discussed in the technical l specifications and USAR Section 18.3.4, l
USAR Section 18.3.4.2 is being revised to incorporate Wolf Creek's updated response to NUREG 0737, " Clarification of TMI Action Plan Requirements," i Item III.D.1.1, as documented in letter KMLNRC B4-046 dated March 30, )
1984. It appears that this change shoulu have been incorporated into the i Final Safety Analysis Report (FSAR) in 1984. The affect of these changes I is to eliminate the discrepancy between USAR Table 5.4A-1 and the actual j Wolf Creek Licensing basis.
Procedure AP 25C-001 "WCGS Leak Reduction of Primary Coolant Sources !
Outside of Containment" is being changed, along with its supporting surveillances. There changes include the addition of a 1 gpm acceptance criteria'for combined Emergency Core Cooling System (ECCS) leakage.
i This USAR revision only affects documentation. The changes to AP 25C-001 and its supporting surveillances do not affect equipment operation. No !
accident initiators are affected.
- The actual licensing basis is not being changed. Acceptr ice criteria for combined ECCS leakage is being added to AP 25C-001 to help assure continuous compliance with the accident analysis. The accident radiological consequences analyses are not affected. Equipment operation ,
is not affected.
There are no new types of equipment or operating practices being introduced. The actual licensing basis is not being changed, and equipment operation is not affected. All SSCs and operating procedures ,
remain as before. !
No equipment important to safety is being added, removed or physically i modified. The USAR change affects documentation only. There are no
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Attachment II to ET 99-0003 Page 73 of 217 Safety Evaluation: 59 98-0040 Revision: 0
' Valve Replacement
' Configuration Change Package (CCP) 03855 provides for the equivalent
. replacement of the Secondary Liquid Radwaste System valve (HFV0124). This Hancock gate valve is replace 1 with an equivalent Vogt gate valve. The specific valve in question '.s the drain valve for the secondary liquid waste discharge pump. This CCP provides approval to substitute the equivalent Vogt valve for other Hancock valve of this type, with the requirement that Maintenance notify Engineering when such substitutions are made.
Also included in the scope of this package is a revision to USAR Table 10.4-15 " Secondary Liquid Waste System Component Data", which currently specifies that piping and valve material will be 3G4SS. The change package allows the option of using either 316SS or 304SS valve material.
-The affected valves are all non-safety related. Some of the affected valves may be Special Scope D-Augmented. The only change that is required is to USAR Table 10.4-15 " Secondary Liquid Waste System Component Data".
All accidents were reviewed. There was no impact discovered which would-result from this change. The new valve has equivalent design requirements, which are equal to or'better than, the original valve. Thus there.are no credible accidents created by this change.
Because the new valve has equivalent design requirements, there are no credible malfunctions of equipment affected by this change.
~Because the new valve has equivalent-design requirements, there are no acceptance limits affected by this change.
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Page 74 of 217 i l
l Safety Evaluation: 59 98-0041 Revision: 0 (
r PNR Primary to Secondary Leak Guidelines f
The proposed activity being evaluated adds guidance from EPRI TR-104788, "PWR Primary to Secondary Leak Guidelines", into procedure AP 26A-003, l
" Primary to Secondary S/G Leakage", step B.1.5, " Guidance for operation ,
with SG Tube leak". One of the actions recommended by EPRI TR-104788 is !
to lower radiation monitor setpoints to provide prompt indication of !
increased leakage during'a monitored Steam Generator (S/G) tube leak. [
Updated Safety Analysis Report (USAR) Tables 11.5-1 and 11.5-3 list the [
affected Radiation Monitors (GE RE-92, BM RE-25 and SJ-RE-02) and list !
their respective alert and high alarm setpoints. !
1 This USQD is in support of a revision to procedure AP 21D-001, Primary to Secondary S/G Leakage. The-procedure provides instruction based on EPRI t guidelines for monitoring small S/G leakage. The guidelines recommend ;
adjusting radiation monitor alarm limits just above what the monitored i leakage creates so that an alarm is given if the leak rate increases. ;
USAR Table 11.5-1 and 11.5-3 gave fixed alarm setpoints for the three .
affected rad monitors. USARCR 98-056 was submitted with this USQD to add f notes to Table's 11.5-1 and 11.5-3 to state that the alarm setpoints may j be adjusted during monitored S/G leaks in order to provide prompt I indication of increases in leak rate. In some cases the alarm setpoints :
may be increased above the fixed USAR table setpoint which may be a non- l l conservative change from the fixed value. However, the guidelines in AP f l 21D-001 can only be used for small monitored S/G leaks up to 150 gpd. If I
or when 150 gpd is exceeded, then a plant shutdown will be initiated. The }
150 gpd limit in the procedure is well below the Tech. Spec. 3.4.6.2.c limits for S/G leakage which are 1 gpm (1440 gpd) through all SG's and [
500 gpd through any one SG. }
l The guidance to adjust alarm setpoints in order to provide prompt I indication of increased leakage during a monitored Steam Generator tube leak will not increase the probability of occurrence of an accident i previously evaluated in the USAR. j i
Adjusting the alarm setpoints in order to provide prompt indication of j increased leakage during a monitored Steam Generator tube leak will provide better control for mitigating the radiological consequences of a I monitored Steam Generator tube leak up to 150 gpd. ;
6 i The guidance to adjust alarm setpoints, in order to provide prompt I i
indication of increased leakage during a monitored Steam Generator tube
- leak, does not degrade any equipment important to safety and therefore i does not increase the probability of occurrence of a malfunction to this equipment. l
- Adjusting the alarm setpoints, in order to provide prompt indication of i
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l Attachment II to ET 99-0003 j Page 75 of 217 increased leakage during a monitored Steam Generator tube leak, does not create any initiators that would create the possibility of a different type of accident than was previously evaluated in the USAR. ,
The guidance to adjust alarm setpoints, in order to provide prompt indication of increased leakage during a monitored Steam Generator tube ,
leak, will not cause initiators that would create the possibility of a different type of malfunction of equipment important to safety than was l previously evaluated in the USAR.
The acceptance limits contained in the bases for the technical specifications are not affected by the proposed procedure change.
Therefore, the proposed change will not reduce the margin of safety for any technical specification.
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t Attachment II to ET 99-0003 :
Page 76 of 217 l Safety Evaluation: 59 98-0042 Revision 0 f closed Cooling Water System Description
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Configuration Change Package (CCP) 07637 provides for changes to the f Updated Safety Analysis Report (USAR) and the Closed cooling Water System ;
Description, M-10EB. These are document changes only. The Closed Cooling i Water System (EB) is non safety related-
. None of these documentation corrections affect the design basis function of the system or its role in the hazards and accident analysis in the USAR. The following items are ;
corrected in the USAR by this change, i I
A discrepancy was found in USAR Table 9.2-24. The Degasifier Vacuum Pumps {
were omitted from this list of components cooled by the Closed Cooling i Water System.
- USAR Table 9.2-24 identifies the components cooled by the CCWS. In this .
table the.Degasifier vacuum Pumps for the Demineralized Water Transfer and ;
Storage system were omitted. The total heat load duty is corrected by this change. !
In USAR Section 9.2.8.2.2, the description of the Closed Cooling Water I
Pumps reference 930 gpm as the approximate flow rate of the pumps. This value is actually the design flow of the closed Cooling Water heat ;
exchangers. The pump design operating flow is actually 1062 gpm. This is i corrected by this change. i This revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type,than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.
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i Attachment II to ET 99-0003 l Page 77 of 217 t
Safety Evaluation: 59 98-0043 Revision: 0 Chenecial and volume Control system Insulation The purpose-of this modification is to replace portions of the insulation 5 l on the Chemical Volume Control System (CVCS) normal letdown piping inside ;
, the containment with FOAMGLASe Super K" insulation. This will prevent ;
i thermally induced overpressurization in the pipe during a loss-of-coolant I accident (LOCA) or high energy line break (HELB) condition. The thermally !
induced overpressurization of isolated water-filled piping sections could i potentially jeopardize the ability of accident-mitigating systems to ;
perform their safety functions, and could also lead to breach of !
L containment integrity via bypass leakage. This modification is part of .
the response to NRC Generic Letter 96-06, " Assurance of Equipment operability and Containment Integrity." I During an accident, FOAMGLAS insulation provides thermal protection for !
l the entrapped liquid so that overpressurization will not occur. Unlike l NUKON insulation,.FOAMGLAS material can withstand the crushing forces associated with a LOCA/HELB, and its insulating properties remain -
undegraded on the high humidity containment atmosphere. The FOAMGLAS insulation performs a safety-related function by maintaining the pressure integrity of the safety-related system piping and components, l
i The FOAMGLAS insulation will be encapsulated with stainless steel jacketing to provide a protective covering for the insulation to minimize damage. The jacketing does not perform any safety-related function during i an accident, nor will its failure will prevent satisfactory accomplishment of a safety-related function. The jacketing is classified as non-safety-related and with the banding is classified as non-safety, Seismic II/I special scope. The installation of jacketing will be additionally secured with 3/4" wing seal bands or banding.
i Updated Safety Analysis Report (USAR) Section 9.3.4.2.2 will be revised to include the safety function of insulation during a LOCA/HELB on the affected CVCS normal letdown piping. The FOAMGLAS8 Super K* insulation I
and the wing seal banding are classified as safety-related for maintaining pressure integrity. The installation of the FOAMGLAS' Super K" insulation will have no effect on the operation of the system. Neither will this j change affect the Wolf Creek's Technical Specifications. Also, there will be no changes required to any procedures, activities or administrative controls.
Because no design basis accidents are identified the probability of occurrence of an accident is not affected. Because no design basis
. accidents are identified, the consequences of accidents are not affected.
l' Because no malfunctions are identified the probability of occurrence of a malfunction is not affected. Because no malfunctions are identified the consequences of a malfunction are not affected. Because no credible ,
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Attachment II to ET 99-0003 Page 78 of 217 ;
accidents that could be created are identified no accidents of a different type can be created. Because no malfunctions are identified no malfunctions of a different type can be created. Because no acceptance limits are identified that could be affected, the margin of safety is not ,
affected.
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Attachment II to ET 99-0003 Page 79 of 217 Safety Evaluation: 59 98-0044 Revision: 0 Steam Generator Blowdown Piping Insulation The purpose of this modification is to replace portions of the insulation on the Steam Generator Blowdown (SGB) piping inside the containment with FOAMGLAS* Super K" insulation. This will prevent thermally induced overpressurization in the pipe during a loss-of-coolant accident (LOCA) or high energy line break (HELB) condition. The thermally induced overpressurization of isolated water-filled piping sections could potentially jeopardire the ability of accident-mitigating systems to perform their safety functions and could also lead to breach of containment integrity via bypass leakage. This modification is part of the response to NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity.
During an accident, FOAMGLAS insulation provides thermal protection for the entrapped liquid so that overpressurization will not occur. Unlike NUKON insulation, FOAMGLAS material can withstand against the crushing forces associated with LOCA/HELB, and its insulating properties remain undegraded in the high humidity containment atmosphere. The FOAMGLAS insulation performs a safety-related function by maintaining the pressure integrity of the safety-related piping and components.
The FOAMGLAS insulation will be encapsulated with stainless steel jacketing to provide a protective covering for the insulation against physical abuse that would reduce the effectiveness of the insulation. The jacketing does not perform any safety-related function during an accident, nor does its failure prevent satisfactory accomplishment of a safety-related function. The jacketing is classified as non safety-related, and with the banding, it is classified as non safety, Seismic II/I special scope. The installation of the jacketing will be additionally secured with 3/4" wing seal bands or banding.
The wing seal bands do not provide a direct safety-related function, but whose failure could cause the jacketing to dislodge and leave the insulation unprotected from flying objects during a LOCA/HELB. Banding l protects the insulation so it can perform its safety-related function-therefore, the banding is claselfied as safety-related. I Updated Safety Analysis Report (USAR) Section 10.4.8.2.2 will be revised to include the safety function of the insulation during a LOCA/HELB on the affected SGB piping. The FOAMGLAS* Super K" insulation and the wing seal I banding are classified as safety-related for maintaining pressure integrity. The installation of the FOAMGLAS* Super K" insulation will have no effect on the operation of the system. Neither will this change affect l the Wolf Creek's Technical Specification. There will be no changes q required to any procedures, activities or administrative controls. ;
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Attachment II to ET 99-0003 Page 80 of 217 Because no design basis accidents are identified the probability of occurrence of an accident is not affected. Because no design basis accidents are identified, the consequences of accidents are not affected.
Because no malfunctions are identified the probability of occurrence of a malfunction is net affected. Because no malfunctions are identified the consequences of a' malfunction are not affected. Because no credible accidents that could be creatnd are identified no accidents of a different type can be created. Because no malfunctions are identified no malfunctions of a different type cr* be created. Because no acceptance limits are identified that could be affected, the margin of safety is not affected.
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l Attachment II to ET 99-0003 Page 81 of 217 Safety Evaluation: 59 98-0045 Revision 0 High Pressure Injection System Relief Valve Installation '
This modification installs a new relief valve EMV0251 in the High Pressure '
Coolant. Injection System (EM) test line EM-068-BCD-3/4" to limit the '
pressure during a Loss-of-Coolant Accident (LOCA) or High Energy Line Break (HELB) . . This system is part of the Emergency Core Cooling System ,
(ECCS). During an accident, as the containment temperature rises, the [
trapped volume of liquid can heat up increasing the pressure in the line i to near piping stress allowables and exceeding the attached safety-related <
valve bonnet bolt stress allowables. This normally closed section of FM test piping is located inside the containment between containment isolation valve (EMHV8871) and the normally closed test-source valves. It is a non safety-related section of piping connected to safety-related ;
piping on all ends through safety-related valves. This modification is part of the response to NRC Generic Letter 96-06, " Assurance of Equipment Operability and Containment Integrity." ;
The new relief valve EMV0251 is a safety-related component installed on a non safety-related section of the EM test piping. The effect of this change is to limit the pressure to ensure the stress allowables for all piping and components, especially connecting safety-related ones, are met.
The location for installing the relief valve assembly was selected based I on the required space for installation, ease of handling and maintenance, and absence of equipment that can be affected by its discharge. The valve's discharge piping is designed such that it vents away from nearby safety-related equipment. Flange connections shall be provided to facilitate maintenance and testing.
There will be no additions or changes to any procedures, activities or administrative controls. The installation of the relief valve assembly >
will have no effect on the current operation of the EM system, or the i specific function of the test line. The test line does not support the normal operation of the plant. There are no plant systems, structures, or components or any requirements listed in the Updated Safety Analysis Report (USAR) affected by this activity. The description of the test line ,
in the USAR is unchanged by the proposed activity. Figure 6.3-1-02 of the USAR is to be updated to include this change. There are no tests or [
experiments added by this activity that are not already described in the USAR. This proposed activity is designed to help mitigate the i consequences of an accident.
This modification will have no impact on the design basis accidents discussed in the USAR. The relief valve will help ensure that the integrity of the associated piping and valves is maintaincd during all LOCA/HELBs. ,
All possible failure modes have been investigated. The relief valve will not reduce the redundancy of equipment important to safety. It has been
Attachment II to ET 99-0003 Page 82 of 217 added to help protect equipment important to safety. Therefore, the proposed change will not create the possibility of a different type of malfunction of equipment important to safety previously evaluated in the USAR. The proposed change does not affect any parameters upon which the Technical Specifications are based, therefore, there is no reduction in the margin of safety.
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l Attachment II to ET 99-0C03
- Page 83 of 217 l I
i Safety Evaluation: 59 98-0046 Revision: 0 i i
Title Changes of Vice Presidents l The titles for the Vice President Engineering.and the Chief Operating l Officer are being changed to better reflect their current j responsibilities. The titles are being changed to Vice President [
Engineering and Information Services, and Vice President and Chief l Operating Officer. This is a title change only. No responsibilities are i being changed or deleted. This change will impact the USAR in that both ;
of these titles appear in the USAR and will require updating. f The responsibilities of these two positions are discussed in several {
places in the USAR. This change will not impact any of these discussions ;
and will only correct the title.
l No design bases accidents rely on the title of an individual. Therefore, I there are no design bases accidents are affected. i A title change of this nature is similar to an editorial change and has no l impact on plant operation or equipment. Therefore, no new accidents could ,
be created. i i
A title change of this nature is similar to an editorial change and has no impact on plant operation or equipment. Therefore, there are no credible I malfunctions of equipment important to safety which may be directly or l indirectly affected by the proposed change. ;
i Because this is a title change and not a responsibility change, there are no acceptance limits associated to this change. Prior NRC approval for the position of Chief Operating Officer was based on responsibility not l title.
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Attachment II to ET 99-0003 i
Page 84 of 217 i 1
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Safety Evaluation: 59 98-0047 Revision: 0 [
r Correction of Document Discrepancies Associated to Configuration, Level, ,
Temperature and Pressure of the Safety Injection Accumulator Tanks l l The proposed change involves correcting a number of document discrepancies !
4 in regard to configuration, level, temperature and pressure of the Safety Injection Accumulator Tanks as o;;iginally identified by Performance
. Improvement Request (PIR) 97-0133. The original design, as well as I
subsequent previously approved design modifications, is not consistently
- represented on design and licensing basis documents.
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USAR Table 6.3-1 will be revised to add a note to distinguish between the
- accumulator normal operating conditions of 45 to 120 degrees Fahrenheit i (F) and the design operating conditions of 60 to 150 degrees F. The note
- explains that an ASME Section XI evaluation allows operation as low as 45 !
degrees F, and that the containment integrity accident analyses limits the !
maximum operating temperature. The actual minimum and maximum accumulator operating temperatures will not be affected. >
In addition, USAR Table 6.3-1 will be revised to distinguish between the ,
585 psig " minimum pressure" and the 600 psig " minimum operating !
pressure." This configuration change is a clarification that is made for censistency and will make this table more consistent with other design d
occuments and with other tables contained in the USAR.
The proposed change will not impact or change the accumulators in any ;
way. This USAR revision will not increase the probability of occurrence {
or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. 'This
} revision does not create a possibility for an accident or malfunction of a ;
j different type than any evaluated previously in the safety analysis ,
report. The margin of safety, as defined in technical specifications, is ;
not reduced by this revision. Therefore, this revision does not involve an l unreviewed safety question. l i
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. L Attachment II to ET 99-0003 Page 85 of 217 I
L Safety Evaluation: 59 98-0048 Revision: 0 '
Addition of New Cable for the Air Conditioning Units (SGK05A/B) [
i The addition of new cable for the air conditioning units (SGK05A/B) will ;
correct cable separation requirements identified in Work Package Tasks j 114212-3, 114213-3, 114214-3 and 114215-3. This new cable will increase !
by an almost insignificant. amount the combustible loading in several rooms !
within the power block (i.e. room #1403, 3403, 3408, 3418, 3419, 3503, [
3504 and 3505). j There are no accident scenarios within the USAR which need to be reviewed l for impact by this change. This rationale is based on the identical !
functionality of the equipment pre- and post-modification, combined with !
the fact that the system will now meet the separation requirements r previously used as part of any accident analysis. There are no credible accident scenarios which this modification could create. This modification will correct identified deficiencies in the routing of non-safety related and safety related cables. This modification will not affect.any failure mechanism or mode for the safety related components of the system. There is no impact on any acceptance limit for the technical ,
specifications bases. ,
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Attachment -II to ET 99-0003 I
Page 86 of 217 ;
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l Safety Evaluationt 59 98-0049 Revision 0 '
Change to Volumetric Information in the Updated Safety Analysis Report ,
The volumetric information for the Reactor Coolant Pump Lube Oil Drain i Tank (TLF-01),. identified in the Updated Safety Analysis Report ('USAR)
Section 9.5.1.2.2.5, 9.5B.7-RB4, USAR table 9.5A-1 sheet 56, and USAR l Table 9.5E-1 sheet 33 are not consistent with each other, nor the reference documents. Although the current USAR is correct in calling this tank (TLF-01) a "300 gallon tank," the actual volume is slightly larger than 300 gallons. Therefore, it would be more accurate to identify it as a " greater than 300 gallon tank." ;
In USAR Section 9.5.1.2.2.5 Paragraph 8, the last sentence is deleted. !
This change will provide clarification.
t In USAR Table 9.5E-1, Sheet 35 and 36 contain the same information.
.Therefore, Sheet 36 is deleted from the table, eliminating duplicate I
information.
l The tank information in the USAR is being corrected for accuracy and f
. consistency purposes only. Pages in the USAR to be changed are 9.5-25, l l 9.5B-173, Table 9.5A-1 sheet 56, and Table 9.5E-1 sheet 34. A review of all the accident scenarios in the USAR chapters indicates there are no references.in regarding to tank TLF-01. j Fixed combustibles for Fire zone RB-1 are not exceeded by this change.
Therefore, the reactor coolant pumps are not affected by this change.
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Changes are made for clarity purposes only. This USAR revision will not increase the probability of occurrence or the consequences of an accident '
or malfunction of equipment important to safety previously' evaluated in l the safety analysis report. This revision does not create a possibility i for an accident or malfunction cf a different type than any evaluated ;
previously in the safety analysis report. The margin of safety, as i defined in technics.1 specifications, is not reduced by this revision. ;
Therefore, this revision does not involve any unreviewed safety question. !
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Attachment II to ET 99-0003 Page 87 of 217 t
Safety Evaluation: 59 98-0051 Revision 0 Double Encapsulated Secondary Source Configuration Change Package 07735 addresses a new design for the secondary source assembly, the double encapsulated secondary source (current design is the single encapsulated secondary source). The double encapsulated design adds an additional stainless steel barrier cladding to the single encapsulated design to preclude a potential failure mechanism, liquid metal embrittlement. Since the double encapsulated design provides ;
less volume for source material over the active height, the number of source rods in the assembly is increased from four to s4x to provide an equivalent neutron source to the detectors. Also, the ac'rce rod loading pattern in the assembly is changed. The double encapsulateo -evv.'dary source contains the same source material as the single encapsulat ed secondary source. The cladding materials for the two designs are identical.
The double encapsulated secondary soure", assembly has the same flow blocking ability as the current design. Fuel handling operations with the double encapsulated secondary source assembly will be identical to that already performed utilizing the same tool used for the current design.
The addition of two source rods will not affect fuel movement or any of the various fuel handling tools or cranes. The design life of the double encapsulated secondary source assembly is equivalent to the current design, as based on the design life of the hold down assembly.
Potential failure mechanisms for the two designs are the same, with the exception that the double encapsulated design precludes a liquid metal embrittlement failure mechanism. The double encapsulated secondary source is about 3 pounds heavier than the current design, but lighter than a control rod assembly; thus there is no concern with this small weight increase. The pellet stack height for the two designs is identical (88"),
but the source material in the double encapsulated design is offset about 0.7" higher than that of the current design. This small difference in source material position will have a negligible effect on the response of the source range detectors.
The purpose of the secondary source assembly is to provide base neutron level to ensure that the neutron detectors are operational and responding to core multiplication neutrons. The source assembly permits detection of changes in the core multiplication factor during core loading and approach to criticality.
Based on the above discussion, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated
h Attachment II to ET 99-0003 Page 88 of 217 previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification.
Therefore, this modification does not involve an unreviewed safety question.
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I Attachment II to ET 99-0003 (
Page 89 of 217 i
Safety Evaluation: 59 98-0052 Revision: 0 ;
Control' Room Carpet Replacement f; Updated Safety Analysis Report (USAR) Section 9.5B and Specification A-196 are being revised to remove specific vendor information. The USAR will be j revised such that selection of ' color, styles and manufacturer is lef t tx) !
the discretion of the operations department There is no change in the {
carpet technical design requirements. !
There are no procedures, activities, administrative controls, or sequences l of plant operations; or any p3. ant SSCs which are affected that would make j the USAR untrue or incorrect. There are no tests or experiments not (
described in the USAR which will adversely' affect the adequacy of the SSCs '
to prevent an accident or mitigate the consequences of an accident. ,
i This change does not affect the amount of combustible loading or fire j hazards analysis. This revision will not increase the probability of !
occurrence or the consequences of an accident or malfunction of equipment ,
important to safety previously_ evaluated in the safety analysis report. I This revision does not create a possibility for an accident or malfunction [
of a different' type than any evaluated previously in the safety analysis ,
report. The margin of safety, as defined in technical specifications, is j not reduced by this revision. Therefore, this revision does not involve an !
unreviewed safety question.
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l Attachment II to ET 99-0003 1 1
Page 90 of 217 Safety Evaluation: 59 98-0053 Revision: 0 Blow Down Heat Exchanger i configuration Change Package 07766 provides a design for staking tubes in steam generator blowdown heat exchangers, EBM01 or EBM02. In addition, j this change package allows plugging greater than 5 tubes in steam generator blowdown non-regenerative heat exchanger EBM02. !
Staking is the insertion of a rod that lifts and holds in place the two ends of a broken, or potentially breaking tube. When a tube suffers a guillotine break or is thought to be subject to forces that could result ,
in a guillotine break, the industry accepted practice is to stake the ;
tube. This restrains the tube ends, reducing the potential for damage to adjacent tubes or further distribution of tubing pieces into the system.
A staked tube is equivalent to a plugged tube, i.e. it is out of service.
l Heat exchangers EBM01 and EBM02 are Special Scope 'D' Augmented. Tube j stakes are non-safety related and not special scope because they are not l pressure retaining parts.
]
The High Energy Line break analysis described in the Updated Cafety )
Analysis Report USAR) is unchanged by this activity. This analysis is not performed for portions of the system in the turbine building which is l where the non-regenerative heat exchange is located. '
The Steam Generator Tube Rupture Analysis described in the USAR is ,
unchanged by this activity. This analysis relies on the radiation l monitors for the Steam Generator Blowdown System liquid flow path and/or )
the condenser air discharge to alarm and terminate steam generator blowdown. No radiation monitors have been altered or modified by this !
I activity.
1 Operation with the non-regenerative heat exchange out of service is I described in the USAR and covered by plant operating procedures. In this )
mode, total steam generator blowdown is limited to about 100 gpm under worst case conditions. Since operation with the non-regenerative heat exchange completely bypassed is already described and allowed, the effect of plugging more that five tubes in the non-regenerative heat exchange is minimal. The maximum design flow rate may not be attained before reaching a 120 degree Fahrenheit non-regenerative heat exchanger outlet temperature; but, blowdown capability will be more than what could be attained with the non-regenerative heat exchange completely out of service.
Based on review of the USAR accidents, this modification will have no
. affect on High Energy Line Breaks for the Steam Generator Blow Down System nor a Steam Generator Tube Rupture. Credible malfunctions of equipment ,
l important to safety are unaffected by tube plugging. Tube plugging is being performed in accordance with the existing manufactures instructions, i
l Attachment II to ET 99-0003 Page 91 of .217 r
which requires a welded connection. The *D' Augmented pressure boundary is being maintained in accordance with existing approved design. There are no acceptance limits affected by the proposed activity ,
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! Page 92 of 217 I
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Safety Evaluation: 59 98-0054 Revision: 0 Condensate.Demineralizer Changes t
The following changes will be made to the Condensate Demineralizer System !
1 Description, M-10AK and the Updated Safety Analysis Report (USAR). These are document changes only. The following USAR sections are being changed: ,
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- USAR Section 10.4.6.2.2, paragraph 2 is changed to read: CONDENSATE .
- DEMINERALIZER VESSELS-The six 20-percent-capacity spherical vessels with f 5
deep-bed regenerable mixed strong acid cation / strong base anion resins, !
are constructed of carbon steel and lined with natural rubber. The design l
, flowrate is approximately 53 gpm per square foot of bed, and the bed depth I is 36 inches. :
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- USAR Section 10.4.6.2.3, paragraph 2, first sentence is changed to read
l The ammonia cycle operation with negligible condenser leakage will allow ;
an extended demineralizer run. !
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i USAR Section 10.4.6.2.3, paragraph 3 first sentence is changed to read: 6
. Condensate flow is passed through up to five of the six demineralizer (
3 vessels, which are pipe in parallel. Delete second sentence.
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t USAR Section. 10.4.6.2.3, paragraph 4c changed to read: Initiate resin :
a transfer from the resin mixing and storage tank to the empty demineralizer (
< vessel. !
J t USAR Section 10.4.6.2.3, paragraph 5 first sentence is changed to read:
l j On termination of a service run, the exhausted demineralizer vessel is i j taken out of service, and a standby unit is put in service by remote l manual operation from the local control pane. Last sentence to read: l Each resin is then backwashed, chemically regenerated, rinsed, and i transferred to the resin and storage tank for final rinsing and mixing. !
USAR Section 10.4.6.2.3, paragraph 6 fourth sentence is changee to read: ,
I During the wash-air scrub process, there is no chemical regeneration involved.
USAR Section 10.4.6.2.3, paragraph 7 first sentence is changee to read: A final rinse is performed on the demineralizer before it is placed in service, j USAR Section 10.4.6.2.3, paragraph 9 first sentence is changed to read: l Regenerant wastes are segregated by total dissolved solid content (TDS) l and directed to the low or high TDS tanks in the secondary liquid waste l system. Change fourth sentence to read: The high TDS is generated from !
the chemical regeneration and the initial stages of the rinsing after
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chemical regeneration. Delete the fifth sentence.
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Attachment II to ET 99-0003
- Page 93 of 217 USAR Section 10.4.6.2.3, paragraph 10 delete a and 10.4.11".
USAR Section Table 10.4-4 for the Demineralizer Vessels delete "(including
. one on standby) " and change the Design flow per vessel to read 4,560.
- No design basis accident are identified or evaluated for the non safety-related Condensate Demineralizer System in the USAR. Since there are no physical changes and the design basis function is not changed, no credible accident that could be created are identified. Since the proposed changes would not affect the system's failure modes, controls on activity j performance, the level of qualification, or the effects on equipment important to safety, no credible malfunctions of equipment important to safety are identified. Since the Condensate Demineralizer System is not included in the bases of Technical specifications, no acceptance limits are identified that could be affected. Also, licensing basis documents do !
not contain any acceptance limits that could be affected. l t
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3 Attachment II to ET 99-0003 Page 94 of 217 r
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Safety Evaluation: 59 98-0055 Revision: 0 '
Correction of Updated Safety Analysis Report Discrepancies f L
This change package is corrective action for Performance Improvement i Request (PIR) 98-1420 which addresses discrepancies associated with l Updated Safety Analysis Report (USAR) Section 10.4.1. This change package l will make the following changes to the Condensate System Description, M- i 10AD: A note is added to 3.3.1 of M-10AD stating the Callaway data is !
"Information Only," !
Section 3.2.1 of M 10AD is to be changed by deleting "and alarmed in the l control room" from the following sentence: " Circulating water leakage occurring within the condenser is detected and alarmed in the control room l by monitoring the condensate leaving each hotwell (six monitoring points l altogether) , " and Section 3.2.4.3 of M-10AD, Loss of one string of LP l Heaters, is to be changed to state: " ... Isolation of one string l necessitates reduction in load to 91%...." in lieu of the 85% of turbine !
guaranteed flow. [
In adiition to the above System Description changes the USAR will be changed as follows: }
Section 10.4.1.1.2 states: " ... condensate oxygen content will not exceed 7 f ppb under any normal cperating condition." The phrase "any normal l operatiri condition" is defined in the USAR as all modes cf operation from '
power e eration down to refueling. Therefore, Section 10.4.1.1.2 will be i changed to state: "... condensate oxygen content should not exceed 7 ppb j under normal full power operating conditions." ;
Section 10.4.1.2.3 will be revised to reflect the changes to Section 3.2.1. l Section 10.4.1.2.3, in the description of turbine building flooding due to j failure of the main condenser, the USAR incorrectly states that there is ;
no safety-related equipment in the turbine building. This section will be l changed to state: "The failure of the main condenser and the resulting [
flooding will not preclude operation of any essential system because the !
limited safety related components, instruments and cabling associated with j the main steam dumps and turbine trip / reactor trip signals, are located l well above the expected flood level of the turbine building .." l USAR Section 10.4 1 does not include a description of the contaminants ,
allowed in the condensate or the length of time the condenser may operate l with degraded conditions without affecting the condensate /feedwater !
quality for safe operation. However, this information is located is section 10.4.6. Therefore a reference to Section 10.4.6 will be added to l Section 10.4.1. ;
The above changea involve clarifications to the USAR Section 10.4. As ;
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l Attachment II to ET 99-0003 Page 95 of 217 such this change will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.
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Attachment II to ET 99-0003 Page 96 of 217 '
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1 Safety Evaluation: 59 98-0056 Revision 0 Temporary Special Order for Steam Generator Atmospehric Relief Valves Special order 17 is being put iato place as a temporary measure until the ,
l evaluation discussed below is completed. 1 1
The limiting Steam Generator Tube Rupture (SGTR) analysis assumes that at i i least 2 Atmospheric Relief Valves (ARVs) on intact steam generators will l be available to perform the required post transient cooldown. The l limiting USAR analysis assumes a single failure in the AFW controller such that excessive Auxiliary Feedwater (AFW) is fed to the faulted steam generator. This analysis results in overfill of the ruptured steam generator, water flow through a safety valve. and a consequential failure ,
i of the safety valve to re-close (assumed to stick open 5 percent). For ;
l this case, 2 ARVs on intact loops would be available for cooldown.
l
- However, with one ARV out-of-service as currently allowed by Technical l Specification 3.7.1.6, and assuming the single failure of 1 ARV on an
( intact loop (in place of the failure in the AFW controller), only 1 ARV is l left on an intact loop to perform.the post transient cooldown (i.e., the ,
j 4th ARV is on the ruptured SG). Having only one ARV used for the cooldown j i significantly lengthens the time to reach Residual Heat Removal (RHR) cut ,
l in conditions, and, if steam generator overfill occurs, to terminate the l transient. The result of the delay before reaching RHR cut-in conditions, and thereby terminating the tran=ient, is an increase in radiological ;
dose.
l This scenario is being evaluated to determine if steam generator overfill occurs. Until the evaluation is complete, Special order 17 will ensure that the USAR analysis of radiological doses remains bounding no matter j
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what the outcome of the evaluation.
l i l Special order 17 will implement Technical Specification Lco 3.7.1.6 such l that four steam generator atmospheric relief valves (ARVs) are required to !
be operable instead of the three currently required in the Technical ,
i Specification. The Action statements are applicable to the four ARV [
l requirement; however, action statement b will be modified to require !
restoration of three valves instead of the current two to operable status.
Section 15.6.3 of the USAR describes the SGTR analysis and includes a discussion of the cooldown of the RCS which identifies the current ;
L Technical Specification requirement of having three SG ARVs operable.
l Implementation of Special order 17 will make this section no longer true. i i
The only design basis accident impacted by this change is the SGTR l accident described in Section 15.6.3., The SGTR analysis is the only
- accident analysis'that relies on the ARVs to mitigate the postulated i accident. In addition, this change is conservative with respect to fire i t
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Attachment II to ET 99-0003 Page 97 of 217 l safe shutdown analysis assumptions for ARV operability.
The change is being implemented to ensure that the steam generator tube rupture analysis contained in the USAR is clearly the limiting analysis while the evaluation of the postulated scenario is performed. In addition, this change is not a physical change to the plant. Therefore, there is no possibility of a different type of accident being created with this change. This change is not a physical change to the plant and therefore will not directly or indirectly affect malfunctions of equipment important to safety.
The acceptance limits affected by the proposed activity are the
, radiological dose results of the SGTR analysis contained in the USAR.
This change will ent:ure that the dose limits are met while the evaluation discussed above is being pe-formed.
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Attachment II to ET 99-0003 Page 98 of 217 Safety Evaluatio'n: 59 98-0057 Revision: O Security Plan Revision 30 This revision replaces shotgun training with rifle training in the Security Training and qualification plan. There are no design basis accidents discussed that are affected by these changes. There is no credible accident these changes would create. Changes do not affect operation of plant equipment; therefore, no credible malfunctions could occur.
The probability of the occurrence of an accident previously evaluated has not been increased. This change does not affect equipment important to safety and does not reduce the protection for the plant resources.
The radiological consequences of an accident previously evaluated has not been increased. This change does not affect the design bases or equipment important to safety and does not reduce the protection for the plant resources.
The probability of the occurrence of a malfunction of equipment important to safety previously evaluated has not been increased. This change does not affect the design bases or equipment important to safety and does not reduce the protection for plant resources.
The radiological consequences of a malfunction of equipment important to safety previously evaluated has not been increased. This c'ange does not affect the design bases or equipment important to safety and does not reduce the protection for plant resources.
The possibility of an accident of a different type than any pre.viously evaluated has not been increased. This change does not affect the design bases or equipment important to safety and does no reduce the protection for plant resources.
The possibility of a different type of malfunction of equipment important to safety than any previously evaluated has not been increased. This change does not affect the design bases or equipment important safety and does not reduce the protection for plant resources.
The margin of safety na defined in the basis for any technical specifications has hot been reduced. This change does not affect the design bases or equipment important to safety and does not reduce the protection for plant resources.
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j Attachment II to ET 99-0003 Page 99 of 217 f i
3 Safety Evaluations, 59 98-0058 Revision: 0 !
Venting of the Emergency Core Cooling System l j- This modification will affect the downstream piping of vent valves EJV0094, EJV0128, EMV0156, and EMV0242 in the Residual Heat. Removal System (EJ) and the High Pressure Safety Injection System (EM). The existing 3/4 l inch vent valves.will have their down stream piping modified and extended [
to facilitate venting of the Emergency Core Cooling (ECCS) system. l A 3/4 inch pipe nipple, flange, and blind down~ stream of the outlet of !
each vent valve presently exists in the field. This downstream configuration will be' removed from each vent valve. A pipe to tube ;
adapter will be installed at the outlet side of each vent valve and then pipe tubing will be routed to an accessible and low dose area. This pipe tubing will have double isolation tubing valves and a threaded end cap.
f EJ vent valves will have pipe tubing routed to the floor area in the same room (1309) . EM vent valves will have pipe tubing run from room 1122 i through an existing spare ceiling penetration OP134SO906 and into room l 1323.
f This modification will not increase the probability of occurrence or the l' consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This f modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the' safety analysis ;
report. The margin of safety, as defined in technical specificatlons, is !
not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.
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Attachment II to ET 99-0003 Page 100 of 217 Safety Evaluation: 59 98-0059 Revision 0 Installation of Pipe Threads and a cap on the Plain End of the Pipe Stub Downstream of EPHCV943.
This. temporary modification will install pipe threads and a cap on the plain end of the pipe stub downstream of Accumulator Safety Injection System (EP) Nitrogen Supply Containment Atmosphere Isolation Valve (EPHCV943). The safety related nitrogen supply isolation valves to accumulators A, B, and C are leaking by the valve seat. The non-safety related nitrogen supply containment atmospheric isolation valve, EPHCV943 on the common header upstream of the safety related nitrogen supply isolation valves is also leaking by the valve seat. The cap will be installed on the non-safety related portion of the nitrogen charging piping downstream of and away from the safety related and tested containment isolation valves for the nitrogen header penetration to containment. Containment boundary is provided by tested safety related piping and containment isolation valves are not affected.
The expected results of this temporary modification are that capping off the pipe stub will prevent leakage out of the nitrogen supply piping and thus from the accumulators. These temporary modification actions will only be implemented if the rework on EPHCV943 fails to acceptably limit the leakage through it. The current leakage rate from the accumulators is within acceptable limits such that the accumulators are still capable of fulfilling their design function.
Capping off the pipe stub downstream of valve EPHCV943 will remove the capability to remotely adjust accumulator pressure and also to exhaust the nitrogen fill piping after filling an accumulator currently depicted on Updated Safety Analysis Report (USAR) Figure 6.3-1. This activity can still be done but not remotely from the control room.
The actions of capping off the non-safety related pipe vent downstream of non-safety related valve EPHCV943 has no impact on the emergency procedures. The "ECCS Failure Modes and Effects Analysis," presented in USAR Table 6.3-5, analyzes the active component failures in the ECCS.
This analysis is not affected by this temporary modification.
Besides the changes to the procedures used to fill and adjust the I accumulator nitrogen pressure, there are no other operating procedures that require changing as a result of the modification.
Capping the vent line downstream of EPHCV943 will enable it to pressurize to a maximum pressure of 700 psig. Pressurizing this line to 700 psig will not compromise the structural integrity of this line because it is capable of being exposed to this pressure. The cap that is being installed on the vent line is for high pressure applications and has a rating of at least 1500 psig. This cap weighs about 3 pounds and does not
Attachment II to ET 99-0003 Page 101 of 217 adversely affect, due to its insignificant mass, the seismic response of the associated piping.
The limiting time before the accumulators inject into the RCS is during a small break LOCA. For the 2 inch small break case, the accumulators never inject and the core is recovered in about 5,700 seconds. For the 3 inch case, the accumulators inject after 3,331 seconds (or 55 minutes) and the core is recovered in about 2,900 seconds. For larger break sizes, the accumulators inject in a progressively shorter time from the initiation of the accident.
If accumulator pressure were at the alarm setpoint of 601 psig at the onset of a 3 inch small break LOCA (the most time limiting case before accumulator injection), pressure would have to decrease 16 psi in 55 minutes to be at 585 psig when accumulator injection started. Pressure must decrease at greater than 17.5 psi / hour for the most limiting small break LOCA case to put the initial accumulator injection pressure below the functional limit. The current leakage rate from any one accumulator is significantly less than 17.5 psi / hour. The modification when installed will not change the current leakage pathway that is through the safety related valves or the leakage orifice size of these valves thus it will ,
not aggravate or adversely affect the leakage rate from any one or all of the accumulators. ,
Based on the information above, this temporary modification does not increase the probability of an accident, malfunction or create the possibility of a new type of accident or malfunction because the modification does not affect the safety design features of the plant.
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F Attachment II to ET 99-0003
- Page 102 of 217 Safety Evaluation: 59 96-0060 Revision 0 Revision of Bullet-Resisting Door Requirements The proposed activity is the revision of Specification A-075 " Technical Specification for Bullet-Resisting Door," and a revision to Section 6.4.2.4 of the Updated Safety Analysis Report (USAR).
I After installation of the control Room Bullet Resisting Door (door 36043),
the required airtightness testing was revised to allow the Control Room j Positive Pressure Test in lieu of the testing for the design requirement
, of having a maximum air leakage rate of 0.1 cfm per linear foot of
- perimeter at one fourth inch water gauge differential pressure.
A statement in the USAR Section 6.4.2.4 indicates that all of the doors
- that form a part of the control room pressure boundary open outward. That statement is incorrect. The section needs to be revised to remove the parenthetical statement on the specified design requirement for airtightness (0.1 cfm leakage per linear foot) .
l There are no expected effects of this change package on any component, system or piece of plant equipment. There are no procedures, activities or any other actions (e.g. administrative controls, system operation) j which will be impacted by this modification. This modification will only j serve to Correct Specification A-075, and to correct and revise a Section j of the USAR. This change is a document change only and can have no effect j on any SSC. There are no accident scenarios within the USAR which need to i be reviewed for impact by this change. The rationale behind this is based l on the fact that this change is a document change only.
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$ There are no credible accident scenarios that this modification could j create. This modification is a document change only. This modification j will revise Specification A-075, and will revise a section of the USAR to
(- make them beth corre;t, and to remove a referenced design requirement from the USIR. This modification will not affect in any manner any failure mechanism or mode for any safety related components of any plant system.
This is a document change only, and will not affect any SSC. There is no impact on any acceptance limit associated with the technical specifications bases. This is a document change only.
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Page 103 of 217 l l I t
Safety Evaluation 59 98-0061 Revision: 0 Updated Safety Analysis Report Revision to Feedwater System Valves I l Updated Safety Analysis Report (USAR) Figure 10.4-6-01 (M-12AE01) Note 9 f i requires continuous presence of authorized individuals whenever Feedwater 1 System (ME) valves AEV0361 or AEV0362 are closed to allow maintenance or
( repair of valves AEV0978 or AEV0979 respectively. Change Package 07782
! removes this restriction. Note 9 will be revised to' read as follows:
k
" Valves AEV0361'and AEV0362 shall remain LOCKED OPEN (L.O.) except for ;
maintenance and. repair purposes'of thermal relief valves AEV0978 AND AEV0979 respectively. . Valves AEHV0034 and AEHV0017 shall be opened before l closing VALVE AEV0361. Valves AEHV0033 AND AEHV0018 shall be opened before closing valve AEV0362." l The underlying assumption is that the pressure relief device is the sole ,
means of preventing overpressurization of the vessel. Continuously l l manning the isolation valve would allow prompt opening of the valve should l l a pressurization event occur. The above assumption is not valid for the q high pressure feedwater heater, so long as the High Pressure Heater Train l inlet and outlet isolation valves (AEHV0018 and AEV0033 for train A, 1 AEV0017 and AEV0034 for train B) are open. The revised note continues to "
meet the requirements of the Code and will provide acceptable protection ;
against overpressurizing the high pressure feedwater heater. i l
Because no design basis accidents are affected the probability of occurrence of an accident is not affected. Because no design basis accidents are identified, the consequences of accidents are not affected.
'Because no malfunctions are identified the probability of occurrence of a malfunction is not affected. Because no malfunctions are identified the consequences of a malfunction are not affected. Because no credible accidents that could be created are identified no accidents of a different type can be created. Because no malfunctions are identified no malfunctions of a different type can be created. Because no acceptance limits are identified that could be affected, the margin of safety is not affected t
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Safety Evaluation: 59 98-0062 Revision 0 .
Organization Change 4
Updated Safety Analysis Report (USAR) Section 13.1-17 is being revised to {
add qualifications for Superintendent Operations Support. i USAR Section 13.1-9 is being revised to change reporting responsibility of the Shift Supervisors from Manager Operations to Superintendent operations.
USAR Figure 13.1-2c is being revised to change reporting of the Shift Supervisors from Manager Operations to Superintendent Operations and changes the Projects group reporting to Supervisor Operations Support.
l This USAR change does not change any administrative controls which would reduce the level of qualification of Wolf Creek Operating Corporation ,
personnel, nor does it affect any structure, system or component. These l changes also do not change the performance of activities that are important to the safe and reliable operation of Wolf Creek Generating ,
Station. The Superintendent Operations is a Licensed Operator and meets all the requirements of ANSI N18.7, ANS 3.1 and Technical Specifications.
Therefore, this. revision will not increase the probability of occurrence I or the consequences of an accident or malfunction of equipment important ;
to safety previously evaluated in the safety analysis report. This l revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.
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Page 105 of 217 l
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Safety Evaluation: 59 98-0064 Revision 0 Increasing High Pressure Condenser Shell Pressure Administrative Limit to e 5.5" Hga i Procedures ALR 00-118B, "Cond C Vac Lo," OFN AF-025, " Unit Limitations,"
and GEN 00-004, " Power Operation," and Updated Safety Analysis Report [
(USAR) Section 10.4.4.2.1 will be revised to reflect an increase in the high pressure condenser shell pressure administrative limit from 5.0" Hga '
to 5.5" Hga. This will allow operation _of the-turbine generator at loads 4 at or above 80 percent when cooling lake temperatures are at or above 88 l degrees Fahrenheit. This will allow operation at reduced plant efficiency and with reduced operating margin of the turbine generator.
i The design basis accident evaluated for.the effect of this activity would be a Decrease In Heat Removal By The Secondary System due to loss of s condenser vacuum and other events resulting in turbine trip. There are no radiological consequences associated with a Loss of Vacuum event. The i change in the administrative limit to allow high power operations (less than 80 percent load) with higher condenser back-pressure (5.5" Hga) will .
not create any credible accidents. The change in the administrative limit !
to allow high power operations (less than 80 percent load) with higher condenser back-pressure (5.5" Hga) will not cause a credible malfunction l of equipment important to safety. This change will not affect turbine trip setpoints. There are no acceptance limits in the bases for technical specifications affected by increasing the administrative condenser back-pressure limit to 5.5" Hga at high power operations (less than 80 percent ;
load). !
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Attachment II to ET 99-0003
( Page 106 of.217 l
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l Safety Evaluation: 59 98-0064 Revision: 1 l l 1
! Increasing the High Pressure Condenser Shell Pressure Administrative Limit ,
j to 5.5" HGA '
! This activity evaluates the same changes a Unreviewed Safety Question l Determination-(USQD) 59 98-0064 Revision 0, except for the provision which '
provided.for operation with cooling lake temperatures above 88 degrees j Fahrenheit.. Cooling lake temperatures above 88 degrees Fahrenheit are not i l evaluated by this revision because Wolf Creek Technical Specifications no longer allow operation of the plant at those temperatures.
Procedures ALR 00-118B, "Cond C Vac Lo,", OFN AF-025, " Unit Limitations,"
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and GEN 00-004, " Power Operation," and Updated Safety Analysis Report ,
(USAR) Section 10,4.4.2.1 will be revised to reflect an increase in the '
i high pressure condenser shell pressure administrative limit from 5,0" Hga to 5.5" Hga. This will allow operation of the turbine generator at loads ,
at or above 80 percent. This will allow operation at reduced plant i efficiency and with reduced operating margin of the turbine generator.
The design basis accident evaluated for the effect of this activity would be a Decrease In Heat Removal By The Secondary System due to loss of l condenser vacuum and other events resulting in turbine trip. There are no 5 radiological consequences associated with a Loss of Vacuum event. The change in the administrative limit to allow high power operations (less than 80 percent load) with higher condenser back-pressure (5.5" Hga) will l not create any credible accidents. The change in the administrative limit ,
( to allow high power operations (less than 80 percent load) with higher l condenser back-pressure (5.5" Hga) will not cause a credible malfunction of equipment important to safety. This change will not affect turbine trip setpoints. There are no acceptance limits in the bases for technical l specifications affected by increasing the administrative condenser back-pressure limit to 5.5" Hga at high power operations (less than 80 percent .
load).
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l Attachment II to ET 99-0003 Page 107 of 217 Safety Evaluation: 59 98-0065 Revision: 0 l
l Reactor Building Hatch Modifications This modification provides the following changes:
1 Reactor Building Personnel Hatch ( ZXO3 ) : .In plant modes 1, 2, 3, and 4: a) Bearing lugs 1, 8, 13, and 20 to be fully effective, b) Upper ,
Seismic Restraints 9 through 12 to be fully effective. c) Lower Seismic Restraints 21 and 22 to be fully effective. d) The two 1-1/2 diameter Pins to be fully engaged. Bearing lugs 2 through 7 and 14 through 19 may be permanently removed. During outages, in Modes 5 and 6 and with the fuel removed from M.e reactor, the Reactor Building Personnel Hatch ( ZXO3 ) may be permanently moved such that it does not interfere with the Reactor building opening at elevation 2045'-6". These configuration changes are shown in drawing C-1S2907 Revision 0.
2 Reactor Building Equipment Hatch ( ZX01 ): Reactor Building Equipment Hatch'( ZX01 ) to be in place with 6 bolts during fuel movement or core alterations as shown in drawing C-151-00400-WO7. The Reactor Building Personnel = Hatch ( ZXO3 ) is not required for tornado generated missile protection.
3 Station procedures: Administrative controls shall be in place to have the Reactor Building Equipment Hatch ( ZX01 ) in place with 6 bolts within 60 minutes upon the approach of threatening weather conditions.
4 Updated Safety Analysis Report (USAR) Section 3.8.2.1.1 is revised to reflect a movable missile shield is provided on the cutside of the reactor
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building to protect the equipment hatch. Administrative controls ensure ,
the hatch cover is in place during the threat of severe weather. ;
1 An analysis has been performed to show the Reactor Building Equipment '
Hatch is capable of protecting the safety related equipment inside l containment against tornado generated missiles without the Containment j Equipment Hatch Radiation and Missile Shield. The calculations meet the i requirements of GDC-2 and Bechtel topical report BC-TOP-3-A. Therefore accidents identified in the USAR are not impacted by this change.
Therefore, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. 1 This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the
. safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.
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Safety Evaluation 59.98-0068 Revision 3 0 4 Updated Safety Analysis Report Change to Residual Heat Removal Pump Seal Cooler Flow Updated Safety Analysis Report (USAR) Tables 9.2-10 & 9.2-11will be revised to clarify flow requirements of the Residual Heat Removal (RHR) ,
pump seal cooler. A minimum seal cooler flow of 4 gpm should be used for -
this table.
Currently RHR pump seal flows are presented inconsistently in the USAR.
USAR Table 9.2-9, Component Cooling Water (CCW) Flows for Normal Operation, lists 6 gpm as the nominal seal cooler flow. USAR Table 9.2-10, j CCW Flows at Four Hours after Shutdown and USAR Table 9.2-11, Post-LOCA .
CCW Flows, list 6 gpm as the minimum flow to the RHR pump seal cooler. '
The normal and four hour cases are normal (non-accident) plant operating modes with the CCW service loop assumed to be in use. A minimum flow of six gpm to the seal coolers cannot be assured in these normal operating '
modes. The manufacturer identifies the seal flow requirement as 4 to 6 gpm. Four (4) gpm is recommended here as-a reasonable minimum seal cooler flow for the four hour shutdown mode (USAR Table 9.2-10) and for post-LOCA operation (USAR Table 9.2-11).
1 The RHR pumps are used for normal shutdown and safe shutdown following ;
j design basis accidents described in the USAR. Any reduction in seal ,
cooler flow resulting from this activity will have no adverse effect on the RER pump seals and, therefore, the RHR pumps, as the reduced flows are consistent with the manufacturer's recommendation.
The proposed activity only changes the allowed minimum flow to the RHR pump seal coolers. No new type of accident is created. Therefore, no credible malfunctions of equipment important to safety are directly or ;
indirectly affected by this proposed activity, j l
CCW cooling water flowrates are not specifically identified in the l Technical Specification bases. The operability of the RHR pumps is not affected by the proposed activity. Therefore, the acceptance limits contained in the bases for the Technical Specifications are not affected.
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- Pa36 109 of 217 l
Safety Evaluation: 59 98-0069 Revision: 0 ,
-Instrument Air Correction I 4
Instrument Air (KA) . valve KAHV0030 is designed as a controllable valve inside the containment building to provide a make-up air source to dilute j
the post LOCA containment building hydrogen concentration. The safety related portion of the Combustible Gas Control system is provided by the !
Hydrogen Recombiners and the Emergency Exhaust System. The addition of compressed. air to the containment is discussed in Updated Safety Analysis i
- Report (USAR) Section 6.2.5.2.2.4. However, valve.KAHV0030 is not relied j upon as the sole means of introducing compressed air for atmospheric j dilution. i The downstream end of line 164-HCD-1M" MUST remain open when valve i KAHV0030 is operable. However, the current wording of the note unduly restricts the availability windows for maintenance on this portion of the KA System. ,
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CCP 07795 will modify Note 2 on drawing M-12KA01, and USAR Figure 9.3 l' 01, and Note 8 on drawing M-13KA43 to read: " Pipe end must remain OPEN whenever valve KAHV30 is considered operable." !
The safety related portion of the. Combustible Gas Control System is {
provided by the Hydrogen Recombiners and the emergency Exhaust system.
Valve KAHV0030 is designed as a controllable valve inside the containment building to provide a make-up air source to dilute the post LOCAA containment building hydrogen concentration. The addition of compressed j air to the containment building aa discussed in USAR Section 6.2.5.2.2.4, '
shows valve that KAHV0030 is not relied upon as the sole mans of I introducing compressed air for atmospheric dilution. The air line will be [
open when valve KAHV0030 is operable. There are no design basis accidents 'I impacted with this change. ;
Since valve KARV0030 is not relied upon as the sole means of introducing $
compressed air for atmospheric dilution, and the air line will be open when the valve is operable,-there is no affect on any system structure, or 1 component. No credible accidents that could be created are identified.
l The proposed change does not affect valve KAHV0030 position or flow rate.
Therefore there is no affect on any system, structure, or component.
l There are no credible malfunctions of equipment important to safety j identified. Review of Technical Specifications and the bases identified l I
no impact on acceptance limits or the margin of safety.
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Page 110 of 217 t
I Safety Evaluation 59 98-0070 Revision: 0 I
Department of Transpc a . ion Regulation Changes
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This revision to the bpdated Safety Analysis Report (USAR) affects the i solid waste management system's storage containers as described in USAR l Sections 11.4.2.4 and 11.4A.3. The USAR references the requirements of 49 CFR 178.125 and 178.118 which have been removed from the Code of Federal :
Regulations. The USAR is being changed to specify packaging that meets t the requirements of Department of Transportation (DOT) approved containers or the current burial site regulations. These changes are a result of a revision to Title 49 of the Code of Federal Regulations, f The following USAR sections are being changed: Section 31.4.2.4, the sixth t pardgraph will be changed to read: "The 55-gallon drums used in the solid
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radwaste system meet the requirements of DOT approved containers." In Section 11.4A.3, the first sentence will be changed to read: " Containers used for packaging of radioactive material, and stored in the IOS, shall l meet the applicable DOT requirements for quantity and form or the current I burial site regulations for disposal (HIC) when placed in storage." In ;
Section 11.4A.3, a typographical error will be corrected. !
These changes to update the USAR reflect a change in regulations. As ;
e such, this revision will not increase the probability of occurrence or the !
consequences of an accident or malfunction of' equipment important to i safety previously evaluated in the safety analysis report. This revision l does not create a possibility for an accident or malfunction of a [
different type than any evaluated previously in the safety analysis ,
report. The margin of safety, as defined in technical specifications, is ,
not reduced by this revision. Therefore, this revision does not involve l any unreviewed safety question.
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Safety Evaluation: 59 98-0071 Revision: 0 I
Post Accident Sampling System Evaluation !
The intent of this evaluation is to formally determine and document l whether the long term effect of the non-functionality of the Post Accident j Sampling System (PASS) Reactor Coolant System (RCS) dissolved hydrogen l analyzer is an unreviewed safety question. In addition, Wolf Creek l Nuclear Operating Corporation (WCNOC) committed in-Letter WM 98-0069, ,
dated July 7, 1998, to complete this evaluation by August 28, 1998. l The PASS RCS dissolved hydrogen analyzer has been inoperable since April '
22, 1992, and in the years prior to that was an extremely high maintenance !
item. WCNOC has submitted to the NRC a request for relief from the j requirements of NUREG 0737 for the analysis of RCS *otal gases and specifically RCS dissolved hydrogen via Letter #WO 98-0047, dated May 11, ,
1998.
l WCNOC Response to the Regulatory Requirements: USAR Section 3A, USAR Table 7A and USAR section 18.2.3.2 describe the design and methodology used to accomplish the required analysis.
The hydrogen analyzer was taken out of service on April 22, 1992 and 1 remains out of. service. In the following years, no field work was l' completed to restore the system to an operable condition. Notice of Violation 50-482/9813-02 was issued which identified that placing the PASS RCS hydrogen analyzer out of service for such a long period of time, ,
without NRC review and approval, constituted a potential unreviewed safety !
question.
J Due to the safety-classification of the equipment required for the l operation of the RCS dissolved hydrogen analyzer (i.e. non-safety), there are no guarantees or certainties that the equipment would be functional in an accident condition. The basic design requirements do not exist to ensure that the equipment would function as needed. The ability of plant personnel (i.e. Operations or E-Plan) to determine hydrogen gas concentration is not negatively impacted by the inability to sample and analyse RCS dissolved hydrogen.
The PASS RCS dissolved hydrogen analyzer is not one of the initiating events or precursors for any accident evaluated in the USAR. The system could be used at the discretion of Chemistry and perations personnel to evaluate the severity of an accident after the accident has occurred. As such, since it is not part of the initiation of any accident scenario, there will be no increased probability of an accident due non-functionality of the RCS dissolved hydrogen analyzer for over 6 years.
The PASS RCS dissolved hydrogen analyzer could be used as an evaluative tool after an accident has occurred. The system will have no impact on
i Attachment II to ET 99-0003 Page 112 of 217 the initiation, therefore any radiological consequences would have already occurred as a result of the accident. The PASS RCS dissolved hydrogen analyzer would then be used to help determine the full scope and severity of the accident.
The PASS panel and specifically the PASS RCS dissolved hydrogen analyzer are classified as non-safety related. In the location of the panel, there is neither any safety related or important to safety equipment. One of i the potential reasons for sampling the RCS for hydrogen is. to help '
determine if reduction of RCS natural circulation flow due to obstructed flow channels from dissolved gas build-up is probable, i
The non-functionality of the PASS RCS dissolved hydrogen analyzer since i 1992 will not have any impact on any malfunction of equipment important to safety. By not impacting any equipment important to safety, there will be no increased radiological consequence.
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Attacraent II to ET 99-0003 Page 113 of 217 Safety Evaluation 59 98-0072 Revision 0 Instrument Upgrade Process Liquid Sampling and Analysis System (RM) Panel RM171 in the Turbine Building Process Sampling Room contains instruments which do not operate properly. Many of the instruments are obsolete and spare parts are no longer available. This modification provides an upgrade to the instrumentation in panel RM171 so that it will perform the required analytical, display, and alarm functions.
The sample lines in panel RM172 will be modified so that conductivity analyzer RMCIT0480 and pH analyzer RMAIT0484 will be monitoring Moisture Separator Reheater Drains A, B, C and D instead of the Circulating Water Inlet. The analyzers and sample line components will be assigned new component numbers for the new application.
The original Beckman conductivity analyzers and the corresponding conductivity cells will be replaced with Rosemount analyzers and cells.
The Leeds & Northrup chart recorders will be replaced with Yokogawa chart recorders. All alarm functions originally performed by the chart recorders will be transferred to the various analyzers. The Leeds &
Northrup chart recorders will be replaced with Yokogawa chart recorders and a programmable controller. Oxygen Analyzers will be replaced with up to date analyzers which have alarm capabilities. Alarms will be activated by the analyzers rather than by the chart recorders.
Alarm modules will be added to sodium analyzers. The alarm modules will perform the alarm functions originally performed by the chart recorders.
Updated Safety Analysis Report (USAR) Table 9.3-5 is revised to incorporate alarm point changes authorized by procedure ALR 412, Rev. 1.
The ranges for several parameters listed in the table have been revised to more accurately define the values measured The modifications proposed in DCP 06544 do not affect any accident discussed in the USAR. Therefore, the probability of occurrence of an ;
acci5ent previously evaluated in the USAR and the radiological consequences of those accidents are unaffected.
l The modifications proposed in DCP 06544 cannot create a malfunction of equipment important to safety. Therefore, the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR is unaffected, and the radiological consequences of a malfunction of equipment will not be affected. The instruments associated with this modification are used to measure turbine cycle water quality and have no affect on plant safety. Therefore, the changes have no affect on the possibility of a different type of malfunction of equipment important to
Attachment II to ET 99-0003 Page 114 of 217 safety than any previously evaluated in the USAR.
Adjustment of the process stream measurement ranges listed in the USAR will more accurately define plant requirements. The range changes do not affect any margins of safety in Technical Specifications.
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! l l Safety Evaluations' 59 98-0073 Revision: 0 i
l Clarification of Relief valve Indications l i The proposed activity involves a text change to Updated Safety Analysis ,
l Report (USAR) Section 5.2.5.2 to clarify statements concerning indications j of relief valve lifting. The USAR text is being changed to read: " Relief j valve lifting is detected by increasing levels of boron recycle holdup !
j tanks which indicate and alarm in the radwaste control room and provide a l
- general system alarm in the main control room". Clarification is needed I i because existing text refers only to " control room" which is generally 3 j understood to mean the main control room. However, the specific '
] indications and alarms under discussion are in the radwaste control room.
l The USAR change is made to provide clear and concise statements with i
t respect to the location of indications and alarms, and is consistent with ;
! information provided on USAR Figures 9.3-11-01 and 9.3-11-02.
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! Because no' design basis accidents are identified the probability _.of f f occurrence of an accident is not affected. Because no design basis {
l accidents are identified, the consequences of accidents are not affected.
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i Because no malfunctions are identified the probability of occurrence of a j malfunction is not affected. Because no malfunctions are identified the $
consequences of a malfunction are not affected. Because no credible accidents that could be created are identified no accidents of a'different l type can'be created. Because no malfunctions are identified no l 1
malfunctions of a different type can be created. Because no acceptance j limits are identified that could be affected, the margin of safety is not :
affected.
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I Safety Evaluation: 59 98-0074 Revision: 0 Technical Requirements Manual i A new document, " Technical Requirements Manual," is being created to j contain and control requirements that have been relocated from the Technical Specifications. These requirements are currently located in ;
Updated Safety Analysis Report (USAR) Chapter 16 which is called the ;
Operational Requirements Manual. A USAR change request associated with this document creation will remove the Operational Requirements Manual from USAR Chapter 16 and replace it with a specific reference in Chapter ;
16 to the Technical Requirements Manual (TRM). The specific reference in :
Chapter 16 clarifies that the TRM is considered part of the USAR.
The NRC Final Policy Statement on Technical Specifications Improvements f for Nuclear Power Reactors (58 FR 39132) allows remeval and relocacion of {
Technical Specification requirements which meet certain criteria. The ;
following statements are contained in the final policy statement-In section IV, "The Commission Policr": I "LCOs which do not meet any of the criteria below may be propomsd for ,
removal from the Technical Specifications and relocation to licensee-l controlled documents, such as the FSAR."
t In section V, " Enforcement Policy"- #
"If a licensee elects to apply these criteria, the requirements of the l
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removed specifications will be relocated to the FSAR or other licensee-controlled documents. Licensees are to operate their facilities in conformance with the descriptions of their facilities and procedures in their FSAR. Changes to the facility or to procedures described in the i FSAR are to be made in accordance with 10 CFR 50.59."
NRC Administrative Letter 96-04, " Efficient Adoption of Improved Standard i Technical Specifications" states: "The staff believes that for consistency 1 and clarity, licensees should incorporate the details of the relocated >
technical specification requirements for which 10 CFR 50.59 is needed to [
control future changes, directly in the FSAR or in the Bases for the improved standard technical specifications." The adminietrative letter !
further indicates that an acceptable approach is to incorporate the j details of the relocated technical specification requirements into a manual, and then reference the manual in the FSAR.
Administrative procedures contain provisions to ensure that the !
requirements of the TRM are implemented and that any changes to the TRM i requirements are considered changes to the USAR and require an Unreviewed l Safety Question Determination. Therefore, the policy statement provisions !
discussed above continue to be implemented after the creation of TRM. The requirements removed from Technical Specifications continue to be required actions per administrative procedures, and changes to these requirements l
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must be evaluated in accordance with 10 CFR 50.59.
r Two technical changes are being made to the Operations Requirements Manual when it is converted to the TRM. Each is associated with approved USAR i Change Requests which have approved Unrviewed Safety Question .
Determinations (USQDs) so those changes will not be further justified '
here. No other technical changes will be made to the Operations Requirements Manual, however it will be re-paginated when converted to the ;
TRM and a list of effected pages will be added, j The Operational Requirements Manual from USAR Chapter 16 is being removed and replaced with a specific reference to the TRM. Administrative i procedures contain provisions to ensure that the requirements of the TRM are implemented and that any changes to the TRM requirements are considered changes to the USAR and require an Unreviewed Safety Question '
Determ!sation. Therefore, the requirements removed from Technical ,
Specifications continue to be required actions, and changes to these requirements must be evaluated in accordance with 10 CFR 50.59.
The result is no reduction in the operational restrictions associated with the Operational Requirements Manual in USAR Chapter 16. All of the restrictions associated with the relocated Technical Specifications ,en they were in Chapter 16 continue to be restrictions in the TRM. !
Therefore, there is no impact on accidents described in the USAR. Since ,
there is no reduction in the operational restrictions associated with the i Operational Requirements Manual in USAR Chapter 16, there are no accidents that could be created. Since the requirements removed from Technical Specifications continue to be required actions, and changes to these ,
requirements must be evaluated in accordance with 10 CFR 50.59, creation !
of a TRM will not affect malfunctions of equipment important to safety.
Since the requirements removed from Technical Specifications continue to !
be required actions, and changes to these requirements must be evaluated l in accordance with 10 CFR 50.59, creation of a TRM will not affect any acceptance limits. i l
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Attachment II to ET 99-0003 Page 118 of 217 Safety Evaluation
- 59 98-0075 Revision: 0 Revision to Security Procedure Revision 24 to procedure SEC 01-206, "High Security Key Control and Issue," deletes the requirement to change or rotate, every twelve (12) months, all keys and lock cylinders used to control access to protected and vital areas. Changes to 10 CFR 73.55 (d) (8) deleted the requirement to rotate all keys, locks, combinations and related access control devices used to control access to protected and vital areas every twelve (12) months. The change is being made pursuant to the change to 10 CFR 73.55.
Because this change implements recent revisions to the Code of Federal Regulations, this revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question, l
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Page 119 of 217 :
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1 Safety Evaluation: 59 98-0076 Revision 0 Opdated Safety Analysis Report Description Correction '
Updated Safety Analysis Report (USAR) paragraph 12.1.2.5.d.1 incorrectly }
states that, to alleviate airborne radioactivity in containment, Wolf l Creek Generating Station design provisions include, "A packless, low- l leakage, ball-type pressurizer spray valve," The valves under discussion !
utilize packing, per approved plant design, to prevent leakage around the i valve stem. The design is consistent with USAR Section 5.4.12.2 which states that throttling type control valves in the Reactor Coolant System (which includes the pressurizer spray valves) are provided with graphite packing. To assure low leakage, the packing design was previously l enhanced by installing live-loading packing under Plant Modification [
Request 02232.
i Therefore, the text is being changed to read; "A low-leakage, ball-type ;
pressurizer spray' valve". The USAR change is made to provide a correct !
description of the pressurizer spray valves. !
The proposed change does not. affect plant design or other information !
provided in the USAR. The proposed correction of USAR text has no effect i on any procedures, activities, administrative controls, or sequences of I plant operations because the clarification is consistent with current i plant design. Also, the correction of USAR text does not constitute a !
test or experiment. A review of USAR analyzed accidents reveals that none ,
are associated with, or affected by, the text change being evaluated. No !
credible accidents are created by a simple USAR text change to correct a j component description. ;
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- 7.ttachment II to ET 99-0003 Page 120 of 217-Safety Evaluation
- 59 98-0077 Revision 3 0 Thermal Piping Insulation Inside Containment The Updated Safety Analysis Report (USAR) specifically identifies NUKON brand insulation. This revision will delete the specific insulation brand and provides an alternative. Configuration Change Package (CCP) 07810 has evaluated the equivalent product for similar characteristics and has concluded that a technical and safety equivalency exists.
USAR Section 6.2.2.1.2.2 discusses the containment recirculation sumps that are used as reservoirs for the containment spray pumps and the Residual Heat Removal (RHR) pumps after the Refueling Water Storage Tank (RWST) has been expended. The sumps are located as far away as possible from the Reactor Coolant System (RCS) components and piping which could become sources of debris. Debris inside containment is a concern because of the potential for adversely affecting the sump due to blockage. The sump is designed to mitigate any potential debris consequences with the use of grating, course screening and fine screening. This should prevent floating debris and high-density particles from entering. The USAR discusses that thermal insulation installed inside containment is not considered a significant source of debris. The USAR references Topical Report OCF-1 which was prepared by Performance Contracting Inc. and discusses the safety aspects of NUKON thermal insulation. The USAR states that most.of the insulation inside containment is NUKON. The USAR states that insulation other '. an NUKON has also been evaluated to ensure that it will not degrade under DBA conditions and become a source of debris.
These references to NUKON are deleted by this CCP.
USAR Table 6.1-6 presents Wolf Creek Generating Station (WCGS) compliance with Regulatory Guide (RG) 1.36, Revision 0, dated February 1973, titled
" Nonmetallic Thermal Insulation for Austenitic Stainless Steel". In essence the RG requires that the levels of leachable contaminants in any insulation that comes in contact with stainless steel fluid systems important to safety should be controlled to prevent the promotion of stress corrosion cracking. USAR Sections 5.2.3.4.1 also discusses the need for protection of austenitic stainless steel materials used in the fabrication and installation of the nuclear steam supply system to prevent contamination which could lead to stress corrosion cracking. USAR Section 5.4.3.3.3 states that prior to application of thermal insulation, austenitic stainless steel surfaces are cleaned and analyzed to halogen limits as defined in Westinghouse Process Specifications. The equivalent Transco Thermal-Wrap insulation has been tested in accordance with the requirements of RG 1.36. Thus, the expected affect of the approval of the alternate manufacturer is that there will be no impact upon safety since the alternate meets equivalent requirements as the original.
CCP 07810 demonstrates the technical and safety equivalency of Transco Thermal-Wrap fiberglass insulation as compared to NUKON fiberglass l
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Attachment II to ET 99-0003
-Page 121 of 217 insulation. The CCP will allow Transco Thermal-Wrap to be installed at WCGS anywhere that the original NUKON insulation has been installed inside containment. USAR Section 6.2.2.1.2.-2 is affected by this change.
Reference to NUKON will be deleted. There are no other sections which would be impacted.
The accidents discussed in the USAR were reviewed including both the small and large break-LOCA, inadvertent operation of the Emergency Core Cooling System (ECCS) during normal operation ar.d all scenarios requiring the operation of the containment. spray or residual heat removal systems.
Because the insulation has been shown to be equivalent, there are no new credible accidents that this change could create.
Debris generated as a result of a LOCA could credibly block the containment Recirculation sump however, as described in the USAR the debris will either be retained in the reactor cavity or refueling pool or must follow a tortuous path to reach the recirculation sump screens.
Therefore, this change will have no effect on sump blockage.
Austenitic stainless steel materials used in the fabrication and installation of the nuclear supply steam system which might be contaminated from insulation could credibly lead to stress corrosion cracking. Thus, prior to application of thermal insulation, austenitic stainless' steel surfaces in the NSSS are cleaned and analyzed to halogen ,
limits as defined in Westinghouse Process Specifications. Therefore, this '
change will have no effect on stress corrosion cracking.
There are no acceptance limits contained in th; Technical Specifications )
that would be directly or indirectly affected by this change.
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} i j safety Evaluation: 59 98-0078 Revision 0 Security Computer Replacement f 1
i This modification will provide fiber optic cabling for the security computer replacement project. It will also provide spare cable to support j future modifications for the Nuclear Plant Information System (NPIS) data j monitoring at the Essential Service Water (ESW) pumphouse. The Security z
- Computer portien cf the job involves cabling to interconnect the hest s l computer with all peripherals, including the field multiplexers and l operator / administrator workstations. These components will be located at {
various points within the Protected Area and at the ESW Pumphouse. The 2 .NPIS. portion of the job involves installing a spare cable between the ESW l j Pumphouse and the back of the Control Room. l The scope of this change involves the cable additions, with the end :
- equipment (microprocessors) being installed on a later revision of this change package, or on separate change packages. The effect of this change ,
{ on the Updated Safety Analysis Report (USAR) will be the addition of l l~ combustible loading in the various fire zones in the power block. The l I additions are identified in USAR Section 9.5B. The cables will be routed i in non-safety related raceway in the power block and to the ESW, and in security ductbank for the yard portion of the installation. The cable installation will be performed in accordance with existing criteria' identified in design documents. No safety related equipment will be affected in the Power Block. At the ESW, a seismic wall will be used to mount a termination cabinet for the fiber. The cabinet will be installed in accordance with existing civil design details identified on the applicable design drawings. This will ensure that all safety related l design criteria is adhered to and that the non safety-related cabinet will not be a seismic II/I concern.
Section 9.5B of the USAR contains a listing of the fire loading in the various zones within the power block. The new cable installed as part of this modification will increase by a small amount the combustible loading in several rooms within the power block (i.e. room # 1101, 1102, 1121, 1122, 1301, 1314, 1320, 1335, 1403, 1408, 3230, 3301, 3306, 3419, 3501, 3SO4, & 7133). USAR Section 9.5B is being revised to indicate the new combustible loading values. The additional combustible loading is due to cable insulation. The new combustible loading for the applicable rooms is still less-than the value needed to declare an area a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating fi.e. 240,000 BTU's/f t2) . USAR Sections 9.5.1.2.2.3 and Table 9.5A-1 reference ASTM E-119 as the applicable fire standard. ASTM E-119 provides 240,000 BTU's/ft2 as a conservative loading estimate for sustaining a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire. 'Specifically, ASTM E-119, paragraph X5.3 notes that cpecifications for fire resistance in regulatory documents continue to be based largely on the fire load concept. The concept incorporates the premise that the duration of a fire is proportional to the fire loading, i I
that is, the mass of combustible materials per unit floor area. The
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Attachment II to ET 99-0003 Page 123 of 217 relationship between the mass of combustible materials and fire duration l was established on the basis of burnout tests in structures incorporating materials having caloric or potential heat values equivalent to wood and paper, that is 7000-8000 BTU /lb. The above premise states that 10 lb. of material per square foot (80,000 BTU /ft2) will produce a fire of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> duration. The NFPA Fire Protection Handbook, page 7-111, also documents this conservative estimate. This estimation is additionally conservative based upon the cable being specified in accordance with IELE-383 (i.e.
qualified for flame retardency).
The fiber optic cable will be installed in accordance with existing ,
separation requirements. This will ensure that safety related equipment will not be adversely affected by the installation of the non-safety !
related fiber optic cable. Because of this criteria compliance, no new considerations exist for accidents identified in the USAR.
The combustible loading that is being added is within the analyzed bounds for the affected zones. The cable is being installed per the approved separation requirements. Based on these facts, and the fact that the cable is a passive material, no credible accidents will be created by the addition of the cable.
Based on the fact that the cable is being installed per the approved separation requirements and the combustible loading is within the analyzed bounds for the affected zones, there is no affect on malfunctions of l equipment important to safety, either direct or indirect.
l With respect to the cable additions, the acceptance limits for combustible loading within the various fire zones has been analyzed and found to be acceptable. The cabinet mounting at the ESW Pumphouse is within analyzed bounds due to the use of seismic mounting details on existing design drawings. The design basis requirements for cable separation and seismic mounting of non safety-related cabinets, are met by the fiber optic cable additions identified on this change package.
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Page 124 of 217 I i
i Safety Evaluation: 59 98-0079 Revision: 0 l
Updated Safety Analysis Report Change to Boron Injection Operation Requirements l Updated Safety Analysis Report (USAR) Section 16.1.2.3 requires that one !
centrifugal charging pump in the boron injection flow path required by i USAR Section 16.1.2.1 shall be OPERABLE and capable of being powered from l an OPERABLE emergency power source. A note for this LCO allows in Mode 6, )
with reactor pressure vessel head removed, an inoperable centrifugal :
charging pump to be substituted by an OPERABLE Safety Injection (SI) pump. j USAR Section 16.1.2.1 requires that as a minimum one of the two boron injection flow paths (Refueling Water Storage Tank or Boric Acid Storage system) be operable with an OPERABLE Centrifugal Charging Pump. l This USAR change clarifies that the SI pump can be substituted for an inoperable CCP in the boron injection flow path from Refueling Water ,
Storage Tank, as an SI pump cannot be used for boron injection from Boric l Acid Storage System. j This change is a document change only. It has no physical impact on the plant. EIt does not change the function of any equipment. It does not' {
change any operating parameters. This change clarifies a note in the USAR l to reflect that an OPERABLE Safety Injection Pump is substituted for in inoperable centrifugal charging pump in the flow path for the Refueling Water Storage Tank only. Therefore, it has no impact on the accidents and ;
malfunctions of equipment mportant to safety previously evaluated in the l USAR. There is no potential to create an new type of an accident. There l la no impact on any margin of safety. This change does not constitute an unreviewed safety question.
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Attachment II to ET 99-0003'-
Page 125 of 217 E
i Safety Evaluation: 59 98-0080 Revision 3 0 l l
Updated Safety Analysis Report Description Changes j
This revision to the Updated Safety Analysis Report (USAR) provides ,
updates to Section 9.5 and 9.5A of the USAR. This change incorporates I title changes, Section renumbering, building name changes, procedure ,
number and name changes, and defines, " Smoking Areas," In addition, titles for off site authorities are updated, and clarification to the Fire l Protection Program is provided.
The above USAR changes are minor editorial changes that clarify the general description, system operation, component description and j administrative statements to reflect the way the fire protection system i and program is operated. Therefore, this revision will not increase the !
probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the !
safety analysis report. This revision does not create a possibility for j an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision.
Therefore, tnis revision does not involve any unreviewed safety question, i
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.i Attachment II to ET 99-0003 t Page 126 of 217 Safety Evaluation: 59 98-0081 Revision: 0 t Updated Safety Analysis P.eport Change to Reflect As-built Conditions The proposed activity being evalusted herein involves an Updated Safety Analysis Report (USAR) change to properly reflect as-built conditions ,
associated with removal of covers over the cask loading pool. These :
covers were removed prior to initial start-up and efforts in locating {
documentation of the final disposition of these covers was unsuccessful. l The function of the covers was to prevent objects (equipment / personnel)
- from accessing the cask loading pool, protect handling tools which are hung in the cask loading pool, and allow for additional space in the exclusion area. The requisite for covers over the cask loading pool has been eliminated. Planning and performing heavy lifts and transferring heavy loads within the power block (including cask loading pool) is !
governed under the Administrative Controls Procedure AP 14-001 Revision.
O, " Control of Heavy Loads." Other administrative controls require ,
workers to wear flotation devices or safety harnesses when working within six feet of the cask loading pool. The USAR change effects USAR Figures:
1.2-20, 1.2-22, 3.8-96, 3.8-98 and 9.3-7.
The proposed activity has no effect on any procedure, activity, administrative control, or sequence of plant operations and does not violate any requirement stated in the USAR. Neither the proposed activity i nor the design change are associated with any test or experiment. The i proposed activity is a correction of USAR text (figures) to properly reflect as-built conditions associated with the cask loading pool design change. Neither the proposed activity nor the design change has any influence on equipment or parameters associated with any design basis ,
accident discussed or referenced in the USAR. '
This change properly reflects as-built conditions. Due to administrative ,
controls within the power block (including cask loading pool), neither the l proposed activity nor the design change creates a potential for, or has ;
influence on, equipment or parameters that may cause any new credible '
accidents. The proposed activity will properly reflect as-built !
conditions in the applicable USAR figures. There are no malfunctions of i equipment important to safety which may be directly or indirectly j associated with the proposed activity or the design change. Neither the l proposed activity nor the design change affects acceptance limits as !
defined in the Technical Specifications or other licensing basis documents. l
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r Safety Evaluationt 59 98-0082 Revisions 0
- Circulating Water system Description Changes !
This Unreviewed Safety Question Determination evaluates Change Package: ;
07857,. " Circulating Water System Description," and Updated Safety Analysis ;
Report (USAR) corrections. As part of the USAR fidelity review, potential ,
discrepancies and inconsistencies were noted in USAR Section 10.4.5, which
- describes the circulating water system. The following USAR discrepancies ,
were found and are corrected or clarified:
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- 1) USAR Section 1.2.2.m, states that chlorination.and acid feed equipment f for the service water system is located there. The acid feed equipment has l been removed under Design Change Package (DCP) 06460. This section is revised to correctly reflect to that chlorine is no longer used in this application and acid feed equipment is not used, t
- 2) USAR Section 10.4.1.2.1 has a description of the new tube cleaning '
system installed undcr DCP 05861. This equipment would be more appropriately scoped under the Circulating Water System. This paragraph ;
describing the condenser tube cleaning system is moved to USAR Section '
10.4.5.2.2.
- 3) USAR Section 10.4.1.5 has a description of the condenser tube cleaning system that is redundant to the previously referenced description and will be deleted, i
- 4) USAR Section 10.4.5 does not list the condenser tube cleaning system as j a major subsystem. The on-line condenser tube cleaning system is added to this paragraph.
- 5) USAR Section 10.4.5.1.2, Power Generation Design Basis Two, the as presently worded implies all three pumps are automatically secured upon a j high condenser pit water level. In the original and existing design ,
logic, two out of three circulation water pumps are secured, one pump will always have to be manually secured upon high condenser pit water level. ;
This will be clarified. j l
Power Generation Design Basis Three specifies the condenser circulating water temperature rise and the inlet temperature ranges. These i temperatures were used in original condenser design; however these i temperatures will be deleted to remove the possibility of interpreting 1 these temperatures references as operating limits. References to expected or approximate temperature ranges will be included in USAR Section !
10.4.5.2.2.
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- 6) USAR Section 10.4.5.2.1, describes the traveling screens as intake screens. For consistency within the USAR, intake screens will be changed to traveling screens.
Attachment II to ET 99-0003 Page 128 of 217
- 7) USAR Section 10.4.5.2.2 the following clarifications need to be made:
The piping to and from the condenser is described as 144 inch inside diameter piping. The piping connections at the condenser standpipes are 120 inches. The 144 inch pipe connection is actually at the site to power block piping interface. This statement will be clarified.
Temperature rise across the condenser is specified as 31.4 degrees Fahrenheit (F), based on heat load and circulating water flow that was used in original design calculations. This will be changed to 32.5 degrees F during full power, three pump operation to reflect the actual temperature rise. The wording is change to reflect that this value is approximate,so the implication is not made that this is an acceptance criteria or limit.
Because of the recent frazil ice event at the circulating water screenhouse, it is appropriate to add a brief description of the warming line in this section.
Specific description of all the types of biocide that could be injected will be deleted. The all encompassing description of any type of biocide that could be used is not necessary and will not change the meaning to simply reference " biocide",
i Specific reference to a particular type of Anti-scale chemical is made.
A generic reference to anti-scale chemical will be incorporated in place j of the specific reference. !
- 8) USAR Section 10.4.5.2.3, incorrectly states that the circulating water pump discharge valves close on a high condenser pit level. Discharge valve operation is correctly described in the next section, 10.4.5.3, which is the safety evaluation. The information on pump discharge valve operation will be moved from section 10.4.5.3 to 10.4.5.2.3.
- 9) USAR Section 10.4.5.4, will be changed to describe the testing performed during initial plant startup in the past tense.
- 10) U ~R Section, 10.4.5.5 incorrectly states that indication is provided in the control room to identify open and closed positions of motor operated valves in the system. This is true for the pump discharge valves ,
but not the waterbox isolation valves. The appropriate corrections are made.
- 11) USAR Table 10.4-1, Condenser Design Data, will have a notation added to identify the data in this table as original design, is historical in I nature, and does not reflect actual operating values.
- 12) USAR Table 10.4-3, Circulating Water System Component
Description:
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Attachment II to ET 99-0003 i Page 129 of 217 ;
The first two paragraphs in this table are redundant to USAR Section ,
.1'.4.5.4.
O These paragraphs will be deleted. l The above floor circulating water piping is specified as 120 inch outside l
diameter, not 119 inch inside diameter. This is corrected.
The CWS is not a safety-related system; however, a flooding analysis of 5 the turbine building was' performed on the CWS which postulated a complete ;
rupture of a single expansion joint. It was assumed that the flow into the condenser pit consists of the water which can drain from both the upstream and downstream' side of the break. For conservatism, it was ;
assumed that the condenser circulating water isolation valves do not fully ;
close, sump volumes in the condenser pit were neglected, and the sump pugs were not operable, f
The USAR discrepancies noted above do not affect assumption or conclusions regarding the role of the circulating water system in these design basis j accidents. No other design basis accidents were identified in the USAR 2 which could potentially be impacted by the Circulating Water System. ,
I Since there are no physical changes and the design basis function of the l system is not affected by this change, no new types of accidents not previously analyzed could be created. Since the proposed change would not l affect the system's failure modes, the systems design. function, the level i of qualification, or equipment important to safety, no credible ;
malfunctions of equipment'important to safety are identified. The i circulating water system is not addressed in the Technical Specifications. j 1
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Attachment II to ET 99-0003 f i
Page 130 of 217 l i
' Safety Evaluation: 59 98-0083 Revision 0 !
Changes to Maintenance Organisation ,
This change deletes the position of Assistant Manager Maintenance and j changes the title of the Superintendent Electrical /I&C to Superintendent ;
Instrumentation & Electrical. The change is administrative only in that ,
one is a nomenclature change and the other is to delete a position that.is not required. All required job functions will continue to be maintained; j therefore, there is no effect on plant processes, functions, or systems.
No processes or procedures are affected by this change. The only portion !
of the Updated Safety Analysis Report (USAR) that is affected is to the l description for the Maintenance organization.
There are no design basis accidents that take credit for the position being deleted. A change in title also has no impact on design bases accidents. I l
1 Deleting an administrative position, not function, or changing a title ,
will not create any new or different accident because all functions j continue to be maintained. l There'are no credible malfunctions of equipment.important to safety which may be directly or indirectly affected by the change because the change is !
administrative in nature and all functions performed by the Maintenance department continue to be performed.
i These positions are not and were not part of acceptance for the Operating License. -The Assistant Manger Maintenance was a recently created position '
ar.d the title of the Superintendent Instrumentation & Electrical is a title change only. Therefore, there in no impact on acceptance limits.
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Attachment II to ET 99-0003 Page 131 of 217 !
l Safety Evaluation: 59 98-0084 Revision 0 Changes to the Domineralizer Water Storage and Transfer System Description Configuration Change Package (CCP) 07859 changes the Demineralizer Water Storage and Transfer System (AN) Description, M-10AN and the Updated l Safety Analysis Report (USAR) to address discrepancies and clarify the operation of the system.
The changes to the AN System Description are minor changes that clarified the general description, system operation and component description section to reflect the way the system is operated to supply demineralized water to plant components and systems. The Demineralizer Water Storage and Transfer System is non safety-related and none of those do:umentation corrections affect the design basis function of the system or its hazards and accident analysis in the USAR.
The following USAR sections are being changed:
9.2.3.2.1, paragraph 1, change the third sentence to read: Check valves are provided to preclude backflow from the DWSTS to the DWST, assuring that contamination of the source is precluded.
9.2.3.2.3, paragraph 5, change the first sentence to read: The supply of demineralized water to the demineralized water storage tank can be initiated manually or automatically controlled by the actuation of tank level switches which cycle pumps in the demineralized woter makeup system (Section 9.2. 3) .
Based on the description above, this revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision.
Therefore, this revision does not involve any unreviewed safety question.
Attachment II to ET 99-0003 Page 132 of 217 Safety Evaluation: 59 98-0086 Revision 0 Updated Safety Analysis Report Clarification on Nuclear Station Operators ,
This revision to Updated Safety Analysis Report (USAR) Section 13.1.2.2.1, Paragraph 6, clarifies that the Nuclear Station Operators (NSO) do not l perform component manipulations within the control Room. Specifically, the revision deletes the following phrase: " operating equipment from the ,
Control Room."
L This revision reorganized the Operations organization chart to correctly i show reporting hierarchy for Operating Crews. The organization chart incorrectly showed the Shift Supervisor directly reported to the Manager Operations. The chart was corrected to show the Shift Supervisor reports j to the-superintendent operations. The chart was further enhanced to show I the reporting hierarchy within the operating crews. Specifically, the 1 Supervising Operator and Shift Engineer report to the Shift Supervisor and the Reactor Operators and Station Operators report to the Superviaing Operators. The change adds the position of Shift Engineer to the-organization chart. The organization chart was modified to change the title of tne position of " Projects," to " Supervisor Operations Support."
These changes ensure that the USAR correctly reflects the operating ;
practices utilized at Wolf Creek Generating Station. This revision will not increase the probability of. occurrence or the consequences of an accident er malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision does not create a ;
possibility for an accident or malfunction of a different type than any l evaluated previously in the safety analysis report. The margin of safety, l 1
as defined in technical specifications, is not reduced by this revision.
Therefore, this revision does not involve any unreviewed safety question.
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I Safety Evaluation: 59 98-0087 Revision: 0 Updated Safety Analyis Report Revision (Technical Requirements Manual) to l
Snubber Requirements i
The proposed activity being evaluated involves an Updated Safety Analysis !
Report (USAR) revision. This change will resolve the inconsistencies
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between USAR Section-16 and the plant conditions. The USAR describes the !
requirements for surveillance test for Hydraulic Snubbers. This type of i snubber has been eliminated in the plant and there are no plans of using l this type of snubbers in the future. Therefore, this requirement is '
deemed obsolete and unnecessary. This change affects USAR Sub-Sections 16.7.2.1.1c, id, if , li and 16.7.2.1.2. j The proposed change does not affect any procedure, activity, administrative controls, sequence of plant operation or any plant structure, system, component or equipment. There is no effect on any experiment as described in the USAR. The proposed activity is a revision to the USAR text and will not have any influence on equipment or j parameters associated with any design basis accident that have been i previously described in the USAR. The proposed activity is a revision to j the text in the USAR and will not create any new type of credible accident.- There are no malfunction of equipment important to the safety !
which may directly or indirectly be associated with the proposed activity which is only a textual revision to the USAR. The proposed activity will not have any effect on the acceptance limits defined in the Technical' i Specification or other Licensing Basis documents.
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Page 134 of 217 Safety Evaluation: 59 98-0091 Revision 0 Updated Safety Analysis Report Revision to Correct Chemistry Specification Values This revision to the Updated Safety Analysis Report (USAR) corrects discrepancies that exist between USAR Table 9.2-16 and System Description M-10EC for various chemistry specification values associated with the Fuel Pool. A review of the potential discrepancies revealed that the USAR Table states that the values listed are " typical" and includes a footnote which states, " Actual plant water chemistry specifications can be found in the WCGS Chemistry Specification Manual." To avoid confusion, the USAR Table values are being changed to agree with the System Description values. This change is editorial in nature since the values remain
" typical" and the footnote remains applicable. The specific changes to ,
USAR Table 9.2-16 are: 1) Revise Fuel Pool values for Chlorides, Fluorides, Calcium and Magnesium from "less than".to "less than or equal to" the values indicated. 2) Revise Fuel Pool value for Boron from "2000/+50" to " greater than or equal to 2000". (Note: This change is consistent with the Boron value stated in USAR Sections 9.1.2.2, 7 9.1.3.2.3.1 and 9.1A.2.3.) Also, a discrepancy exists within USAR Table 9.1-4 in that the Boron concentration value of 2000 ppm is not indicated as a minimum value. Therefore, USAR Table 9.1-4 is also being revised to add the word ' minimum' for this parameter.
As indicated in Table 9.2-16, the values listed are " typical" and actual plant water chemistry specifications are provided in the AP 02-003,
" Chemistry Specification Manual " Also, the change in the Boron value in Tables 9.1-4 and 9.2-16 are needed to make the information consistent with USAR Sections 9.1.2.2, 9.1.3.2.3.1 and 9.1A.2.3 and plant Technical Specifications. The actual water chemistry specifications in AP 02-003 are more restrictive than the values listed in this table. As a result, changes made in the USAR Tables are considered editorial in nature. This change is for clarification purposes only and does not affect any procedures, activities, administrative controls or sequence of plant 5 operations as described in the USAR.
Because no design basis accidents are affected the probability of occurrence of an accident is not affected. Because no design basis accidents are affected, the consequences of accidents are not affected. .
Because no malfunctions are affected the probability of occurrence of a malfunction is not affected. Because no malfunctions are identified the consequences of a malfunction are not affected. Because no credible ,
accidents that could be created are identified no accidents of a different !
type can be created. Because no malfunctions are identified no l malfunctions of a different type can be created. Because no acceptance limits are identified that could be affected, the margin of safety is not affected
Attachment II to ET 99-0003 Page 135 of 217 Safety Evaluation: 59 98-0092 Revision 0 Oripper Modification Configuration Change Package (CCP) 07702 modifies a gripper disengage fixture so that it will welded to the liner plate of the floor of the refueling pool. The gripper disengage fixture would be utilized in the event that thc refueling machine gripper became engaged without being latched onto a fuel assembly. In this event, the gripper would be mechanically locked in the engaged position. Lowering the gripper to the gripper disengage fixture would remove the weight of the gripper from the mast and release the mechanical lock, allowing the gripper to be cycled to the disengaged position.
The addition of the gripper disengage fixture to the refueling pool floor, as well as its intended use, will have no detrimental effects on interfacing equipment (refueling pool and refueling machine).
USAR Section 15.7.4 discusses the fuel handling accident inside the reactor building. Several feet of vertical clearance would exist between the gripper disengage fixture and a fuel assembly latched by the refueling machine and in the full up position; thus there would be no potential of impact. The analysis assumec that all the rods in the dropped assembly fail; thus the shape of the fixture (and the number of rods which could potentially fail due to a fuel assembly falling on it) does not have any impact on analysis. The location and height of the fixture are such that at least 23 feet of water would exist above a dropped assembly laying on the fixture, as assumed in analysis. There are no other design basis accidents potentially impacted by the proposed modification.
The concrete structure for the refueling pool is designed to maintain leaktight integrity to prevent the loss of cooling water (reference USAR section 9.1.2.2). The concrete is designed to support compressive forces greater than the weight of the gripper and inner mast (fixture weight negligible); thus the leaktight integrity of the refueling pool will not be challenged.
The purpose of the stainless steel lining of the refueling pool is to provide an easily decontaminable surface and to provide a construction form for the concrete pour. The joint welds of the liner plate are provided with a leakchase system for initial testing and subsequent monitoring of weld integrity. Following installation and testing, a breach of the liner plate (which could result in any significant loss of water through the leakchase system) is not considered credible.
(Reference USAR section 9.1.2.2). The gripper disengage fixture will not be placed directly over a leak chase, thus will have no potential of affecting this system. The gripper disengage plate does not provide any surfaces for the gripper to latch onto, thus there is no potential of providing an upward force on the liner.
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Attachment II to ET 99-0003 Page 136 of 217 1
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The gripper disengage fixture is a passive component utilized to unlatch the gripper of the refueling machine should it become locked in the t
' latched position without being engaged in the upper nozzle of a fuel !
assembly. When lowered onto the fixture, the gripper can be cycled in !
open water between the unlatched and latched. positions. Implementation of this modification'and use of the fixture under. appropriate administrative
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- controls will' not result in any impact to the gripper. !
Because no design basis accidents are affected the probability of. {
occurrence of an accident is not affected. Because no design basis accidents are affected, the consequences of accidents are not atfected.
Because no malfunctions are identified the probability of occurrence of a malfunction is not affected. Because no malfunctions are identified the l consequences of a malfunction are not affected. Because no credible
)
accidents that could be created are identified no accidents of a different ;
type can be created. Because no malfunctions are identified no malfunctions of a different type can be created. Because no acceptance limits'are identified that could be affected, the margin of safety is not affected.
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Attachment II to ET 99-0003 ;
Page 137 of 217 4
Safety Evaluation: 59 98-0094 Revision: 0 Reduced Essential Service Water Flow to Containment Cooling Water System
, Configuration Change Package (CCP) 07866 revises Drawing M-11EF01, System
, Flow Diagram Essential Service Water, and Drawing M-11GN01, System Flow
, Diagram Containment Cooling Water System, to reflect the reduced Essential ;
j Service Water (ESW) flow to the Containment Air Coolers (GN) in the post-LOCA mode of operation. Plant Modification Request (PMR) 03478 originally
- . reduced the ESW flow to the Containment Air Coolers during the post-LOCA mode of operation from 4,000 gpm/ train (2,000 gpm/ cooler) to 2,000 gpm/ train (1,000 gpm/ cooler) .
In conjunction with PMR 03478, Wolf Creek Technical Specifications (TS) were revised by Amendment 50 to the Operating License to require a minimum of.2,000 gpm/ cooler group in the post accident mode of operation. The !
surveillance requirement is contained in TS 4.6.2.3b. Surveillance ,
procedures STS EF-925A and STS EF-925B verify a minimum ESW flow of 2,000 gpm/ cooler group to the Containment Air Coolers following a Safety Injection test signal.
PMR 04478 removed ESW valves EFHV0047 and EFHV0048 and associated piping (bypass line around valves EFHV 0049 and EFHV0050) to the Containment Air Coolers and installed flow restricting orifices EFOOOO5 and EF00006 to minimize throttling of valves EFHV0049 and EFHV0050 to achieve the required flow rate through the Containment Air Coolers. With the flow restricting orifices installed, the 4,000 gpm/ train flow rate is no longer achievable. The flow orifice installation is shown on ESW P&ID M-12EF02.
- These changes make M-11EF01 and M-11GN01 consistent with existing documentation and the installation in the field. Previously the reduced ;
ESW flow information was contained in the notes section of the drawings but not in the flow point tables. There is no field work associated with this CCP.
Updated Safety Analysis Report ('USAR) 2able 9.2-3 will also be revised as a result of this CCP to reflect the reduced ESW flow to the Containment Air Coolers during post-LOCA operation. A note in the USAR table contains ,
the reduced flow information but the actual table retains the 4,000 gpm/ train flow rate.
These documentation changes will make the ESW and GN system flow diagrams and USAR Table 9.2-3, ESW System Flow Requirements Post-LOCA Operation, consistent with the plant design and operation since the implementation of PMR 03478. The documentation changes do not affect any system, structure or component (SSC) nor do they change the performance of activities that ,
are important to the safe and reliable operation of Wolf Creek Generating l Station. !
l USAR Table 9.2-3 is the only place in the USAR where the 4,000 gpm/ train ;
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L Attachment II to ET 99-0003 Page 138 of 217 ESW flow to the Containment Air Coolets in the post-LOCA mode is retained.
The remainder of the USAR is consistent with the changes previously made with PMRs 03478 and 04478. The change to USAR Table 9.2-3 vill make it consistent with the remainder of the USAR as well as plant design and '
operating procedures. No other USAR descriptions or conclusions will change or be made untrue as a result of this change.
I when PMR 03478 was implemented it was determined that 2,000 gpm/ train (1,000 gpm/ cooler) ESW flow to the Containment Air Coolers in the post-LOCA mode of operation was adequate removal of heat from the containment.
The changes proposed by CCP 07866 are consistent with changes made to the plant by PMR 03478 and PMR 04478.
Because no design basis accidents are affected the probability of occurrence of an accident is not affected. Because no design basis accidents are af fected, the consequences c f accidents are not af fected.
Because no malfunctions are affected the probability of occurrence of a t malfunction is not affected. Because no r21 functions are identified the consequences of a malfunction Are not affected. Because no credible accidents that could be creat<4d are ider.tified no accidents of a different type can be created. Because no malfu. actions are identified no l ma'. functions of a different type can be created. Because no acceptance -
l limits are identified that could be affected, the margin of safety is not l
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Page 139 of 217 '
i Safety Evaluations- 59 98-0095 Revision 0 i Updated Safety Analysis Report Changes to Personnel Titles l
This revision to the Updated Safety Analysis Report (USAR) changes the I title of the " Superintendent of Operations" to " Superintendent :
Operations". In addition, the title of " Equipment / Auxiliary Operator" was !
changed to the title of
- Nuclear Station Operator." The changing of these l titles does not affect the operation of the plant.
i This revision adds information to identify that Wolf Creek Nuclear j Operation Corporation (WCNOC) has suspended the operator bachelor degree j program A statement was added to identify that WCNOC has either a Senior !
Reactor Operator or Shift Engineer who meets the on-shift technical :
advisor training requirements on each shift. f i
This change deletes USAR Section 18.1.9.2, bullet "a" in its entirety. In l
bullet ab" of this section, the phrase "a plant status board" was deleted j to allow plant administrative procedures to govern the tools that will be !
used to convey equipment condition. !
This change to the USAR is implemented to ensure accuracy. No other USAR !
descriptions or conclusions will change or be untrue because of this j change. These changes do not result in any test or experiments not :
described in the USAR which could adversely affect the adequacy of l' systems, structures, or components (SSCs) to prevent accident or mitigate the consequences of an accident.~
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There are no design basis accidents identified because these USAR changes {
affect the description of administrative organizations and controls that do not reduce the level of qualification of WCNOC personnel. j l
These USAR changes only affect the description of administrative {
organizations and controls and there is no affect on any SSC. Therefore, no credible malfunctions of equipment important to safety are identified ;
and no acceptance limits are identified that could be affected.
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Page 140 of 217
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Safety Evaluation 59 98-0096 Revision: 0 Use of sodium Hypochlorite and Sodium Bromide For Biocide Treatment This modification provides for the use of sodium hypochlorite and sodium ;
bromide for biocide treatment of the service water and circulating water l
- systems. In addition, this modification discontinues using
' l BromoChloroDiMethylHydantoin (BCDMH) for biocide treatment. l i
i Liquid sodium hypochlorite (bleach) is proposed to be stored in a 7,900
- ga]lon capacity tank. Liquid sodium bromide is proposed to be stored in an 8,000 gallon tank. The two storage tanks, along with associated metering pumps and controls, are to be located within the chlorine storage ,
structure. The chlorine storage structure, Z052, is a non safety related, !
non seismic structure. Abandoned chlorine header piping, located in the !
chlorine structure, is to be removed. All electrical power for the new equipment is supplied from the same distribution panels which were used for the BCDMH equipment. '
! An existing three inch service water branch line runs from the Circulating Water Screen House to the chlorine storage structure and returns to either j the service vater or circulating water system. The treatment chemical is :
added to tlis service water flow at the chlorine storage structure. This l service water flow path provides the carrier service for either the BCDMH treatment or the sodium hypochlorite and sodium bromide treatment. i 1
Updated Safety Analysis Report (USAR) Section 2.2.1.2, describes " storage .
vessels, which are not part of the standard plant." This section of the 1 I USAR does not currently describe storage tanks for sodium hypochlorite and sodium bromide. Adding a description of the two new chemicals fulfills the need of thic USAR section. Chapter 2 of the USAR discusses onsite storage vessels for chemicals which are not part of the standard plant.
Evaluation of potential accidents from onsite or offsite hazards is addressed. l Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," addresses control room habitability during a postulatad release of chlorine gas. Currently, chlorine is stored in six - 150 pound vessels in the chlorinator room in j the shop building, and it is used to treat the influent to the makeup demineralizer system. Lesser amounts of chlorine (1ers'than 20 pounds) i may be used routinely for laboratory work, Using these amounts of
- chlorine on site conforms to the recommendations of Regulatory Guide 1.95, and control room habitability is not affected. USAR Table 6.4-2 summarizes the Wolf Creek position. The installation of sodium
, hypochlorite does not change the existing chlorine evaluation because sodium hypochlorite is not volatile like chlorine, i
l Regulatory Guide 1.78, " Assumptions for Evaluating the Habitability of a 4
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Nuclear Power Plant Control Room During a Postulated Hazardous Chemical {
Release," addresses control room habitability during a postulated release i of potentially hazardous chemicals other than chlorine. It describes two types of. industrial chemical accidents to consider for control room f i
habitability: these are maximum concentration chemical accidents and {
maximum concentration - duration chemical accidents. l t
A maximum concentration accident is proposed to result from an l instantaneous release of the total contents of a storage tank. In the event of a tank rupture, the spilled contents will be contained within a concrete berm. Because sodium hypochlorite is a liquid solution, not a pressurized or liquefied gas, the spilled contents will not become air j borne. The vapor pressure of sodium hypochlorite is quite low, similar to ;
that of water. A rapid boiloff of the spilled liquid will not occur. The !
storage tanks are continuously vented preventing any accumulation of gases within the tank. Oxygen is a major component of the natural decomposition of sodium hypochlorite. A maximum concentration accident is not possible [
with sodium hypochlorite because the low rate of flashing or boiloff of )
toxic vapors, during normal conditions or after a tank rupture, will not create a hazardous environment beyond the immediate vicinity. :
i In a maximum concentration - duration accident, a continuous release of i hazardous chemical from the largest safety relief valve on the source is ;
proposed by the regulatory guide. A pressurized release through an open ;
safety valve is not possible because the sodium hypochlorite tank is l continuously vented. The tank is not a pressurized storage vessel. A j maximum concentration - duration accident is not possible with sodium r hypochlorite.
l In addition, the chlorine storage structure, where the chemicais are to be 4 stored and used, is far from the control room. The chlorine storage j structure, 2052, is located near the circulating water screen house, the ,
chlorine storage structure is approximately 1900 feet from the control j room. This distance enables the wide dispersion of vapors before they could drift to the control room. !
I Sodium hypochlorits does not present a control room habitability hazard. (
Sodium bromide is a stable solution with a pH ~ 7.5, and exhibits no !
specific hazard. Sodium bromide does not present a control room l habitability hazard.
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No credible accidents are possibly created. The nearest type of credible accident which may occur would be one where a storage tank ruptures and i causes control room habitability problems. As discussed above, the ,
storage of sodium hypochlorite at the chlorine storage structure does not !
create a hazard for the control room. The control room habitability i systems are unaffected. There is no safety related equipment in the I vicinity of the chlorine structure or the circulating water screen house I
-to be harmed by the chemical biocide. ]
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Attachment II to ET 99-0003 Page 142 of 217 l
No credible malfunctions are affected by this modification. There is no l system interface of the biocide treatment system beyond the service water ;
and circulating water systems. The concentrated sodium hypochlorite solution only exists within the chlorine storage structure. It is present ,
in the treated water systems in trace amounts, less than 1 part per l million of the residual oxidant at the Essential Service Water system outfall. The chemical properties of the treated water are essentially l unchanged. 1 The planned continuous, daily treatment of the service water and ,
circulating water systems will keep the systems clear of biological organisms. The frequent treatments will prevent blockage events which could possibly occur if treatments were not performed regularly. No basis is identified which could be affected. There is no acceptance limit in the technical specification concerning chlorine.
The original plant design used 1 ton chlorine containers for biocide treatment, at the same location now to be used for the sodium hypochlorite. When the chlorine cylinders were removed, the chlorine detectors for the control room were disconnected from the system and spared in place. From USAR: "If single chlorine containers greater than 150 lbs are brought back on site, the chlorine detectors shall be made operable and provide alarm indication at the control room and automatically isolate the control room in seven seconds for chlorine concentrations of 5 ppm or greater in control building supply air intake duct. Chlorine accident analysis will be performed and the USAR and Technical specification revised to reflect the use of greater than 150 lbs ,
containers of chlorine."
Chlorine is not being reintroduced back on site. Chlorine detectors at the control room are not required as a result of installing a sodium hypochlorite and sodium bromide treatment system. The requirement for chlorine detectors at the control room is unaffected.
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Attachment II to ET 99-0003 Page 143 of 217 l
i Safety Evaluation: 59 98-0097 Revision 0 i Revision 30 to the Security Plan Revision 30 to-the Security Plan incorporates.the following changes: l updates.the Coffey County Sheriffs manning for response to Wolf Creek Nuclear Operating Corporation (WCNOC) requests for. assistance because i upgrades have been made to sheriffs department. This change affects the Security Contingency Plan and changes the test requirements for door 22021 !
to allow testing only when the barrier is removed. This revision changes the retention period for security procedure changes and approvals from three years to five years. This revision corrects errors found in ;
Security Plan Figures 3.1-1, 3.1-4, 3.2-1, 5.1-3, and 5.1-5 to reflect I current security configurations.
i This revision changes the requirement for approval of the current vital i sector authorization list from every 31 days to at least quarterly. This l revision changes vehicle escort requirements to allow company owned or leased vehicles driven by an individual with unescorted access to enter i the protected area without security escort. This revision changes i Security Operations Supervisor title to Security Shift Lieutenant operation. This change does not affect equipment important to safety and ;
does not reduce the protection for the plant resources.
i These changes have no affect on any systems structures or components. ,
Because no design basis accidents are impacted, the probability of occurrence of an accident, and the consequences of any accident are not ;
affected. Because no malfunctions are impacted, the probability of occurrence and the consequences of a malfunction are not affected. i Because there are no credible accidents that could be created, no i accidents of a different type can be created. Because no malfunctions are identified, no malfunctions of a different type can be created. Because f no acceptance limits affected, the margin of safety is not affected. !
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Attachment II to ET 99-0003 5
-l Page 144 of 217 i Safety Evalurvion: 59 98-0098 Revision 0 !
Zero Filtration System The Diversified Technology's 2ERO Filtration System is being installed under Temporary Modification TMO 98-018-HB to evaluate the effectiveness . ;
and benefit to Wolf Creek Generating Station (WCGS). The ZERO Filtration !
System will be connected in series upstream of the existing vendor ;
supplied Demineralizer Skid. l l
It has been identified that certain insoluble materials in the liquid radwaste system can not be removed by the Demineralizer Skid and must be ;
physically removed using filtration. The current filtration provided in j the Liquid Radwaste System has the ability to remove particulate material of 9.45 micron and larger.
l The Zero Filtration System consists of three major skids:
- 1) The Tubular Ultrafiltration (TUF) Skid removes all particulate material larger than 0.05 microns in size.
- 3) The Mobile Drum Dryer (DD) Skid drums and dewaters the concentrated contamination removed by the TUF and SRO skids for disposal as solid l radwaste.
l The equipment will be physically located on the 2031'-6" elevation of the Radwaste Building. Evaluations have been completed by Engineering [
determining the unit's acceptability for use at WCNOC. Regulatory Guide l 1.143, " Design Guidance for Radioactive Waste Management Systems, !
Structures and Components Installed in Light-Water Nuclear Power Plants" is being followed for the equipment and installation of the ZERO ;
Filtration System. Evaluations included floor loading, hypothesized .
shielding requirements and electrical power requirements. f b
Hoses will be routed through existing pipe penetrations to connect the ZERO Filtration System to the existing components of the Demineralizer j Skid for further processing. The processed water can then be either '
released to the cooling lake or reused internally at WCGS. ;
i The filter cartridge in filter housing FHB06 will be removed or modified for the duration of this TMO. The ZERO Filtration System will provide better filtration than any filter that we place in FHB06. In addition, ;
the-filter removal or modification will eliminate the cost of labor, personnel radiation dosage and radwaste involved with the change out of ;
.the. filters. !
Based on the above evaluation, this temporary modification will not increase the probability of occurrence or the consequences of an accident :
or malfunction of equipment important to safety previously evaluated in l
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l l Attachment II to ET 99-0003 Page 145 of 217 l
l the safety analysis report. This modification does not create a possibility for an accident or malfunction of a different type than any l evaluated previously in the safety analysis report. The margin of safety, ;
as defined in technical specifications, is not reduced by this !
modification. Therefore, this modification does not involve any unreviewed safety question, l
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Safety Evaluation: 59 98-0099 Revision: 0 I i Revision of Essential Service Water Warning Line Procedures t Procedures STN EF-020A, "ESW Train A Warming.Line Verification," and STN EF-0203, "ESW Train B Warming Line Verification," measure the Essential !
Service Water (ESW) system warming line operating pressures and flow rates. The acceptance criteria for the warming line flow rate has been ;
5000 gpm i 400 gpm. A corresponding system flow rate is not specified. .
Current Updated Safety Analysis Report (USAR) Section 9.2.1.2.2.3, System ,
operation, states that the warming lines divert a minimum of 4500 gpm of
- the return flow back to the ESW pump intake bays during cold weather. A corresponding system flow rate is not specified. ;
i Calculation EF-M-030, Revision 0 finds a minimum warming line flow rate of 4,415 gpm is needed to warm an ESW pump suction flow rate of 15,010 i' 3
gpm. 4,415 gpm was conservatively rounded to 4,500 gpm for conservatism
! (the USAR value). The conclusion statement says that the warming line
- flow rate should be established at a nominal 5,000 gpm (the procedure !
value). l Calculation EF-M-036, Revision 0 finds that a maximum lake temperature of 58 degre+ Fahrenheit (F) could briefly result in an ESW pump suction .
temperat. of 95 degrees F assuming a maximum warming line flow rate of l 5,500 gpm ui an ESW pump flow rate of 13,576 gpm. I Calculation Problem: Revision 0 of both of the above calculations, ,
calculated a result for only one ESWS operating point. No projections I were made for other flow rates. Resolution: Calculation EF-M-036 was l l
revised. It was revised to determine the maximum and minimum warming line flow rates for a range of ESW system flow rates, considering the warming line flow rate will vary with the ESW system flow rates and considering the transit time for water to flow througS .he buried ESW supply and return piping and considering different late temperatures. Calculation EF-M-030 was also revised. The original design points were used as the basis for the revisions.
Procedure Problem: The surveillance procedures state an acceptance criteria for the warming line flow rate, and they do not reference a corresponding ESW system flow rate. Resolution: The addition of a new figure to each of the procedures, showing the maximum and minimum warming line flow rates, will replace the original acceptance criteria. The maximum and minimum values may vary from that shown as determined by originating calculations. The expected effects of changing the flow rate acceptance criteria in the surveillance procedures will be a clear and defined set of bounding values for warming line flow rates and corresponding ESW system flow rates. l l
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Attachment II to ET 99-0003 Page 147 of 217 Loss of normal feedwater, loss of non-emergency AC power and feedwater line break accidents assume the availability of the ESW system to provide emergency makeup water to the auxiliary feedwater system. The ESW system provides heat removal for safety related equipment (emergency diesels, Component Cooling Water heat exchangers, pump room coolers, containment coolers and Control Room and Class 1E refrigerant condensers). '
The proposed procedure changes do not affect the ability of the ESW system to respond to an activation signal. The surveillance procedures do not change any setpoints or system hardware. The new set of flow rate acceptance criteria defines the flow rate limits over a range of flow rates, for the warming line and for the system The surveillance procedures measure the ESW warming line flow rate. The original acceptance criteria of 5,000 gpm i 400 gpm is satisfactory for ESW system flow rates at and near the ESW system total flow rate as designed. But, the ESW system total flow rates vary from design because some excess water flows through each component to maintain a minimum flow, and because the flow through each containment cooler train may range from 2,000 gpm minimum to 4,000 gpm, and because the emergency makeup flow paths are normally not flowing, and because flow control valves regulate water flow through the refrigeration unit condensers.
The revised acceptance criteria provide acceptance values over a range of ESW system total flow rates as normally operated. The new acceptance criteria better defines the limits of flow for conditions found over a range of ESW system total f3ow rates. System flow rates are not changed by the procedure, but only monitored.
The ESW system flow rates and the warming line flow rates are parameters that are set to provide cooling to safety related components. The ESW system pumps valves and orifices which control the system flow rates are not changed, adjusted, or affected by the warming line flow verification procedures, STN EF-020A and STN EF-020B. Using an expanded range of ,
acceptance criteria enables a greater degree of confidence for acceptable operating parameters.
Technical Specification 3/4.7.4 presents the ESW system requirements.
There are no flow rates discussed and no acceptance limits for the warming line in the related basis. Procedures STN EF-020A and STN EF-020B exist to periodically show that there is satisfactory flow through the warming lines. The only reference to warming line flow rate is found in the identified USAR section 9.2.1.2.
Because nc design basis accidents are affected the probability of occurrence of an accident is not affected. Because no design basis j accidents are affected, the consequences of accidents are not affected. i Because no malfunctions are affected the probability of occurrence of a malfunction is not affected. Because no malfunctions are identified the consequences of a malfunction are not affected. Because no credible ;
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Attachment II to ET 99-0003 Page 148 of 217 accidents that could be created are identified no accidents of a different type can be created. Because no malfunctions are identified no malfunctions of a different type can be created. Because no acceptance limits are identified that could be affected, the margin of safety is not affected.
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f Safety Evaluation: 59 98-0100 Revision 0 i
Acceptance of Degraded Neld Condition
, f During the performance of maintenance activities under WR 98-129416-001, l the inlet' spool' flange for the SGK05B condenser was discovered to have l
approximately 3 inches of the length of the interior fillet weld eroded i from an original 1/4 inch to an as-found (field measured) 3/16 inch. !
t ASME Section III, Subsection ND, Aappendix XI, indicates that the minimum l required fillet weld leg size is 1/4 inch (0.250 inch) for a slip-on type i flange. Based on the as-found condition, the minimum weld size code ;
requirements are not met. ;
i Further engineering evaluation by stress analysis of the as-found welded connection configuration, based on the resultant forces and moments acting at the welded connection under all postulated design loading conditions j (reference pipe stress calculation P-192), determined that a minimum weld ;
size of 0.126 inch is required to maintain the structural integrity of the ;
connection. j i*
since the actual weld size (0.1875 inch) envelopes the minimum required weld size (0.126 inch), it has been concluded that there is enough margin of safety at the welded connection, to maintain the pressure boundary under all postulated loading ccnditions, including a seismic event.
The proposed conditional.use-as-is evaluation does not adversely affect any system, component, or procedures required to mitigate the consequences of an accident previously evaluated in the USAR.
Since the as-found degraded condition was found by stress analysis to be !
structurally acceptable to perform its original design intent, it was I concluded that a new credible failure mode is not introduced. Therefore, l there are no malfunctions of equipment important to safety identified, and i all system functions will continue to be performed as designed. j
. Since the Stress analysis determined that the as-found weld size of 0.126 inch is acceptable, based on the resultant forces and moments acting at ;
the welded connection under all postulated design loading conditions, and l' determined that the equipment will continue to perform its intended design function, no accidents referenced in Updated Safety Analysis Report (USAR) will ta impacted.
Since the Stress analysis determined that the as-found weld size of 0.126 inch is acceptable, based on the resultant forces and moments acting at ,
the welded connection under all postulated design loading conditions, and l determined that the equipment will continue to perform its intended design l function, no credible accidents could be created. ?
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i Attachment II to ET 99-0003 Page 150 of 217 Since the Stress analysis determined that the as-found weld size of 0.126 inch is acceptable, based on the resultant forces and moments acting at the welded connection under all postulated design loading conditions, and determined that the equipment will continue to perfcrm its intended design function, no credible malfunctions of equipment important to safety are identified.
Since, the as-found degraded condition was found by stress analysis to be structurally acceptable to perform its original design intent, no acceptance limits which could affect the basis for any technical specification are affected.
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Attachment II to ET 99-0003 i
Page 151 of 217 i
Safety Evaluation: 59 98-0101 Revision 0
{
Revision to Design Drawing to Reflect As-Built Configuration f Configuration Change Package (CCP) 07885 revises the Chemical and Volume !
Control System P&ID M-12BG03, to correct two errors when it was taken from Revision 6 to Revision 7. The P&ID is also Updated Safety Analysis Report (USAR) Figure 9.3-8-03. The two changes being made to M-12BG03 are: !
a.) The line. designator for the Volume control Tank (VCT) outlet drain to the RHUT (between line 020-HCB-4" and valve V245) is being changed from 316-HCB-1" to 318-HCB-1". This change is supported by MS-01, Revision 41, f Piping Class Summary for WCGS, which gives the sequence number for this '
line as 318. In addition, Piping Isometric M-13BG01, Revision 4, CVCS - t Minimum Charging Flow Auxiliary Building, shows the drain line as 318-HCB-1". These two docuna nts reflect the as-built condition of the plant.
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b.) The piping classification change (HCB to HCD) for this drain line is being moved from the upstream side of valve V245 to the downstream side of ;
valve V245. This change is supported by M-13BG01 which shows the piping l classification change on the downstream side of valve V245. M-13BG01 i reflects the as-built condition of the plant. These drawing changes j correct inadvertent errors that entered M-12BG03 during an earlier '
revision and restore the drawing to reflect the as-built status of the plant.
No physical changes are made to any system, structure, or component. This CCP doas not change the performance of activities that are important to the safe and reliable operation of Wolf Creek Generating Station.
.Therefore, this modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.
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Attachment II to ET 99-0003 f
- Page 152 of 217 f l
l Safety Evaluation: 59 98-0102 Revision 0 '
Revision to Drawing to Reflect As-built Condition ;
Configuration Change Package (CCP) 07886 revises Feedwater Heater Extraction Drains and Vents P&ID M-12AF01 to correct an error that entered l the drawing when it was taken from Revision 7 to Revision 8. The P&ID is also Figure 10.4-6-03 of the Updated Safety Analysis Report. This revision to M-12AF01 changes the line designator for a portion of the 3rd -
Stage Extraction To Heater 7B line from 635-QAD-12 inches to 635-GAD-12 inches. This change is supported by MS-01, Revision 41, Piping Class !
Summary for WCGS, which shows AF line sequence 635 as pipe class GAD and j service as 3RD Stage Extraction to Heater 7B. Piping Isometric M-13AC08, !
l Revision 8, "3RD. Stage Extr. To HTR. 7A & 7B Turbine Building," also !
shows the line designator as 635-GAD-12 inches. MS-01 and M-13AC08 reflect the as-built condition of the plant.
This drawing change corrects an inadvertent error that entered M-12AF01 during an earlier revision and restores the drawing to reflect the as-built status of the plant. No physical changes are~made to any SSC. This ,
! CCP does not change the performance of activities that are important to f the safe and reliable operation of Wolf Creek Generating Station. l I
This modification will not increase the probability of occurrence or the )
consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this modification. Therefore, this modification does not involve any unreviewed safety question.
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Attachment II to ET 99-0003 Page 153 of 217 Safety Evaluation: 59 98-0103 Revision 0 Elimination of Snubbers Snubbers were eliminated / reduced from Main Steam (MS) and Main Feed Water (FW) lines, inside the Containment building, by Plant Modification Request (PMR) 05785. The field changes to this PMR were implemented in RF VII.
Also, Steam Generator (SG) hydraulic snubbers were eliminated through a re-analysis of the main loop using time history seismic inputs and Leak Before Break (LBB) technology, as allowed by the revised GDC-4 implementing guidance document, Branch Technical Position MEB 3-1, Revision 2. This project was implemented in RF VI under Plant Modification Requests (PMRs) 04579 and 04373. As a part of implementation, two SG hydraulic snubbers per loop were converted to compression bumpers, and the remaining two per loop were abandoned in place. Also, shims of cross over leg whip restraints and certain other whip restraints were removed.
It was identified that plant modifications 04579, 04373, and 5785 had used the Branch Technical Position MEB 3-1, Revision 2 criteria for High Energy i Pipe Break analysis, to eliminate the Arbitrary Intermediate Breaks from l MS and FW lines. Branch Technical Position MEB 3-1," Postulated Rupture Locations in Fluid System piping inside and outside Containment", provides ;
acceptance criteria for detining postulated pipe rupture locations inside i and outside containment and methodology for pipe whip restraint dynamic analysis. The original design basis for Wolf Creek Generating Station is Branch Technical Position MEB3-1, Revision O.
The NRC issued Generic letter (GL) 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements," (June 19, 1987) along with Branch Technical Position MEB 3-1, Revision 2. The objective of Generic letter 87-11 was to provide Relaxation in Arbitrary Intermediate Pipe Rupture Requirements, and to allow elimination of Arbitrary Intermediate Break (AIB) requirement for High Energy lines. AIBs are selected based on the highest stress values, even though stress values are less than the allowable code limits. AIBs also require postulation of pipe breaks, jet img:ngement force evaluation and addition of whip restraints to protect nes.-by safety related pipes and structures.
Besides the relaxation, Revision 2 of Branch Technical Position MEB3-1, introduces some minor changes and updates ASME stress limits, to be consistent with current or committed ASME Codes.
While implementing changes for PMRs 4579, 5785 and 4373, re-analysis of the MS and FW systems and the Reactor Coolant Loops (RCLs) were performed.
Leak Before Break (LBB) technology was re-demonstrated using revised GDC-4 implementing guidance methodology for the RCLs. The analysis results met the requirements of the ASME code section III, revised GDC-4 implementing guidance and GL 87-11. Therefore, accidents identified in chapters 2, 3, 6, 9 or 15 of the USAR are not impacted by this change.
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Attachment II to ET 99-0003 Page 154 of 217 While implementing changes for PMRs 4579, 5785 and 4373, re-analysis of the MS and FW systems and the RCLs was performed. Leak Before Break (LBB) technology was re-demonstrated using revised GDC-4 implementing guidance methodology for the RCLs. The analysis results met the requirements of the ASME Code Section III, revised GDC-4 implementing guidance and GL 87-11.
Therefore, the proposed design changes will not create any credible accidents.
Re-analysis of MS and FW systems and RCLs comply with industry criteria, j Branch Technical Position MEB3-1, and GL 87-11. This provides a basis for safety requirements of components designed in accordance with the l requirements of Section III of the ASME Code. Therefore, the proposed i change will not create any credible malfunction to equipment associated with the MS and FW systems and RCLs.
The methodology of Branch Technical Position MEB3-1, Revision 2 is j approved for use by the NRC in GL 87-11, DT June 19, 1987 The following
- statement in the GL 87-11 allows the elimination of AIBs in the high energy lines. " Licensees of operating plants desiring to eliminate previously required effects from arbitrary intermediate pipe ruptures may do so without prior approval unless such changes conflict with the license !
or technical specification". There is no impact on the technical specification systems, structures, or components due to these changes.
The Leak Before Break (LBB) technology was re-demonstrated using revised GDC-4 implementing guidance methodology. The results of this analysis is documented in WCAP 10691. Therefore, acceptance limits as identified in the USAR or other documents are not affected by this change.
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l Attachment II to ET 99-0003 f Page 155 of 217 i
Safety Evaluation: 59 98-0104 Revision: 0 Piping Replacement Because of Flow Accelerated Corrision >
l Configuration Change Package (CCP) 07755 is issued to provide guidelines j for the modification / replacement of Main Turbine System (AC) and Feedwater !
Heater Extraction-Drains and Vent System (AF) piping lines AC-072-GBD-20, AC-073-GBD-20, AC-006-GBD-20 and AF-051-GBD-20, to mitigate abnormal pipe-wall thinnir.g due to Flow Accelerated Corrosion (FAC). These line are being replace with a low alloy steel (2 1/4 Cr - 1 Moly) which improves l piping resistance to FAC. .
l The proposed pipe replacement did not change the cross sectional "
properties (section modulus, moment of inertia) or the geometric j configuration of the piping. The mechanical properties such as tensile strength and code allowable stresses will remain unchanged. The lower !
yield strength and higher Young's Modulus is judged to have insignificant impact on the original analysis. Therefore, the change does not adversely }
affect the existing safety margins or structural integrity of the affected l piping system. The piping stresses will remain within code allowables.
The proposed replacement does not adversely affect any system, component or procedures required to mitigate the consequences of an accident i previously evaluated in the Updated Safety Analysis Report. The proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings).
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain and mechanical creep are credible failure modes for which the proposed piping replacement has been evaluated, through a critical characteristics comparison to the existent piping system design.
Based on the evaluation, it was concluded that a new credible failure mode is not introduced. Therefore, there are no malfunctions of equipment important to safety identified.
The proposed change will restore a degraded section of the affected piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents are affected.
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents could be created.
t Attachment II to ET 99-0003 Page 156 of 217 ;
The proposed pipe replacement did not change the croso sectional properties (section modulus, moment of inertia), or the mechanical properties (tensile strength, and/or code allowable stresses), or the geometric configuration, no new malfunction of equipment important to safety is identified.
[
Since the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits which could affect the basis for any Technical specification are affected.
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Attachment II to ET 99-0003 l I
Page 1$7 of 217
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Safety Evaluations- 59 98-0105 Revision 0 l Piping Replacement Because of Flow Accelerated Corrision !
Configuration Change Package (CCP) 07756'is issued to provide guidelines for the modification / replacement of Feedwater Heater Extraction-Drains and l Vent System (AF) piping lines AF-467-GBD-16 and AF-038-GBD-16, to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC) . This
. piping is being replaced with a low alloy steel (2 1/4 cr - 1 Moly) which improves piping resistance to FAC.
The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertia), or the geometric configuration of the piping. The mechanical properties such as tensile strength and code allowable stresses will remain unchanged. The lower yield strength and higher Young's Modulus is judged to have insignificant impact on the original analysis. Therefore, the change does not adversely ,
affect the existing safety margins or structural integrity of the affected !
piping system. The piping stresses will' remain acceptable within code allowables.
The proposed replacement does not adversely affect any system, component, i or procedures required to mitigate the censequences of an accident ;
previously evaluated in the Updated Safety Analysis Report. The proposed :
change will restore a degraded section of the affected piping system, to !
its original design configuration (piping geometry, cross section, support ,
location, fittings). l l
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical !
properties, excess strain, mechanical creep etc., are credible failure modes for which the proposed piping replacement has been evaluated, j through a critical characteristics comparison to the existent piping system design. Based on the evaluation, it was concluded that a new credible failure mode is not introduced. Therefore, there are no ;
malfunctions of equipment important to safety identified. ;
The proposed change will restore a degraded section of the affected piping system, to perform its original design intent. The proposed replacement j does not involve or affect any safety related system or component. All system functions will continue to be performed as designed. (
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents are affected.
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, t cross section, support location, fittings), no new accidents could be {
created. !
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Attachment II to ET 99-0003 Page 158 of 217 The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertia), or the mechanical properties (tensile strength, and/or code allowable stresses), or the -
geometric configuration, no new malfunction of equipment important to safety is identified.
Since the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits which could affect the basis for any Technical specification are affected.
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l Attachment II to ET 99-0003 Page 159 of 217 l
Safety Evaluation: 59 98-0106 Revision 0 Piping Replacement Because of Flow Accelerated Corrision Configuration Change Package CCP 07758 is issued to provide guidelines for -
the modification / replacement of a degraded section of Feedwater Heater
- Extraction-Drains and Vent System (AF) piping line AF-451-DBD-4, to.
mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion ;
(FAC). This piping is being replaced with a low alloy steel (2 1/4 Cr - 1 ;
Moly) which improves piping resistance to FAC. !
l
'The proposed pipe replacement did not change the cross sectional l properties (section modulus, moment of inertia), or the geometric configuration of the piping. The mechanical properties such as tensile ;
strength and code allowable stresses will remain unchanged. The lower '
yield strength and higher Young's Modulus is judged to have insignificant !
impact on the original analysis. Therefore, the change does not adversely l affect the existing safety margins or structural integrity of the affected !
piping system. The piping stresses will remain acceptable within code allowables.
The proposed replacement does not adversely affect any system, component, or procedures required to mitigate the consequences of an accident [
previously evaluated in the Updated Safety Analysis Report. The proposed I change will restore a degraded section of the affected piping system, to ;
its original design configuration.(piping geometry, cross section, support j location, fittingt.). :
i Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure j modes for which the proposed piping replacement has been evaluated, through a critical characteristics comparison to the existent piping system design. Based on the evaluation, it was concluded that a new {
credible failure mode is not introduced. Therefore, there are no j malfunctions of equipment important to safety identified. j l
The proposed change will restore a degraded section of the affected piping .
system, to perform its original design intent. The proposed replacement {
does not involve or affect any safety related system or component. All l system functions will continue to be performed as designed. ;
Since the proposed change will restore a' degraded section of the affected l piping system, to its original design configuration (piping geometry, '
cross section,' support location, fittings), no accidents are affected.
Since the proposed change will restore a degraded section of the affected {
piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no new accidents could be created.
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Attachment II to ET 99-0003 Page 160 of 217 The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertin), or the mechanical properties (tensile strength, and/or code allbwable stresses), or the geometric configuration, no new malfunction of equipment important to safety is identified.
Since the proposed change will restore a degraded section of the affected
-piping system, to perform its original design intent, no acceptance limits which could affect the basis for any Technical specification are affected.
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i Attachment II to ET 99-0003 )
Page 161 of 217 i
Safety Evaluation: 59 98-0107 Revision: 0 ;
Piping Replacement Because of Flow Accelerated Corrision 4
Configuration Change Package 07797 is issued to provide guidelines for the ;
modification / replacement of Condensate System (AD) piping line AD-319-HBD- ,
2,-to mitigate abnormal pipe-wall thinning due to Flow Accelerated l Corrosion (FAC) . This piping is being replaced with a-low alloy steel (2 l 1/4 Cr - 1 Moly)' which improves piping resistance to FAC.
l The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertia), or the geometric configuration of the piping. The mechanical properties such as tensile ,
strength and code allowable stresses will remain unchanged. The lower !
yield strength and higher Young's Modulus is judged to have insignificant !
impact on the original analysis. Therefore, the change does not adversely l affect the existing safety margins or structural integrity of the affected !
piping system. The piping stresses will remain acceptable within code ;
allowables.
The proposed replacement does not adversely affect any system, component, or procedures required to mitigate the consequences of an accident previously evaluated in the U pdated Safety Analysis Report. The proposed ;
change will restore a degraded section of the affected piping system, to i its original design configuration (piping geometry, cross section, support ('
location, fittings).
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical ;
properties, excess strain, mechanical creep etc., are credible failure modes for which the proposed piping replacement has been evaluated, through a critical characteristics comparison to the existent piping syster design. Based on the evaluation, it was concluded that a new ;
credible failure mode is not introduced. Therefore, there are no '
malfunctions of equipment important to safety identified.
The proposed change will restore a degraded section of the affected piping ;
system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All ,
system functions will continue to be performed as designed. [
i Since the proposed change will restore a degraded section of the affected !
piping system, to its original design configuration (piping geometry, (
cross section, support location, fittings), no accidents are affected.
Since the proposed change will restore a degraded section of the affected -
piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no new accidents could be created.
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Attachment II to ET 99-0003 Page 162 of 217 The proposed pipe replacement did not cl.7nge the cross sectional properties (section modulus, moment of inertia), or the mechanical properties (tensile strength, and/or code allowable stresses), or the geometric configuration, no new malfunction of equipment important to safety is identified.
l Since the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits I which could affect the basis for any Technical specification are affected.
i
Attachment II to ET 99-0003 Page 163 of 217 Safety Evaluation: 59 98-0108 Revision O Piping Replacement Pecause of Flow Accelerated Corrision Configuration Chance Package 07804 is issued to provide guidelines for the modification / replacement of Feedwater Heater Extraction-Drains and Vents System (AF) and Condensate System (AD) piping lines AF-271-HBD-4; AF-273-HBD-3; AF-270-HBD-1.5; AD-102-HBD-3; AF-325-HBD-1; AF-328-HBD-3; AF-326-HBD-1.5; AD-111-HBD-3; AF-289-HBD-1; AF-290-HBD-2; AF-292-HBD-3 and AD-105-HBD-3, to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC) . This piping is being replaced with a low alloy steel (2 1/4 Cr. - 1 Moly) which improves piping resistance to FAC (see NUREG-1344 for reference).
The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertia), or the geometric configuration of the piping. The mechanical properties such as tensile strength and code allowable stresses will remain unchanged. The lower yield strength and higher Young's Modulus is judged to have insignificant impact on the original analysis. Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.
The proposed replacement does not adversely affect any system, component, or procedures required to mitigate the consequences of an accident previously evaluated in the Updated Safety Analysis Report. The proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, _ fittings).
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the proposed piping replacement has been evaluated, through a critical' characteristics comparison to the existent piping system design. Based on the evaluation, it was concluded that a new credible failure mode is not introduced. Therefore, there are no malfunctions of equipment important to safety identified.
The proposed change will restore a degraded section of the affected piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents are affected.
Since the proposed change will restore a degraded section of the affected
. . _ . _ - . . . -_._m..
Attachment II to ET 99-0003 ,
Page 164 of 217 piping system,.to its original design configuration (piping geometry, cross section, support location, ' fittings), no new accidents could be l' created. !
l The proposed pipe replacement did not change the cross sectional !
properties (section modulus, moment of inertia), or the mechanical i properties (tensile strength, and/or code allowable stresses) ,' or the ,
geometric configuration, no new malfunction of equipment important to '
-safety is identified.
Since the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits I which could affect the basis for any Technical specification are affected. I i
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F Attachment II to ET 99-0003 Page 165 of 217 l
Safety Evaluation: 59 98-0109 Revision O Control of Heavy Loads Generic Letter (GL) 81-07, " Control of Heavy Loads," was issued on December 22, 1980. The WCNOC 6-month (Phase I) GL 81-07 response was submitted to the NRC under the cover of SLNRC 81-48, dated June 22, 1981.
NUREG-0612, " Control of Heavy Loads," was issued by the USNRC on July 6, 1982, and provided guidance on Phase I and Phase II requirements. The 9 month (Phase II) response was submitted to the NRC under the cover of letter SLNRC 82-033 dated August 4, 1983. The Phase II response was resubmitted with additional information in letters SLNRC 84-0008 and SLNRC 84-0056.
The NRC reviewed the Wolf Creek Phase I and II responses and found them acceptable, as documented in Appendix K & L of Supplement 5 to NUREG- )
0881. Wolf Creek was one of four plants selected by the NRC to have the j Phase IZ response reviewed under a pilot program. GL 85-11 informed !
licensees that, based on the improvements in heavy load handling j subsequent to the implementation of NUREG-0612 (Phase I), further action )
was not required to reduce the risks associated with the handling of heavy loads. The NRC determined that a detailed Phase II review of heavy loads )
was not necessary for any additional plants. The summary below describes I changes made in WCNOC's commitment to NRC Generic Letter 81-07. These changes have been submitted to the NRC by letter dated February 2, 1999.
Summary of Changes WCNOC is making two primary changes to our Phase II response submitted under the cover letter SLNRC 82-033. First, WCNOC is correcting a i statement made in letter SLNRC-82-033 regarding RHR train operability in l Modes 5 and 6 to clarify that only one train may be operable during certain conditions. Secondly, WCNOC is providing a commitment change to allow crane hook movement over an open reactor vessel during refueling i outages. Both of these changes concern movement of heavy loads when the I plant is not operating at power but in a shutdown mode. Each change has been evaluated under the guidance provided in NUREG-0612 and determined to be acceptable.
WCNOC has revised the Reactor Building Analyses section of the Heavy Load Report, Section 2.4, to accurately reflect Modes 5 and 6 Residual Heat Removal (RHR) requirements as stated in our Technical Specifications and to clarify control of the movement of heavy los.ds over the two RHR trains. The original submittal stated that both trains of RHR are operable in Mode 5 pursuant to WCNOC Technical Specifications. WCNOC Technical Specifications require only one train if the Reactor Coolant System (RCS) loops are filled and at least two steam generates contain a water level as specified in Technical Specifications.
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Attachment II to ET 99-0003 Page 166 of 217 To incorporate this change and allow movement of heavy loads in Containment during Mode 5 under Technical Specifications conditions. WCNOC evaluated the effects of a heavy load drop on the RHR system with only one train operable. Should the impacted line break on an operating train of RHR, the RCS fluid would be non-flashing; however, the accident could cause drain down of the RCS to the bottom of loop pipe level and loss of the operating RHR pump if the return line is ruptured. If the ruptured RHR return line is from the only operable (and operating) train, RHR would be lost until the out-of-service train of RHR could be restored. WCNOC concluded that under worst case conditions and assuming no administrative controls that other Emeegency Core Cooling Equipment and inventory would be available to keep :w core cool and from boiling. However, WCNOC has l established administrative controls which will not allow the movement of '
heavy loads over the only operable RHR return line. Similar
- administrative controls are in place for Mode 6. The changes are consistent with the guidance of NUREG-0612.
WCNOC also changed its commitment relating to crane hook travel over an open reactor vessel. Letter SLNRC 82-033 stated that crane hook travel would not be allowed over an open reactor vessel once the reactor upper internals are removed, except during required vessel serving activities.
WCNOC revised Section 1.0 to state that once the upper internals have been removed and fuel is in the reactor vessel, crane hook travel will be prohibited over the open vessel except for the occasional need for reversing the orientation of the main / auxiliary hoists and for required vese.el servicing activities. When there is fuel in the vessel, I administrative procedures will restrict use of the hoist controls when traveling over the open vessel to reverse the hoist orientation and the only item attached to either hook may be the load cell linkage attached to the main hook. The drop of the load block / hook is not considered credible i because redundant and independent limit switches provided on the polar ;
crane main and auxiliary hoists ensures that two-blocking accidents are '
prevented. Pre-operational inspections include the main and auxiliary hoists wire rope and frequent demonstrations that the primary upper limit switch is operable. The stresses in the hoist system are extremely low due to the weight of the load block / hook / linkage compared to the capacity l of the crane. In addition, when there is fuel in the vessel, raising or lowering the hook while traveling over the open reactor vessel will not be i s allowed. This change is consistent with the guidance of NUREG-0612 l
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Attachment II to ET 99-0003 f Page 167 of 217 I I
Safety Evaluation: 59 98-0110 Revision 0 f Technical Requirements Manual Revision he Updated Safety Analysis Report (USAR) Fidelity Review identified h potential discrepancies associated with the Technical Requirements Manual -
(TRM) USAR Sections 16.6.1. Below are the described discrepancies to the f
'TRM.
The revision to TRM Chapter 16 is to remove duplicate information from the !
TRM (USAR) requirements and add Mode 5 of operation to the applicability {
portion. l Mode 5 was added to the applicability because the temperature limit of :
less than or equal to 200 degrees Fahrenheit is applicable to Mode 5.
TRM (USAR) Section 16.6.1.1.a is being revised ;o provide action required to restore the leakage to the acceptable limits when leakage exceeds the i Containment Leakage Rate Testing Program values. Acceptance criteria is i removed the from the action section. ,
TRM (USAR) Section 16.6.1.1.b is being revised because the information in this section is already defined in the TRM requirements.
f TRM (USAR) Chapter 16.6.1.1.b, stated that when the RCS is above 220 degree Fahrenheit the measured combined leakage for all penetrations and l valves subject to Type B and C testing is not to exceed 0.60La or take ]
corrective action and restore within four hours. j Technical Specification 3.6.1.1 requires that Containment Integrity be maintained in Modes 1, 2 , 3, and 4 or restore within one hour. The definition of containment integrity as noted in TRM 1.7.e requires the containment leakage rates determined by Technical Specification 4.6.1.1.c )
are within defined limits to have containment integrity. Technical i Specification 4.6.1.1.c states that the containment leakage rates shall be accordance with Technical Specification 6.8.4.1. This section requires that the leakage rates for Type B and C tests be less than 0.60 La.
However, the time required to restore is one hour per Technical Specification 3.6.1.1.
Thus, the Technical Specification imposes the same requirements as the TRM (USAR) 16.6.1.1, but has more restrictive time to restore.
TRM Section 16.6.1.lb is being revised to refer to Technical Specification 3.6.1.1. The noted section of the TRM is being revised to reflect the applicability of Mode 5 operation. The changes do not change any other information in the USAR. The changes to the USAR do not affect any of the design basis accidents discussed or referenced in USAR Chapters 2, 3, 6, 9, or 15. No plant structures, systems, components or equipment are
Attachment II to ET 99-0003 Page 168 of 217 affected by this change. Since the changes do not affect the operation or function of any SSC as described in the USAR, no credible malfunction of equipment important to safety are identified. These changes to the USAR do not change the acceptance limits that are contained in the bases for the technical specification.
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Attachment II to ET 99-0003 Page 169 of 217 Safety Evaluation: 59 98-0111 Revision: 0 Updated Safety Analysis Report Change to Emergency Core Cooling System Discrepancies in the Updated Safety Analysis Report This Updated Safety Analysis Report (USAR) change is being issued to resolve the discrepancies identified by the USAR Fidelity Review Team. j For clarification purposes, USAR Section 3.1.6 (Discussion for Criterion 37), Paragraphs 4 & 5 are revised as follows: !
Design provisions include special instrumentation, testing, and sampling lines to perform certain tests during plant shutdown to help demonstrate ;
proper automatic operation of the Emergency Core Cooling System (ECCS),
Several subsystems / components of the ECCS can also be tested during normal {
plant operation. A test signal is applied to initiate automatic action and i verifiertion is made that the safety injection pumps attain required {
discharge heads. The test demonstrates the operation of the valves, pump j circuit breakers, and automatic circuitry. In addition, the periodic '
testing of the pumps and valves verify the delivery capability of the l ECCS. USAR Section 7.1.2.5 & Table 7.1-3 provide a discussion of Regulatory Guide 1.22, " Periodic Testing of Protection System Actuation Functions." ;
r The design provided capability to test initially, to the extent practical, the full operational sequence up to the design conditions, including I transfer to alternate power sources, for the ECCS to demonstrate the state I of readiness and capability of the system. This functional test was i performed with the water level below the reactor pressure vessel flange j with the reactor coolant system initially cold and depressurized.
l The statement in the fourth paragraph seemed to imply that the tests can ;
be performed during plant shutdown only. A statement has been added to clarify that some tests can also be performed during normal plant j operation. The reference to Appendix 3A was misleading as the discussion i about Regulatory Guide 1.22, is contained in Section 7.1.2.5 & Table 7.1- l
- 3. The statement about recirculation to the refueling water storage tank I was misleading as a Centrifugal Charging Pump (CCP) is tested with recirculation to volume control tank and Residual Heat Removal (RHR) with I recirculation to suction header. The last sentence about the extent to which the recirculation test is performed is deleted as this section provides a general description of how Wolf Creek Generating Station (WCGS) l meets the General Design Criteria.
The last sentence in the fifth paragraph incorrectly describes how the i initial functional test was performed. This sentence has been corrected. l In addition to the above clarification, this change also revises Figure 3.1(B)-3 to Figure 3.11(B)-3 to correct a typographical error. The first j sentence of Note 6 of Table 3.11(B)-2 (Sheet 6) is revised to read as )
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- Attachment II to ET 99-0003 Page 170 of 217 follows
Following a postulated main steam line break the containment vapor'could become superheated, and the temperature of the vapor could exceed the containment design value of 320 degrees Fahrenheit (F) for a short period of time.
The temperature of 320 degrees F is consistent with containment design temperature of 320 degrees F listed in Table 6.2.1-2 and Mechanical Design
! Criteria M-000, Revision 9. Per Section 6.2.1, the containment design temperature is defined as the design temperature for safety-related i equipment and instrumentation located within the containment and not the maximum temperature allowed for the containment atmosphere vapor during normal plant operation.
l These changes are made to resolve discrepancies identified in PIR 98-2438. These changes do not impact the plant or its operation in any way. '
l This revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision i does not create a possibility for an accident or malfunction of a j different type than any evaluated previously in the safety analysis l report. The margin of safety, an defined in technical specifications, is I
not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question.
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Attachment II to ET 99-0003 Page 171 ef 777 j I
Safety Evaluations. 59 98-0114 Revision 0 i Updated Safety. Analysis Report Change Resulting From Fidelity Review The proposed Updated Safety Analysis Report (USAR) changes are based on ;
comments contained in Performance Improvement Request (PIR) 98-1044 which '
was initiated as a result of the USAR Fidelity Review. This change will:
- 1. Replace the reference to National Bureau of Standards (NBS) with National Institute of Standards and Technology (NIST) traceabl6 standards. '
The change is to update the USAR to reflect the current title of the
{
governmental department associated with traceable standards. :
- 2. Remove mention of performing a "special calibration" in USAR section 13.5.2.2.4 and replace with a more generic discussion of corrective i actions to be taken when validity of analysis results are suspect.
The changes are consistent with existing plant procedures. Therefore, !
there are no effects as a result of the proposed change, t
Change 1 revises the, title of the governmental department associated with traceable standards and does not affect any requirements or regulations.
Change 2 modifies the USAR. discussion of actions taken when the validity ,
of a chemical analysis is suspect. The two procedures associated with !
this are AP 02-008, " Verification of Analytical Performance," and AP 02E- ;
001, "Che:nistry. Calibration Program. " The existing USAR wording focuses !
too narrowly on instrument calibration and fails to recognize that other !
actions may be appropriate to validate analysin results. AP 02-008 states j that the Manager Chemistry / Radiation Protection or designee is responsible for ensuring that corrective actions are taken when data indicates a potential problem (Step 5.1.3). The procedure also provides direction for initiating out-of-calibration reports if an instrument or reagent is not wit hin acceptable limits, and discusses other corrective actions which may '
be appropriate including performance of a second analysis, reviewing procedures and instrumentation, and conducting retraining in the event of ,
technician errors (Section 2.0). The proposed USAR text change recognizes l that actions other than, or in addition to, instrument re-calibration may ;
be necessary to confirm analysis validity. The text change does not constitute a test or experiment. [
h The proposed change does not affect and is not associated with any design basis accident discussed or referenced in the USAR. The proposed change ;
is textual only and, therefore, doesn't create any type of credible !
accident. The proposed text changes do not affect any equipment important to safety either directly or indirectly and, therefore, do not affect any f credible equipment malfunctions. .The proposed changes do not affect any
. acceptance limits for technical specifications or other licensing basis ,
documents.
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Attachment II to ET 99-0003 Page 172 of 217 I
Safety Evaluation: 59 98-0115 Revision 0 k Technical Requirements Manual Revision The Loose Parts Munit9 ring System (LPMS) monitors the Reactor Coolant [
System (RCS) for the presence of metallic loose parts. The LPMS consists [
of 12 active instrumentation channels, each comprising a piezoelectric ?
accelerometer (sensor) , signal conditioning, and diagnostic equipment.
l The system complies with NRC Regulatory Guide 1.133, " Loose-Part Detection l Program for the Pri?ary Systea of Light Water Cooled Reactors," except as !
( noted in Updated Safety Analysis Report (USAR) Chapter 3, Appendix A, j l Conformance to NRC Regulatory Guides, ,
Two redundant sensors are fastened mechanically to the RCS at each of the following potential loose parts collection regions: !
Reactor pressure vessel - upper head region !
Reactor pressure vessel - lower head region !
Each steam generator - reactor coolant inlet region
)
This change revises the Technical Requirements Manual (TRM) as follows: ;
(i, TRM 16.3.1.5, Action a, is revised such that the Limiting condition i for Operation (LCO) will be entered if all channels in one or more !
collection regions become inoperable. The current revision requires I entering the LCO with one or more channels inoperable. f (ii) TRM 16.3.1.5, Action b, is revised to delete the exception to l
General Operational Requirement 16.0.3.3. ;
l Discussion of Item (1) : TRM 16.3.1.5, LCO, states, "The Loose-Part [
( Detection System shall be OPERABLE." The specified function of the LPMS l is to monitor the RCS for loose parts. Since each collection region is ,
j monitored by two redundant sensors, one or more channels con be inoperable l l without affecting the operability of the LPMS provided that no two {
j channels are in the same collection region. j
- l. t Discussion of Item (ii): General Operational Requirement 16.0.3.3 states, "When an LCO is not met, except as provided in the associated ACTION l guidance, action shall be implemented in a timely manner to place the unit >
in a safe condition as determined by plant management." TRM 16.3.1.5,
, Action a, requires that an inoperable LPMS be returned to OPERABLE status l within 30 days. If this requirement can not be met, then entry into j 16.0.3.3 is appropriate. l No design basis accident assumes detection of a loose part by the LPMS.
Therefore, the' proposed change has no potential impact on design basis (
l accidents. The proposed change does not affect the specified function of t the LPMS nor any other SSC. Therefore, no credible accidents could be j
+
created by the proposed change. The proposed change does not affect.the ;
specified function of the LPMS nor any other SSC. Therefore, no credible .
j malfunctions of equipment important to safety could be affected by the ;
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Attachment II to ET 99-0003 Page 173 of 217 proposed change. There are no acceptance limits in the bases for technical specification or any other licensing basis documents that could .
be affected by the proposed change.
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f Attachment II to ET 99-0003
-Page 174 of'217 Safety Evaluation: 59 98-0116 Revision: 0 !
Technical Specification Bases Change l This. proposed change will add additional wording to the " Bases" section of Technical Specifications section 3/4.7.1.2, Auxiliary Feedwater System.
The added wording will clarify the specific requirements for operability ;
of the suction and discharge flow paths that have been determined to be appropriate. This wording change consists of added information to provide-better clarification and ensure consistent application of LCO actions when i various equipment is taken out of service for maintenance or becomes !
The following will be added to the " Bases" section of Technical ;
Specification 3.7.1.2:
"The Auxiliary Feedwater (AFW) system is configured into three independent AFW pumps and associated flow paths. An AFW pump and associated '
discharge flow path is considered OPERABLE when the components and flow paths required to provide redundant AFW flow to the steam generators are OPERABLE. This requires that the two motor-driven AFW pumps be OPERABLE in two diverse paths, each capable of automatically transferring the suction from the condensate storage tank to an Essential Service Water (ESW) supply and supplying AFW to two steam generators. The Turbine-Driven AFW (TDAFW) pump is required to be OPERABLE with redundant steam !
supplies from each of two main steam lines upstream of the Main Steam Isolation Valves (MSIVs), and shall be capable of automatically transferring the suction from the condensate storage tank to an ESW supply l and supplying AFW to the steam generators. The piping, valves, instrumentation, and controls in the required flow paths are also required !
to be OPERABLE. Because each ESW supply flow path to the TDAFW pump ,
provides 100 percent capacity, the " Required ESW Supply" to the TDAFW Pump j is provided by a single, OPERABLE, supply flow path (the suction flow path l begins at the branch connection to the Motor-driven AFW Pump piping and l continues to the suction of the TDAFW Pump) and associated OPERABLE l suction isolation valve."
Risk Considerations Regarding Proposed Technical Specification Wording Changes The general design descriptions for the AFW System, and specifically for the TDAFW pump train, are consistent with the logic and assumptions for the AFW system logic model in the Wolf Creek Generating Station (WCGS) J Probabilistic Safety Assessment (PSA). The AFW system logic model assumes that either supply source from the ESW system would be adequate for suCCensful operation of the TDAFW pump train. Likewise, the AFW system logic model assumes that either steam supply source to the TDAFW pump turbine driver would be adequate for successful TDAFW pump operation.
Application of this change to the WCGS PSA model will not result in an increase in overall plant Core Damage Frequency.
i
Attachment II to ET 99-0003 Page 175 of 217 The above statements have been formulated based on information contained in the various Wolf Creek Nuclear Operating Corporation (WCNOC) design basis documents. The information has been verified as true and correct and data base searches of licensing and design basis documents did not identify any commitments or design restrictions that would prevent the implementation of this additional wording to the basis section of Technical Specification Bases section 3/4.7.1.2. The implementation of this change will not affect the wording in the Updated Safety Analysis Report (USAR) and implementation of this change will not make the information in the USAR no longer true and/or accurate. The implementation of this change will not violate any requirements stated in the USAR.
The AFW System provides sufficient feedwater to the steam generators following a Design Basis Accident (DBA) to remove decay heat and prevent damage to the reactor core. The AFW System is made up of three pumps, two motor driven and one turbine driven. Each of the motor driven pumps has adequate capacity to supply 100 percent of the required feedwater flow to remove decay heat. The one turbine driven pump has twice the capacity of a motor driven pump. The flow from a motor driven pump is sufficient to cool the reactor coolant system down at a rate of 50 degrees Fahrenheit per hour, to 350 degrees Fahrenheit, following a reactor trip from full power. USAR section 5.4.7 provides that the cooldown to 350 degrees Fahrenheit can be accomplished within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the reactor shutdown.
For decay heat removal following a design basis accident, at least two auxiliary feedwater pumps are available, typically one motor driven pump and the turbine driven pump.
In addition to the requirements for DBA, the turbine driven pump must also be capable of functioning for four hours following a station blackout (SBO). During a SBO, the turbine driven pump, taking suction from the Condensate Storage Tank, is the only means to remove decay heat from the Reactor Cooling System (RCS) using auxiliary feedwater.
This wording change does not create any accidents of a different type than those that have been previously evaluated. The function of the AFW system l is not changed or affected by this change. The wording change will provide clarification of existing requirements and will not affect any accident analysis documents.
This wording change will not cause any malfunctions of equipment important j to safety and will not affect any analysis performed previously. The wording change is intended to provide clarification and standardization of l the implementation of the LCO requirements of Technical Specification l 3.7.1.2. This is not a change in procedures or processes but l
clarification to eliminate confusion and possible errors associated with operability determinations when certain equipment is out of service for maintenance or testing.
The Technical Specifications Bases discusses testing requirements to ,
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Attachment II to ET 99-0003 Page 176 of 217 !
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verify .t hat ' the AFW system will provide the proper flow at the appropriate ,
discharge pressure to ensure decay heat is removed and the RCS temperature '
is reduced to less than 350 degrees Fahrenheit so the RHR System may be . ;
placed into operation. The bases does not discuss the requirements for i suction and discharge flow paths. Since the backup suction supply for the- [
TDAFWP consists.of redundant 100 percent capacity supplies from each ESW !
- pump only one supply.is required to maintain operability of the TDAFWP. ;
This lack of bases for the required operable flow path equipment has led '
to confusion when determining OPERABILITY of the AFW system. This change I will add information that supports the design basis of the AFW system to I ensure consistent and accurate identification of required equipment.
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J Attachment II to ET 99-0003 ,
Page 177 of 217 !
1 Safety Evaluation: 59 98-0118 Revision: 0 f
Building Name Change The proposed activity being evaluated herein involves a Updated Safety l
- Analysis Report (USAR) change to appropriately reflect the function of the facility associated with a non-controlled structure. The structure in i question is designated as the Uninterrupted Power Source (UPS) L'_11 ding in I j USAR Figure 2.4-3-07. The building is currently being used as a storage
] facility for snacks and drinks for vending machines throughout the site. ;
, There'is a proposal to use part of this building to store re-cycleable r paper in the future. The basic function of the building remains .
unchanged, i.e., storage. Therefore, this building is being renamed as a l
" Storage Facility" to store' miscellaneous items. This change affects USAR i
! Figure 2.4-3-07.
e The proposed activity has no effect on any procedure, activity,
- administrative control, or sequence of plant operations and does not !
j violate any requirement stated in the USAR. The proposed activity is not !
4 associated with any test or experiment. ,
The proposed activity is a correction of the USAR to properly reflect the i existing condition of the facility, as it has been operated, associated with the UPS building change. The proposed activity has no influence on equipment or parameters associated with any design basis accident .
discussed or referenced in the USAR. f i The proposed activity properly reflects the function of a non-controlled
structure. Due to nature of the function, (storage of miscellaneous items in the affected building) the proposed activity does not create a i j potential for, or has influence on, equipment or parameters that may cause any new credible accidents.
]
. The proposed activity will properly reflect function of the building in the applicable USAR figures. There are no malfunctions of equipment i important to safety which may be directly or indirectly associated with the proposed activity. ,
i The proposed activity does not affect acceptance limits as defined in the Technical Specifications or other licensing basia documents. ,
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Attachment II to ET 99-0003 Page 178 of 217 Safety Evaluation: 59 98-0119 Revision 0 Organization Change The proposed Updated Safety Analysis Report (USAR) change is based on comments contained in Performance Improvement Request (PIR) 98-2721, which was initiated as a result of the USAR fidelity review. The USAR change will add text in Section 13.1.2.2.5 to indicate that responsibilities of the Manager Integrated Plant Scheduling include daily scheduling activities.
The changes are are consistent with existing plant procedures. All required job functions are maintained therefore there are no effects on plant processes, functions, or systems as a result of the proposed change.
No processes or procedures are affected by this change. Also, no other sections of the USAR are affected. The text change does not constitute a test or experiment.
The proposed text change does not affect and is not associated with any design basis accident discussed or referenced in the USAR.
The proposed change is textual only and, therefore, doesn't create any type of credible accident.
The proposed text change does not affect any equipment important to safety either directly or indirectly and, ther<> fore, dces not affect any credible equipment malfunctions.
The proposed change is textual only and does not affect any acceptance limits for technical specifications or other licensing basis documents.
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I Attachment II to ET 99-0003 Page 179 of'217 l i
Safety Evaluations 59 98-0121 Revision: O Piping Replacement Because of Flow Accelerated Corrision !
Configuration' Change Package (CCP) 07837, Revision 0, is issued to provide i guidelines for the modification / replacement of the Auxiliary Turbines ,
System (FC) piping lines FC-001-DBB-4, FC-002-DDB-4, FC-008-DBB-1 and FC- [
009-DBB-1,'to mitigate abnormal pipe-wall thinning due to Flow Accelerated '
Cerrosion (FAC). The piping is being replaced with a low alloy steel (2 {
1/4 Cr - 1 Moly) which improves piping resistance to FAC. l i
The proposed modification changes the sockolet connections in lines FC-008- !
DAB-1" and FC-009-DAB-1" to weldolet connections. This change enhances the i original design configuration by improving the stress levels at the branch connection points. [,
The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertia), or the geometric ;
configuration of the piping. The mechanical properties such as tensile !
strength and code allowable stresses will remain unchanged. The lower l yield strength and higher Young's Modulus is judged to have insignificant !
impact on the original analysis. Therefore, the change does not adversely i affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.
The proposed replacement does not adversely affect any system, component, i or procedures required to mitigate the consequences of an accident ;
previously evaluated in the Updated Safety Analysis Report. The proposed l' change will restore a degraded section of the affected piping system, to '
its original design configuration (piping geometry, cross section, support location, fittings).
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical l properties, excess strain, mechanical creep etc., are credible failure ,
modes for which the proposed piping replacement has been evaluated, !
through a critical characteristics comparison to the existent piping ,
system design. Based on the evaluation, it was concluded that a new credible failure mode is not introduced. Therefore, there are no malfunctions of equipment important to safety identified.
The. proposed change will restore a degraded section of the affected piping [
system, to perform its original design intent. The proposed replacement i does not involve or affect any safety related system or component. All f system functions will continue to be performed as designed. [
i Since the proposed change will restore a degraded section of the affected j piping system, to its original design configuration (piping geometry, !
cross section, support location, fittings), no accidents are affected, j I
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Attachment II to ET 99-0003 Page 180 of 217 i
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no new accidents could be created.
The proposed pipe replacement did not change the cross sectional properties (section modulus, moment of inertia), or the mechanical properties (tensile strength, and/or code allowable stresses), or the geometric configuration, no new malfunction of equipment important to safety is identified.
Since the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits which could affect the basis for any Technical specification are affected.
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Att? ; dent II to ET 99-0003 Page 181 of 217 Safety Evaluation: 59 98-0122 Revision 0 Addition of Welding Recepticles This modification provides for the addition of two cables to supply new welding receptacles inside containment. The new welding receptacles will be installed on the 2000' and 2068' elevation (one new receptacle on each
- elevation). The new cables installed as part of this modification will increase by an small amount the combustible loading in zone RB8 within containment. The combustible fire loading for zone RB8 will increase from 8,869 btu /ft2 to 8,915 btu /ft2. Updated Safety Analysis Report (USAR)
Section 9.5B is being revised to indicate the new combustible loading values. The additional combustible loading is due to cable insulation.
The new combustible loading for the applicable rooms is ctill less than the value needed to declare an area a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating (i.e. 240,000 BTU's/ft2). USAR sections 9.5.1.2.2.3 and Table 9.5A-1 both reference ASTM E-119 as the applicable fire standard. ASTM E-119 provides 240,000 BTU's/ft2 as a conservative loading estimate for sustaining a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire. Specifically, ASTM E-119, paragraph X5.3 notes that specifications for fire resistance in regulatory documents continue to be based largely on the fire load concept. The concept incorporates the premise that the duration of a fire is proportional to the fire loading, that is, the mass of combustible materials per unit floor area. The relationship between the mass of combustible materials and fire duration was established on the basis of burnout tests in structures incorporating materials having caloric or potential heat values equivalent to wood and paper, that is 7000-8000 BTU /lb. In other words, the above premise states that 10 lb. of material per square foot (80,000 BTU /ft2) will produce a fire of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> duration. The NFPA Fire Protection Handbook, page 7-111, also documents tis conservative estimate. This estimation is based upon the cable being specified in accordance with IEEE 383 (i.e. qualified as flame retardant).
l There are no procedures, activities or any other actions (e.g.
administrative controls, system operation) which will be impacted by this modification. This modification will help ensure that plant personnel do not violate any cable separation requirements by installing more available
)
l power inside containment. l I
There are no accident scenarios within the USAR which need to be reviewed )
for impact by this change. This modification will only add more available l power receptacles inside containment. l There are no credible accident scenarios which this modification could create. This modification will help eliminate transient cable separation violations.
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l Attachmeat II to ET 99-0003 Page 182 of 217 -
Safety Evaluation: 59 98-0123 Revision: 0 ;
Revision to Procedure SYS GK-200 !
i Procedure SYS GK-200, " Inoperable IE A/C Unit," is being modified to ,
incorporate interim compensatory measures allowed by configuration Change i Package (CCP) 07905. This CCP is a temporary use-as-is disposition which is applicable.only during the time that corrective actions are being i implemented in accordance with the corrective action plan developed for ;
Performance Improvement Request (PIR) 98-3259. Therefore, this Unreviewed l Safety Question Determination is applicable only during the time that CCP ;
07905 is applicable.
This procedure revision implements the following changes:
Temperature of the rooms shall be maintained at or below 80 degrees [
Fahrenheit (F) at all times. ,In addition to required monitoring, room [
temperature trending is required, and Engineering must be notified if the i room temperature exceeds 75 degrees F. j l
The surrounding area temperatures must not be in excess of the following:
Outside ambient - 97 degrees F, Auxiliary Building - 104 degrees F, Diesel Generator Building - 104 degrees F, Communication-Corridor - 97 degrees F, !
Access Control - 85 degrees F, Lower Cable Spreading Room, - 104 degrees F. [
If an Engineered Safety Features (ESF) actuation occurs, all doors shall remain open for the duration of the actuation time. If a fire occurs in !
any room with the doors listed below in the open position, the doors in l all the rooms cooled by both Class IE Electrical Equipment A/C Unit l (SGK05) units shall be closed immediately, even if an ESF actuation has occurred. The interior doors listed below must be opened and remain open f with appropriate firewatches for the duration of normal plant operation l with one SGK05 unit out of service: 33011, 34051, 34081, 34131, 33023, [
34052, 34082, 34141, 34041, 34071, 34101, 34042, 34072, 34111. !
l This temporary use-as-is disposition provides Engineering justification l that interim operation is allowable, with restrictions, with one SGKOL {
unit out of service. It provides a list of contingence actions, ;
acceptable operating conditions, and restrictions that would provide reasonable assurance that, if one SGK05 A/C unit becomes unavailable during normal operations or required maintenance activities, the other train SGK05 A/C unit would provide adequate cooling for both trains.
The activities associated with establishing a maximum room temperature and ;
monitoring temperatures are routine in nature and have no impact on other SSCs not does it adversely impact or interfere with other plant i activities. -Opening the doors between cooled rooms violates the train separation criteria assumed in the current safety analysis, but the associated CCP justifies continued acceptability of this configuration ,
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Attachment II to ET 99-0003 Page 183 of 217 until the end of applicability of the CCP. In addition, opening the doors between cooled rooms violates the hazards barrier functions of the doors. 7 However, other existing provisions of SUS GK-200 mandate that the open doors are addressed procedure by AP 10-104, " Breach Authorization," and procedure AP10-103, " Fire Protection Impairment Control Procedure."
Therefore the activities associated with this procedure change have no adverse impact on other parts of the Wolf Creek Generating Station.
Based on the above, this revision will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. '
This revision does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report. The margin of safety, as defined in technical specifications, is not reduced by this revision. Therefore, this revision does not involve any unreviewed safety question. l
Attachment II to ET 99-0003 Page 184 of 217 Safety Evaluation: 59 98-0124 Revision: 0 Revisision of Drawings to Reflect Room Use and Portal Monitor Locations Configuration Change Package (CCF) 07906 revises Drawings 10466-A-1701, (Radiation Zones - Normal Operation EL. 1974'-0"), 10466-A-1704, (Radiation Zones - Normal Operation EL. 2047'-6"), and M-1G050, (Equipment Locations Control Bldg. & Comm. Corridor Plan EL. 1974'-0" & EL. 1984'-
0"), and the associated Updated Safety Analysis Report (USAR) Figures.
These drawings are revised to show the correct locations of portal monitors in the plant and to reflect the current usage of rooms utilized by Health Physics based on comments documented in Performance Improvement Request (PIR) 98-2615, which was initiated during the USAR Fidelity Review. Specific changes to drawing 10466-A-1701 include:
a) Delete the two portal monitors shown in room, b) Delete the portal monitor shown at the door to room 3201.
c) Add a portal monitor at the entrance to room 3207 (zone 3-H).
d) Delete the portal monitor shown at the sign in/out area (rooms 3220/3221). Delete the walls shown between rooms 3220 and 3221 (zone 3-H).
e) For room 1115 in the drawing legend, delete Positive Displacement Charging Pump Room and add Normal Charging Pump Room.
f) For room 3202 in the drawing legend, delete Controlled H.P. Tool &
Storage Room and add H.P. Count Room.
g .) For room 3222 in the drawing legend, delete Health Physicist's Office and add Health Physics Office.
Specific changes to drawing 10466-A-1704 include:
a) Delete the portal monitor shown at stairwell A-2 (zone 3-D).
Specific changes to drawing M-1G050 include:
a) At zone 6-F, delete Controlled H.P. Tool & Inst. Storage Rm. and add H.P. Count Room.
b) At zone 4-F, delete Health Physicist's Office and add Health Physics Office.
In addition to the drawing changes discussed above, the following USAR changes are being made to resolve the comments raised in PIR 98-2615:
a) Add Area Radiation Monitors 0-SD-RE-43 and 0-SD-RE-44 to Table 12.3-4, Power Supplies For Area and In-Plant Airborne Monitors. 0-SD-RE-43 is supplied by Technical Support Center Emergency Diesel Generator on loss of
.offsite power. 0-SD-RE-44 is supplied by Emergency Operating Facility Emergency Diesel Generator on Loss of offsite power. These Area Radiation Monitors were previously included in Table 12.3-2, Area Radiation Monitors but not Table 12.3-4.
b)- Revise Table 12.5-1, Health Physics and Lab Equipment, to reflect the current equipment available:
l Attachment II to ET 99-0003 Page 185 of 217 ,
- 1) Computer Based Gamma Spectroscopy System - change two to seven.
- 2) Gas Proportional Counter - change one to two.
- 3) Pocket Dosimeter Charger - change five to ten
- 4) Condenser R Meter or Digital Dosimeter - change one to three, c) In section 12.3.4.2.2.1.9, Ranges and Retpoints, change the first paragraph to read: "The ranges of the various airborne radioactivity monitors were chosen based on the detection of radioactivity in concentrations ranging from 10 DAC-hours or lower in compartments served up to those from postulated accidents." This change makes section 12.3.4.2.2.1.9 consistent with sections 12.3.4.2.1.2, Power Generation Design Basis Two, and 12.3.4.2.2.1.8, Sensitivities, both of which use DAC-hours as a measure of airborne radioactivity.
The CCP 07906 drawing changes will make drawings 10466-A1704, 10466-A-1704 and M-1G050 consistent with the current plant configuration. Specific areas corrected include:
a) Room descriptions are changed to reflect the current usage. These changes were concerned with areas used by Health Physics except for room 1115 which contains the Normal Charging Pump. PMR 04590 replaced the Positive Displacement Charging Pump with the Normal Charging Pump.
b) The location of Portal Monitors was revised to reflect the current location of Portal Monitors in the plant.
c) The removal of the wall between rooms 3220 and 3221 as shown on drawing 10466-A-1701. This wall does not exist as shown on drawing M-1G050. There is no field work associated with CCP 07906. The additional changes included in the USAR change revise Table 12.3-4 to include the power supplies for two area radiation monitors previously omitted; revise Table 12.5-1 to reflect the increase number of Health Physics and Lab Equipment available; and revise section 12.3.4.2.2.1.9 to use DAC-hours for airborne radioactivity concentrations. The drawing changes made by CCP 07906 and the associated USAR changes are documentation changes that !
do not affect any SSC nor do they change the performance of activities that are important to the safe and reliable operation of WCGS. The drawing changes made by CCP 07906 and the associated USAR changes do not impact any procedures, activities, administrative controls, or sequence of plant operations nor are any plant structures, systems, components or i equipment impacted. No requirements outlined in the USAR are revised by l these changes. No other USAR descriptions or conclusions will change or be made untrue as a result of these changes. No test or experiments are involved with these changes.
The drawings revisions made by CCP 07906 and the associated USAR changes )
revise room descriptions to reflect current usage; update the location of ;
portal monitors in the plant; remove a wall shown on a radiation zone l drawing that does not exist; add power supplies for two area radiation monitors previous omitted; update the numbers of Health Physics and Lab ,
equipment available; and revise the units used for airborne radioactivity for consistency. There is no additional impact on the performance of plant l activities nor affect on any SSC. Therefore, no Design Basis Accident is I
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l Attachment II to ET 99-0003 f
i Page 186 of 217 l identified for review.
i The drawings revisions made.by CCP 07906 and the associated USAR changes !
revise room descriptions to reflect current usage; update the location of [
portal monitors in the plant; remove a wall shown on a radiation zone drawing that does not exist; add power supplies for two area radiation {
monitors previous omitted; update the numbers of Health Physics and Lab l equipment available; and revise the units used for airborne radioactivity j for consistency. CCP 07906 and the associated USAR changes make no !
additional changes to the plant, do not affect the performance of plant ,
activities and do not affect any SSC. Therefore,'no credible accidents !
that could be created are identified. l l
The drawings revisions made by CCP 07906 and the associated USAR changes ;
revise room descriptions to reflect current usage; update the location of i
. portal monitors in the plant; remove a wall shown on a radiation zone !
drawing that does not exist; add power supplies for two area radiation j monitors previous omitted; update the numbers of Health Physics and Lab I I
equipment available; and revise the units used for airborne radioactivity for consistency. CCP 07906 and the associated USAR changes make no '
additional changes to the plant, do not affect the performance of plant activities and do not affect any SSC. Therefore, no credible malfunctions ;
of equipment important to safety are identified. '
i The drawings revisions made by CCP 07906 and the associated USAR changes i revise room descriptions to reflect current usage; update the location of
{
portal monitors in the plant; remove a wall shown on a radiation zone :
drawing that does not exist; add power supplies for two area radiation monitors previous omitted; update the numbers of Health Physics and Lab ]
equipment available; and revise the units used for airborne radioactivity ,
for consistency. CCP 07906 and the associated USAR changes make no ]
additional changes to the plant, do not affect the performance of plant ,
activities and do not affect any SSC. Therefore, no acceptance limits are i identified that could be affected. !
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Attachment II to ET 99-0003 Page 187 of 217 Safety Evaluation: 59 98-0125 Revision 0 Modification /Heplacement of Degraded Piping Configuration Change Package (CCP) 07757 is issued to provide guidelines for the modification / replacement of a degraded section of Feedwater Heater Extraction-Drains and Vents (AF) System piping line AF-451-DBD-4, to mitigate abnormal pipe-wall thinning due to Flow Accelerated Corrosion (FAC). This piping is being replaced with a low alloy steel 2 1/4 Cr-1 Mo (The replacement section is identified by new line number AF-642-DAD-4) which improves piping resistance to FAC.
The proposed pipe replacement does not change the cross sectional properties (section modulus, moment of inertia), or the geometric configuration the pipe. The mechanical properties such as tensile strength and code allowable stresses will remain unchanged. The lower yield strength and higher Young's Modulus is judged to have insignificant impact on the original analysis. Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.
l The proposed replacement does not adversely affect any system, component, or procedures required to mitigate the consequences of an accident previously evaluated in the Updated Safety Analysis Report. The proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings). Ductile fracture, erosion / corrosion, loss of mechanical properties, excess strain, mechanical creep etc., are credible failure modes for which the proposed piping replacement has been i evaluated, through a critical characteristics comparison to the existent piping system design.
Based on the evaluation, it was ccsacluded that a new credible failure mode is not introduced. Therefore, there are no malfunctions of equipment
( important to safety identified. The proposed change will restore a l degraded section of the affected piping system, to perform its original l design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.
j Since the proposed change will restore a degraded section of the affected piping system to its original design configuration (piping geometry, cross section, support location, fittings), no accidents are affected as being associated with this change.
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no new accidents could be created.
l Attachment II to ET 99-0003 Page 188 of 217 The proposed pipe replacement does not change the cross sectional l properties (section modulus, moment of inertia), or the mechanical properties (tensile strength, and/or code allowable stresses), or the >
geometric configuration. Therefore, no malfunction of equipment important to safety are identified.
Since, the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits which could affect the basis for any Technical specification are identified.
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l_ Attachment II to ET 99-0003 i
l Page 189 of 217 Safety Evaluation: 59 98-0126 Revision: 0 Modification /Replacsment of Piping to Mitigate Abnormal Pipe-wall Thinning Configuration Change Package (CCP) 07798 is issued to provide guidelines ,
[ for the modification / replacement of Main Steam System (AB) piping line AB-l 188-DBD-4, to mitigate abnormal pipe-wall thinning due to Flow Accelerated .
Corrosion (FAC). The piping is being replaced with a low alloy steel (2 1/4 Cr - 1 Moly) which improves piping resistance to FAC.
1 The-proposed pipe replacement does not change the cross sectional properties (section modulus, moment of inertia) , or the geometric configuration of the pipe. The mechanical properties such as tensile !
strength and code allowable stresses will remain unchanged. The lower yield strength and higher Young's Modulus is judged to have insignificant i impact on the original analysis. Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.
l The proposed replacement does not adversely affect any system, component, or procedures required to mitigate the consequences of an accident l previously evaluated in the U pdated Safety Analysis Report (USAR). The proposed change will restore a degraded section of the affected piping i l
system, to its original design configuration (piping geometry, cross l section, support location, fittings). Ductile fracture,
- i erosion / corrosion, loss of mechanical properties, excess strain, ;
mechanical creep etc., are credible failure modes for which the proposed l piping replacement has been evaluated, through a critical characteristics comparison to the existent piping system design. Based on the evaluation,
' it was concluded that a new credible failure mode is not introduced.
Therefore, there are no malfunctions of equipment important to safety identified.
The proposed change will restore a degraded section of the affected piping l system, to perform its original design intent. The proposed replacement ;
l does not involve or affect any safety related system or component. All l system functions will continue to be performed as designed.
! Since the proposed change will restore a degraded section of the affected ;
piping system, to its original design configuration (piping geometry, 1 cross section, support location, fittings), no accidents are affected as '
being associated with this change.
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, crosa section, support location, fittings), no new accidents could be
. created.
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Attachment II to ET 99-0003 Page 190 of 217 The proposed pipe replacement does not change the cross sectional properties (section modulus, moment of inertia), or the mechanical properties (tensile strength, and/or code allowable stresses), or the geometric configuration. Therefore, no malfunctions of equipment important to safety are identified.
Since, the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits which could affect the basis for any Technical specification are identified. ,
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Attachment II to ET 99-0003 Page 191 of'217 I
t Safety Evaluation: 59 98-0127 Revision: 0 Modification / Replacement of Piping to Mitigate Abnormal Pipe-wall Thinning' Configuration Change Package (CCP) 07754 is issued to provide guidelines i for the modification / replacement of Auxiliary Turbine System (FC) FC-095- l
.RBD-1, FC-085-HBD-1, FC-075-HBD-1 piping lines, and Condensate System (AD) !
piping line, and AD-151-HBD-4, to mitigate abnormal pipe-wall thinning due '
to Flow Accelerated Corrosion (FAC) . The piping is being replaced with a low alloy steel (2 1/4 Cr - 1 Moly) which-improves piping resistance to FAC.
'The proposed pipe replacement does not change the cross sectional ,
properties (section modulus, moment of inertia), or the geometric l configuration of the pipe. The mechanical properties such as tensile l I
strength and code allowable' stresses will remain unchanged. The lower yield strength and higher Young's Modulus is judged to have insignificant ;
impact on the original analysis. Therefore, the change does not adversely affect the existing safety margins or structural integrity of the affected piping system. The piping stresses will remain acceptable within code allowables.
The proposed replacement does not adversely affect any system, component, i or procedures required to mitigate the consequences of an accident previously evaluated in the Updated Safety Analysis Report. The proposed [
change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings).
Ductile fracture, corrosion, erosion / corrosion, loss of mechanical i properties, excess strain, mechanical creep etc., are credible failure modes for which the proposed piping replacement has been evaluated, through a critical characteristics comparison to the existent piping i system design. Based on the evaluation, it was. concluded that a new credible failure mode is not introduced. Therefore, there are no malfunctions of equipment important to safety identified.
The proposed change will restore a degraded section of the affected piping system, to perform its original design intent. The proposed replacement does not involve or affect any safety related system or component. All system functions will continue to be performed as designed.
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry, cross section, support location, fittings), no accidents are affected as being associated with this change.
Since the proposed change will restore a degraded section of the affected piping system, to its original design configuration (piping geometry,
Attachment II to ET 99-0003 Page 192 of 217 cross section, support location, fittings), ne accidents could be created.
The proposed pipe replacement does not change the cross sectional properties (section modulus, moment of inertia), or the mechanical properties (tensile strength, and/or code allowable stresses), or the geometric configuration. Therefore, no malfunction of equipment important to safety are identified.
Since, the proposed change will restore a degraded section of the affected piping system, to perform its original design intent, no acceptance limits which could affect the basis for any technical specification are identified.
i
Attachment II to ET 99-0003 Page 193 of 217 i Safety Evaluation: 59 98-0128 Revision: 0 Temporary Modification to Defeat Power Load Unbalance Circuitry This-Temporary Modification will remove Power Load Unbalance (PLU) circuit cards 1PU2-A001 and 1PU3-A005, from AC119 to defeat the PLU circuitry.
The affects of removing these circuit cards disables the " anticipatory" overspeed protection function described in Updated Safety Analysis Report (USAR) Section 10.2.2.3.2 and Vendor Manual M-800-0231. This circuit provides a rate sensitive loss of load signal to initiate control valve (CV) and intercept valve (IV) fast closure and.is accurately described as ;
' anticipatory turbine overspeed protection. At lower loads, the circuit acts when turbine power exceeds generator load by approximately 40 percent ,
and the load was lost at a rate equivalent to going from rated to zero i load in approximately 35 msec. At full load, the rate function is reduced so that a full load loss rate of approximately 55 msec is required for circuit actuation.
Upon loss of generator load, the Electrohydraulic Control (EHC) system acts to prevent rotor speed from exceeding design overspeed. Failure of any single component will not result in rotor speed exceeding design overspeed (120 percent of rated speed). Redundancies are employed between the following:
Main stop valves / Control valves, t Intermediate stop valves / Intercept valves, l Primary speed control / Backup speed control, i Fast acting solenoid valves / the Emergency trip fluid system (ETS),
Speed control / Overspeed trip / Backup overspeed trip.
]
The fast acting solenoid valves would normally initiate fast closure of turbine control valves by dumping ETS pressure under load reject i conditions signaled by the PLU. Under rotor acceleration, whether the PLU signal to the solenoid valves works or not, ETS pressure would be dumped by any one of several devices in the front standard including the mechanical and electrical overspeed trips. These overspeed trip devices are set at 110 percent (mechanical) and 111 percent (electrical) of normal operating speed (1800 rpm) and will trip all of the turbine valves (Control, Stop, Intercept, and Intermediate Stop). These overspeed trip devices will remain in operation to preclude an overspeed event. l Event table 10.2.1 of the USAR postulates equipet t failures following a loss of turbine load. The maximum speed that cou._ be reduced upon full loss of load from 100 percent is 109 percent with a normally operating control system which includes the primary or backup speed control systems. Assuming that both the power to load unbalance and speed control systems had failed prior to a loss of load, the following maximum speeds could be achieved during this event: 119 percent peak transient speed with normal control system failure and operation of mechanical overspeed trip, 1
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Attachment II to ET 99-0003 Page 194 of 217 120 percent peak transient speed with failure of both normal control systems and mechanical overspeed trip, but proper operation of backup overspeed trip.
i The turbine / generator assembly is designed to withstand normal conditions and anticipated transients without loss of structural integrity. Multiple failures of the system would be required to exceed the original design criteria of the unit. Additionally, each fully bucketed turbine rotor assembly has been spin tested at 20 percent overspeed at time of manufacture, and there are no refueling generator, rotor, or disk inspection results have been noted which would jeopardize the original design criteria of the unit.
USAR Section 10.2.2.3.2 describes the PLU circuit as being an
" anticipatory" overspeed protection circuit. The PLU protection failed >
due to failure of power supply IPI-G003. Repair of this supply and reinsertion of cards 1PU2-A001 and 1PU3-A005 at power could result in a turbine and reactor trip. With this circuit disabled. this paragraph will no longer be true until the power supply is replaced and circuit cards are placed in service. l There is only one accident previously evaluated in the USAR to which this change is applicable. That event is a' Main Generator load rejection which i
has the potential to ree'21t in a turbine / generator overspeed condition.
In this event, USAR Sectaon 15.2.2.1 states that any increased frequency to the reactor coolant pump motors will result in slightly increased flow rate and subsequent additional margin to safety limits. No other safety related pump motors, reactor protection system equipment, or other safety l related loads are subjected to overfrequency conditions. No l turbine / generator unit design limits will be exceeded.
This change is not related to any of the events or conditions associated with the initiation of credible accidents. This change does not pose any ,
credible malfunctions to any equipment important to safety, either directly or indirectly.
I Upon loss of generator load, the EHC system acts to prevent rotor speed (
from exceeding design overspeed. The discussion of overspeed protection J in USAR Section 10.2.2.3.2 and Table 10.2-1 describes how the failure of this " anticipatory" overspeed protection does not cause the design overspeed value to be exceeded. This change does not affect any I acceptance limits specified in the USAR. The only applicable acceptance limits specified in the licensing basis documents are those related to a turbine trip event. The minimum departure from nucleate boiling ratio remains greater than the limit of 1.3 and the RCS pressure limit of 2250 psia is not exceeded. l i
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Attachment 'II to ET 99-0003 :
Page 195 of 217 i
Safety Evaluation: 59 98-0128 Revision: 1 I Temporary Modification to Defeat Power Load Unbalance Circuitry }
Revision 1 to Unreviewed Safety Question 98-0128 provides corrected values f for minimum departure from nucleate boiling ratio and the Reactor Coolant
-System pressure limit. This Temporary Modification will remove Power Load }
Unbalance (PLU) circuit cards 1PU2-A001 and 1PU3-A005, from AC119 to I defeat the PLU circuitry. I The affects of removing these circuit cards disables the anticipatory I overspeed protection function described in Updated Safety Analysis Report (USAR) Section 10.2.2.3.2 and Vendor Manual M-800-0231. This circuit provides a rate sensitive loss of load signal to inftiate control valve 1 (CV) and intercept valve (IV) fast closure and is acc.'rately described as '
anticipatory turbine overspeed protection. At lower leads, the circuit ;
acts when turbine power exceeds generator load by approximately 40 percent I and the load was lost at a rate equivalent to going from .ated to zero i load in approximately 35 msec. At full load, the rate function is reduced i so that a full load loss rate of approximately 55 msec is required for i circuit actuation.
I Upon loss of generator load, the Electrohydraulic Control (EHC) system i acts to prevent rotor speed from exceeding design overspeed. Failure of )
any single component will not result in rotor speed exceeding design 1 overspeed (120 percent of rated speed). Redundancies are employed between l the following: l Main stop valves / Control valves, ,
Intermediate stop valves / Intercept valves, Primary speed control / Backup speed control, )
Fast acting solenoid valves / the Emergency trip fluid system (ETS), i Speed control / Overspeed trip / Backup overspeed trip. 1 The fast acting solenoid valves would normally initiate fast closure of i turbine control valves by dumping ETS pressure under load reject )
conditions signaled by the PLU. Under rotor acceleration, whether the PLU l signal to the solenoid valves works or not, ETS pressure would be dumped i by any one of several devices in the front standard including the j mechanical and electrical overspeed trips. These overspeed trip devices J are set at 110 percent (mechanical) and 111 percent (electrical) of normal operating speed (1800 rpm) and will trip all of the turbine valves (Control, Stop, Intercept, and Intermediate Stop). These overspeed trip devices will remain in operation to preclude an overspeed event.
Event table 10.2.1 of the USAR postulates equipment failures following a loss of turbine load. The following speed that could be reduced upon full loss of load from 100 percent is 109 percent, with a normally operating control system which includes the primary or backup speed control systems. Assuming that both the power to load unbalance and speed control
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l Attachment II to ET 99-0003 l Page 196 of 217 l
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systems had' failed prior to a loss of load, the following maximum speeds !
could be achieved during this event: 119 percent peak transient speed l with normal control systein failure and operation of mechanical overspeed l trip, 120 percent peak transient speed with failure of both normal control j systems and mechanical overspeed trip, but proper operation of backup i overspeed trip. i I
The turbine / generator assembly is designed to withstat.d normal conditions ,
and anticipated transients without loss of structural integrity. Multiple l
failures of the system would be required to exceed the original design I criteria of the unit. Additionally, each fully bucketed turbine rotor j assembly has been spin tested at 20 percent overspeed at time of manufacture. No refueling generator, rotor, or disk inspection results have been noted which would jeopardize the original design criteria of the i unit.
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USAR Section 10.2.2.3.2 describes the PLU circuit as being an anticipatory overspeed protection circuit. The PLU protection failed due to failure of l power supply 1PI-G003. Repair of this supply and reinsertion of cards l 1PU2-A001 and 1PU3-A005 at power could potentially result in a turbine and i reactor trip. With this circuit disabled, this paragraph will no longer be !
true until the power supply is replaced and circuit cards are placed in I service. l USAR 15.2'2.1
. LOSS OF EXTERNAL ELECTRICAL LOAD: i Main Generator load rejection has the potential to_ result in a turbine / generator overspeed condition. In this event, USAR Section 15.2.2.1 states that any increased frequency to the reactor coolant pump motors will result in slightly increased flow rate and subsequent l additional margin to safety limits. No other saft.?r related pump motors, !
reactor protection system equipment, or other safety related loads are l subjected to over frequency conditions. However, the PLU circuitry is l assumed to have failed prior to loss of load and no turbine / generator unit j design'11mits will Le exceeded. Therefore, Loss of External Electrical l Load is not affected by this change.
USAR 3.5.1.3 TURBINE MISSILES -
Event table 10.2.1 of the USAR assumes the PLU circuitry failed prior to f loss of. load. Therefore, the turbine misrile analysis is not affected by i defeating the PLU circuitry. {
i The NNSR PLU circuitry was installed by the turbine manufacturer for i commercial reasons. Event table 10.2.1 of the USAR assumes the PLU l' circuitry failed prior to loss of load. Therefore, this change is not related'to any of the events or conditions associated with the initiation l of credible accidents. l There is no equipment important to safety associated with the NNSR PLU f i
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Attachment II to ET 99-0003 Page 197 of 217 circuitry. .This' circuitry was installed by the turbine manufacturer for l commercial reasons. Therefore, this change does not pose.any credible l malfunctions to any equipment important to safety, either directly or j indirectly. ;
t i
Upon loss of generator load, the EHC system acts to prevent rotor speed l from exceeding design overspeed. The discussion of overspeed protection j in USAR Section 10.2.2.3.2 and Table 10.2-1 describes now the failure of l this anticipatory overspeed protection does not cause the design overspeed !
value to be exceeded. This change does not affect'any acceptance limits ;
specified in the USAR. The acceptance limits specified in Amendment 99 to !
the Facility Operating License No. NPF-42 continue to be met. The minimum departure from nucleate boiling ratio remains greater than the limit of (
1.76 and the RCS pressure limit of 2750 psia is not exceeded. -
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I l-Attachment II to ET 99-0003 l Page 198 of 217 5 Safety Evaluation: 59 98-0129 Revision 0 -
Clarification of Updated Safety Analysis Report Description of Plant .
Winterization Activities !
This change to the Updated Safety Analysis Report (USAR) results from the !
USAR Fidelity Review and is applicable to non-safety related systems. I This USAR change will modify the following Sections of the US 9.4.3.2.3 i (18th paragraph) and Section 9.4.10.2.3, as follows: l
- 1) Section 9.4.3.2.3 (18th paragraph) : The phrase " Normal Charging Pump I Room Cooler" was changed to " Normal Charging Pump Fan Coil Unit" for !
accuracy and consistency with the other units in the paragraph. !
- 2) 9.4.10.2.3: The etatement was changed from "The CeCWS provides the !
cooling medium for the ventilation cooling coils all year round," to "The CeCWS provides the cooling medium for various ventilation cooling coils all year round, except for those cooling coils that are isolated and drained for plant winterization." ,
These changes to the USAR are necessary to ensure consistency and !
completeness and will ensure the USAR reflects the current, approved method of plant operations. These changes correct the title of a non- !
safety related cooler, and clarify how our plant winterization activities i are performed. As a result, these changes do not affect safety related equipment.
Therefore,. these changes do not affect the approved physical plant or {'
equipment operations during either normal or transient conditions.
Finally, since no equipment functional requirements are affected, no -l change to the plant response or accidents are possible as a consequence of l the USAR wording changes represented by this change. )
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Page 199 of 217 i
Safety Evaluationt 59 98-0137 Revision: 0 danhole Removal on Essential Service Water Electrical Vaults This Unreviewed Safety Question Determination (USQD) is written to i evaluate Revision 9 to AP 10-104, " Breach Authorization," which will !
permit removal of Essential Service Water (ESW) Electrical Vaults manhole :
access covers on one safety related train of equipment. This revision is !
applicable to the personnel access covers for either train and for one or i more access covers to a single train of safety related equipment which ;
will be removed to facilitate inspection, maintenance, or modifications. i The cover (s) will be removed while the equipment within the enclosure is !
operable and intended to remain operable as defined by Wolf Creek Generating Station Technical Specifications. Covers may only be off or i removed from one train of equipment at a time.
i Removal of the vault or access covers will not create a seismic II/I !
condition or missile hazard. The outside diameter of each personnel f access cover or vault cover is significantly larger than its respective ;
opening inside diameter; thus, cover drop failure into the enclosure is i not likely. The vault covers are reinforced concrete, and rigging controls limit lift height to a minimum; thus, a dropped cover will not result in fracture and multiple missiles that could enter the opening. Lift l instructions will be specified via the Work Package when the equipment within the enclosure is operable. f After lifting, the vault covers will be placed aside from the vault openings and attached rigging relaxed to assure the vault covers do not remain suspended. All rigging or hoists that are to remain suspended over the vault (for example, a temporary A- i frame with hoists) will be moused or restrained so as to prevent them from ,
posing a threat from unscable equipment. ,
In the unlikely ever.t of a load drop, train separation ensures that the ESW system function would not be lost. Thus lifting a load over one train at a time would not be inconsistent with the guidelines of NUREG -0612 (Control of Heavy Loads).
No protective measures are required to compensate for missiles generated from the failure.of plant equipment or postulated accidents. While .
missiles from' rotating component failure, pressurized component failure, l and from accidents occurring within the proximity of the site have been postulated, the missiles from these sources have a sufficiently low probability that no specific protective measures for the other vaults are ;
required. Plant layout minimizes the possibility of a turbine missile ;
impacting plant structures and equipment essential for safe shutdown requirements. The ESW Electrical Vaults are located South of the expected trajectory of potential missiles that would be generated by a turbine / generator failure.
Contingency protection will be provided for postulated missiles from ,
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Attachment II to ET 99-0003 Page 200 of 217 inclement weather. Sufficient personnel and equipment will remain available to restore any removed vault or vault access covers within One hour of notification from the Shift Supervisor or control room personnel.
Control room personnel will remain cognizant of current and approaching weather at all times while the vault covers are removed. Prior to removal and during periods of time when the covers are removed, the control room will verify current weather and weather forecasts for inclement weather.
Covers will not be removed if thunderstorm warnings or tornado watches are in effect or weather patterns sufficient to issue thunderstorm warnings or tornado watches are expected to enter the area within one hour.
The USAR does not specify any controls on lifting or removing the covers. AP 10-104 does not include any discussion of how to treat the removal of the manhole covers. This change will ensure protective measures are in place to minimize the risk to the safety related structure, system or component.
Loss of Normal Feedwater, Loss of non-emergency A/C, and Feedwater Line Break all assume the use of ESW for supplying the Auxiliary Feedwater System with water to provide long term core cooling. ESW also provides heat removal for safety related equipment (emergency diesels, component Cooling Water system, room coolers, containment coolers, Class 1E air conditioning) during emergency operation of the Emergency Core Cooling System during a LOCA and also provide the means to bring the plant to a safe shutdown condition using only safety related equipment.
No credible accidents would be created by this activity. Removal of the ,
vault or access covers will not create a seismic II/I condition or missile hazard. The outside diameter of each personnel access cover is significantly larger than its respective opening inside diameter, thus, j cover drop failure into the enclosure is not possible. The covers are )
reinforced concrete, and rigging controls limit lift height to a minimum, thus, a dropped cover will not result in fracture and multiple missiles that could enter the opening. Vault covers will be placed aside from the vault openings and attached rigging relaxei to assure the vault covers do not remain suspended. All rigging or hoists that are to remain suspended over the vault (for example, a temporary A-frame with hoists) will be moused or restrained so as to prevent them from posing a threat from unstable equipment.
The proposed activity will not affect the ESW or Emergency Diesel Generator Systems since the compensatory measures in the procedure will maintain the system per the design. No credible malfunctions of equipment important to safety are identified. ;
Since the accidents identified in have been shown to have not been j affected, the margins of safety are not impacted.
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Page 201 of 217 Safety Evaluation: 59 98-0142 Revision: O' )
i Updated Safety Analysis Report Correction for Valve Leakoffs and Equipment Drains .;
This Updated Safety Analysis Report (USAR) change provides for the l following changes. ;
I Item g) under USAR Section 9.3.6.2.1 states that there are valve leakoffs )
and equipment drains outside of containment which go directly to the Boron l Regeneration System (BRS). There are no valve leakoffs or equipment ,
drains which go directly to the BRS. All piping which is seen on )
essential drawing M-12HE01 entering the BRS was inspected and the details ;
of the pipe flow and intended service were verified. In USAR Section l 9.3.6.2.1, the second paragraph is being revised to remove the existing letter g) qualifier and its contents. This is considered an editorial
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change such that the USAR accurately reflects the configuration of the j plant. There are no design bases which state that the BRS must accept i liquid from valve leakoffs or equipment drains. Conversely, there are no I design bases that state that the BRS must not accept liquid from valve I leakoffs or equipment draina. This is only a descriptive statement.
Revising the statement will have no effect on plant operation or on any accident or hazards analysis.
In USAR Section 9.5B.7 there is a description of particular sumps pumped to a drain tank located in the radwaste building. In fact, the sumps are pumped to the Secondary Liquid Waste Oil Interceptor which is located in the Turbine Building. This USAR Sections' last sentence is being revised to read, " Sump pumps drain the sump as depicted on Figure 9.3-5." This is considered an editorial change such that the USAR accurately reflects the configuration of the plant. This will have no effect on the plant operation, nor on any accident or hazards analysis.
In USAR Section 9.5B.7 there is a statement that drains are piped to a sump at elevation 1974. This sump is actually located in a pit within the floor with the bottom of the sump at Elevation 1967. However, access to this sump is 'from' floor elevation 1974. To clarify wording in the USAR, this section is being revised to read, "The drains are piped to a sump
- g. located on floor El. 1974." This will have no effect on plant operation, nor on any accident or hazards analysis.
In USAR 9.5B.7 second paragraph, last sentence, the USAR inaccurately describes the number of drains in rooms 1405 and 1415. Essential drawing M-0P1411 shows Room 1405 having one (1) floor drain while Room 1415 having two (2) drains. This USAR section is being revised to read, "One 4-inch floor drain is provided in Room 1405 and two 4-inch floor drains in Room 1415."
There are no procedures, activities, administrative controls, or sequences i
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Attachment II to ET 99-0003 Page 202 of 217 of plant operations involved with these administrative changes. These ,
USAR chapters do not describe or involve any of the discrepancies to be '
changed in the USAR. There are no accidents these administrative changes could possibly create. These changes do not affect or influence any SSC's ,
important to safety. There are no license bases or Technical ;
Specifications involved with these administrative changes.
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Safety Evaluations' 59 98-0143 Revision 0 [
Temporary Procedure TMP 98-007 {
Temporary Procedure TMP 98-007, " Temporary Power NK12/NK14 Batteries and Distribution," provides instructions for connecting temporary power to l
distribution panels NK42, NK44, and NK54 and batteries NK12 and NK14. The >
temporary configuration will allow train B equipment to be powered in a manner which isolates buses NK02 and NK04. The procedure is limited for i use only when the plant is in Modes 5 or 6 and the opposite train (train ,
A) is operable. (
Updated Safety Analysis Report (USAR) Section 8.3 identifies the overali f design of the 125v DC System. USAR Figure 8.3-6-01, Rev. 10, shows the -
permanent power to Distribution Panels NK42, NK44 and NK54 and batteries i NK12 and NK14. During the B train outage, USAR Section 8.3 and Figure 8.3-6-01, will no longer be accurate, due to maintenance activities. During ;
train outages, the equipment is not necessarily configured in the j permanent manner due to maintenance. This is acceptable since no credit ;
is taken for the. functioning of the train that is out of service. The l configuration for temporary power identified by TMP 98-007 affects a non- !
functioning safety equipment trazn. Therefore, none of the design basis ;
accidents discussed or referenced in UFsR Chapters 2, 3, 6, 9 or 15 are impacted by temporary procedure TMP 90-007.
Temporary Procedure TMP 98-007 involves providing auxiliary power to equipment that is not being relied upon to perform its safety related function due to a train outage. Therefore, there are no credible accidents which can be created by the proposed temporary procedure.
Because the equipment in question is part of a non-functioning safety train, it is not required for safe shut down of the reactor, to maintain the reactor in a safe shutdown condition or to prevent or mitigate the consequences of accidents that could result in potential off-site exposures in excess of radiological limits. Because the equipment in question is out of service, there are no credible malfunctions which may be directly or indirectly affected by TMP 98-007.
Technical Specification requirements are not affected to any additionai extent beyond those items which are normally identified for a safety related train outage. Technical Specification 3.8.2.2 identifies the Mode 5 & 6 requirement for one train of 125v batteries being available.
None of the accidents identified in the USAR are affected by this temporary procedure. The equipment affected by the implementation of Temporary Procedure TMP 98-007.will be out of service at the time of impidmentation. Only one train is required to be operational in modes 5 and 6. Because the affect of having only one train operable in Modes 5 and 6 has been' factored into the existing probabilities, and one train
Attachment II to ET 99-0003 Page 204 of 217 will be operable. Therefore, there is no change to the probability of any accident.
-Since none of the accidents previously evaluated in the USAR are affected by TMP 98-007, the temporary procedure has no affect on the radiological consequences of. accidents previously evaluated in.the USAR.
l The procedure is being_. implemented in conjunction with B train outage, during which time the equipment is assumed to be unavailable. Equipment !
moortant to safety in the affected train is not required and therefore l
.cs malfunction is of_no consequence. Therefore, use of procedure TMP 98- [
007 does not increase the probability of occurrence of a malfunction of )
equipment important to safety previously evaluated in the USAR.
l Equipment malfunctions with this temporary procedure are encompassed by !
existing malfunctions scenarios. Therefore, no increase to the i radiological consequences of a malfunction of equipment important to ]
safety, will be incurred. The use of temporary power to the equipment <
does not create the possibility of an accident of a different type than )
any previously evaluated in tbt USAR.
l Since Temporary Procedure TMP 98-007 will be performed only in Modes 5 or 6,-equipment powered from Distribution. Panels NK42, NK44 and NK54, is not
. required for safe shut down of the reactor. Since the procedure is part of a safety related train outage, the equipment is not required to maintain the reactor in a safe shutdown condition or to prevent or mitigate the consequences of accidents that could result in potential off-site exposures in excess of radiological limits. Therefore, the proposed procedure does not create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the USAR.
Technical Specification 3.8.2.2 requires that at least one train of safety related 125VDC Power be available. If nr.t , all operations involving core alterations, positive reactivity changes or movement of irradiated fuel,-
must be suspended. Since the affected train is already in a "not required" status, there_is no additional effect caused by Temporary Procedure TMP 98- l 007. Therefore, the margin of safety as defined in the technical I specifications is unchanged.
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Page 205 of 217 i I
Safety Evaluation: 59 98-0144 Revision: 0 I
Temporary Procedure 98-008 l,
Temporary Procedure TMP 98-008, " Temporary Power NK11/NK13 Batteries and l Distribution," provides. instructions for connecting temporary power to !
Distribution Panels NK41, NK51 and NK43 and batteries NK11 and NK13. The i temporary configuration will allow train A equipment to be temporarily 4 powered in a manner which_ isolates Buses NK01 and NK03. The procedure is j limited for use only when the plant is in Modes 5 or 6 and the opposite !
train (train B) is operable. !
Updated Safety Analysis Report (USAR) Section 8.3 identifies the overall design of the 125v DC System. USAR Figure 8.3-6-01, shows the permanent ;
power to Distribution Panels NK41, NK51 and NK43 and batteries NK11 and :
NK13. During the train A outage, Section 8.3 and Figure 8.3-6-01, will no i longer be accurate, due to maintenance activities. During a train outage, the equipment is not necessarily configured in the permanent manner due to maintenance. This is acceptable since no credit is taken for the functioning of'the train that is out of service. l The configuration for temporary power identified by TMP 98-008 affects a ,
non-functioning safety equipment train. Therefore, none of the design j basis accidents discussed or referenced in USAR Chapters 2, 3, 6, 9 or 15 are impacted by temporary procedure TMP 98-008. ]
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Temporary Procedure TMP 98-008 involves providing auxiliary power to equipment that is not being relied upon to perform its safety related function due to a train outage. Therefore, there are no credible accidents which can be created by the proposed temporary procedure.
Because the equipment in question is part of a non-functioning safety train, it is not required for safe shut down of the reactor, to maintain the reactor in a safe shutdown condition or to prevent or mitigate the consequences of accidents that could result in potential off-site exposures in excess of radiological limits. Because the equipment in question is out of service, there are no credible malfunctions which may be directly or indirectly affected by TMP 98-008.
Technical Specification requirements are not affected to any additional extent beyond those items which are normally identified for a safety related train outage. Technical Specification 3.8.2.2 identifies the Mode r
5 & 6 requirement for one train of 125v batteries being available.
None of the accidents identified in the USAR are affected by this temporary procedure. The equipment affected by the implementation of Temporary Procedure TMP 98-008 will be out of service at the time of implementation. Only one train is required to be operational in Modes 5 and 6. The affect of having only one train operable in Modes 5 and 6 has i
Attachment II to ET 99-0003 Page 206 of 217 been factored into the existing probabilities, and one train will be operable. Therefore, there is no change to the probability of any accident.
Since none of the accidents previously evaluated in the USAR are affected !
by TMP 98-008, the temporary procedure has no affect on the radiological consequences of accidents previously evaluated in the USAR.
The procedure is being implemented in conjunction with the A train outage, during which time the equipment is assumed to be unavailable. Equipment important to safety in the affected train is not required and therefore its malfunction is of no consequence. Therefore, use of procedure TMP 98-008 does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR.
Equipment malfunctions with this temporary procedure are encompassed by existing malfunctions scenarios. Therefore, no increase to the radiological consequences of a malfunction of equipment important to safety, will be incurred. The use of temporary power to the equipment does not create the possibility of an accident of a different type than any previously evaluated in the USAR.
Since TMP 98-008 will be performed only in Modes 5 or 6, equipment powered from distribution panels NK41, NK51 and NK43, is not required for safe shut down of the reactor. Since the procedure is part of a safety related train outage, the equipment is not required to maintain the reactor in a safe shutdown condition or to prevent or mitigate the consequences of accidents that could result in potential off-site exposures in excess of radiological limits. Therefore, the proposed procedure does not create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the USAR.
1 Technical Specification 3.8.2.2 requires that at least one train of safety related 125VDC Power be available. If not, all operations involving core alterations, positive reactivity changes or movement of irradiated fuel, must be suspended. Since the affected train is already in a "not required" status, there is no additional effect caused by procedure TMP 98-000. Therefore, the margin of safety as defined in the technical specifications is unchanged by Temporary Procedure TMP 98-008. 1 l
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Page 207 of 217 i i
Safety Evaluation: 59 98-0146 Revision 0 !
Independent Safety Review Group Organization Change .
l This is an organization change. This change will move reporting of the ?
Independent Safety Review Group (ISEG) otherwise known as Nuclear Safety Engineering (NSE) from Performan,e Improvement and Assessment to Licensing !
and Correr:tive Action and-more specifically the NSE function will report !
to the Superintendent Licensing (which is a title change from Supervisor Licensing). The ISEG function will have a direct reporting chain to the [
Chief. Executive Of ficer (CEO) . The organization change will replace the i person currently assigned to the position of Manager Performance [
, Improvement and Assessment with the person currently assigned as acting l Manager. (This person was previously evaluated as the person qualified as r the position described in ANSI /ANS 3.1-1978, 4.4.5, Quality Assurance). .
In addition, Supplier / Material Quality will report to the Manager Performance Improvement and Assessment. There will be no direct effect on the normal day to day activities associated to these groups. All i functions will cantinue to be performed.
This change will affect the organization discussed in Chapter 13, the i reporting of Supplier / Material Quality described in Chapter 17 and the [
ISEG discussion in chapter 18 of the USAR. WCNOC is currently committed j to NUREG 0737, I.B.1.2, Organization and Management. This commitment l includes a reporting requirement to a corporate official who holds a high- !
level, technically oriented position that is not in the management chain f for power production. This will be maintained with the direct reporting i ability to the CEO. WCNOC is also committed to ANSI /ANS 3.1 - 1978 for the l qualifications of the Manager Performance Improvement and Assessment.
This commitment will also continue to be met. j l
This is an organization change. It has no effect on any design basis j accidents discussed in the USAR. No functions are being deleted.
This is a change in organization - (title, personnel and reporting) of :
organizations which do not perform approve cecident analysis or effect s plant operations. Independence from plant operations is being i maintained. Therefore, the change will not create any new accidents.
This is an organization change that does not affect plant operations or I equipment operation. Therefore, there are no direct or indirect affects ;
on equipment important to safety. ;
1 The discussion in Technical Specifications associated to ISEG would be !
supported by the discussion in Chapter 18 and in the SER. However, the )
commitments made to NUREG 0737 are being maintained. Therefore, the ;
acceptance limits (or in this case more appropriately acceptance criteria) !
are still being met. ,
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Safety Evaluation: 59 98-0147 Revision 0 l l
Control of Heavy Loads Clarification The proposed changes being evaluated to WCNOC-4, " Report on Control of l Heavy Loads," and the Updated Safety Analysis Report (USAR) statement are I to clarify that the polar crane load blocks / hooks and load cell linkage 5 are not to be treated as heavy loads. In addition to this clarification, the proposed change to WCNOC-4 will also include the new allowed. condition to reverse hook orientation if necessary by trollying over the open vessel with fuel in it with or without the load cell linkage attached. In the i following discussion, analysis will refer to the original heavy loads report and should not be considered or construed with the USAR analysis or ,
USAR previously evaluated consequences, malfunctions or accidents. When referring to the USAR analysis or USAR previously evaluated consequences, malfunctions or accidents the acronym "USAR" will be utilized. ,
I The containment building polar crane is used in conjunction with the various lifting rigs to remove the reactor vessel head, vessel upper and i lower internals. The 260-ton capacity main hook is used for these services. The 25-ton-capacity auxiliary hook on the polar crane is used for routine maintenance and in-service inspection. Each hook has an asse;ssted block or pulley assembly for the cables. The main hook weighs l 2100 lbs and its block weighs slightly over 5000 lbs. 'The estimated i weights'of the auxiliary hook and its block are 400 and 2500 lbs I respectively. A load cell linkage is connected between the main hoist i hook and the reactor head or internals rig to monitor the load during lifting and lowering to ensure no excessive loading is occurring. It weighs 2750 pounds and has a rated capacity of 175 tons. '
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The original. report or analysis on the control of heavy loads, now known ,
as WCNOC-4, does not definitively discuss the polar crane load ;
blocks / hooks and load linkage cell as being treated as heavy loads. I currently these polar crane features are treated by the heavy loads l procedure as heavy loads by default. NUREG-0612, " Control of Heavy Loads (
at Nuclear Power Plants," defines a heavy load as "Any load, carried in a :
given area after a plant becomes operational, that weighs more than the ;
combined weight of a single spent fuel assembly and its associated '
' handling tool for the specific plant in question." At Wolf Creek this heavy load definition translates to approximately 2000 lbs. The ;
guidelines in NUREG-0612 Section 5.1.3 mention that the load block may be moved.over fuel in the reactor when handling smaller loads or no load at ~
all. It continues to say in this situation that, due to the weight of the load block alone, this should also be considered as a heavy load. The '
guideline in this section thus does not explicitly state that the load !
block is to be considered a heavy load. There exists no specific analysis of the load blocks / hooks or load cell linkage as heavy loads in our {
analysis or submittals to the NRC in regards to heavy loads. It can be inferred from the original analysis submitted to the NRC that these polar j
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Attachment II to ET 99-0003 i
Page 209 of 217 crane features were not considered as heavy loads, or not considered at >
, all, as seems the case for.the load linkage cell. This statement is based ;
on the fact that the analysis never restricts the movement of.the polar
~
crane when unloaded or loaded with a light load in Modes 3 or 4, thus inferring that the load blocks and hooks are not to be considered as a ,
heavy load. The analysis does not mention the load cell linkage. The !
analysis does say that "Once the reactor vessel head and upper internals are removed, the polar crane will be administratively controlled to !
preclude travel over the open vessel while fuel is in the reactor except l
for required vessel servicing operations such as irradiation sample i removal". The original analysis was reviewed by the NRC, who concurred with it and stated that it complies with the guidelines of NUREG-0612.
This conclusion is documented in NUREG 0881 Safety Evaluation Report (SER) for Wolf Creek Generating Station, Supplement 5. Nothing in thie SER !
conflicts with the conclusions just discussed or provides any more clarification of these polar crane features, j Not having to treatof the polar crane blocks / hooks and load cell linkage features as loads will enable movement of them over restricted areas.
This will allow 360 degree movement and complete containment area access i of the polar crane when not lifting heavy loads or no load at all. It will also enable trollying (no hook / block manipulation) over the open l reactor vessel with fuel in it, when necessary to reverse hook !
orientation. This ability allows flexibility in the orientation of a hook i (main or auxiliary) that needs to be next to the containment wall for a !
desired load lift.
The auxiliary hook, due to its size, can get closer to the. wall than the ,
main hook. When lifting the reactor head, the main hook has to be on the i outside or next to the circular containment wall. If, after lifting the i reactor head, the auxiliary hook is desired on the outside, it can not be oriented that way without traveling over the open reactor vessel. The i proposed change will allow the ability to reorient the hooks. The affects !
of this change do not result in a compromise of the decay heat removal ;
function, create the potential to crush the fuel such that criticality ;
could be challenged, or cause 10 CFR 100 limits to be exceeded. The basis for this statement follows.
I Stress calculations were made on the major components of the containment building polar crane. Stresses on structural load-carrying components are in compliance with requirements. The seismic analysis performed on the containment building polar crane has considered Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) for unloaded conditions and safe shutdown for earthquake for loaded conditions up to 200 tons. The stresses were found to be in compliance with OBE and SSE seismic !
requirements. Positive means are also provided to limit motion of the polar crane trolley during a seismic event. Trolley earthquake restraints l are provided to limit vertical motion of the trolley. These restraints j are attached to both sides of the trolley girders and project under the I flanges supporting the rails on which the trolley runs. To help limit i
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E Attachment II to ET 99-0003 Page 210 of 217 horizontal motion of the trolley during a seismic event, rail capture bars are provided. Thus, the containment building polar crane is designed to maintain its structural integrity with load during an SSE.
The load cell linkage complies with the guidelines of NUREG-0612 and ANSI N14.6 for "special lifting devices." A retainer plate which fits into a "U" slot on the pin and is bolted to the linkage side plates, prevents the load cell pin from falling out of the linkage. The drop of the load cell linkage when attached to the main hoist hook without any load is not considered credible.
Redundant and independent limit switches provided on the polar crane hoists ensures that two-blocking accidents are prevented. The two types of hoist upper limit switches, geared and weight-operated, are redundant and independent. This is to say that if the geared-type limit switch were to fail the weight-operated limit switch would stop the load block from rising higher and would prevent the occurrence of a two-blocking event that are the major cause of load drop events. Pre-operational inspections include the main and auxiliary hoists wire rope and frequent demonstrations that the primary upper limit switch is operable. The stresses in the hoist system are extremely low due to the self-weight of the load block / hook / linkage compared to the capacity of the crane. In addition, when there is fuel in the reactor vessel, use of hoist controls over the open reactor vessel will not be allowed other than those already allowed by the analysis for servicing the vessel. These restrictions on the hoist controls will eliminate any increase in the potential to lower or drop the blocks / hooks / load cell linkage into the open vessel or create a two-blocking event that could result in a drop event into the vessel I when trollying cver the vessel to reorient the hooks. Therefore the potential for the occurrence of a drop of the blocks / hooks / load cell linkage into the vessel when reorienting the hooks as a result of a two- ,
blocking event will not be created by the change. The potential of these items to drop as a result of their disintegration when they are trollying over the open vessel remains within the acceptable limits already permitted in the analysis that clearly allows the blocks / hooks over the open vessel for servicing it. Heavy load drops of these items into the open vessel with fuel in it were not analyzed because it is postulated that these items were not considered as heavy loads, were considered as integral parts of the crane and were only allowed over the open vessel under certain conditions for brief time periods as necessary. Therefore, criticality and damaged core calculations are not required and are still not required as a result of the minor change allowed by this revision to the analysis.
The text changes to USAR Section 9.1 clarify that the polar crane blocks / hooks and load cell linkage are not to be treated as heavy loads.
This text change will provide consistency between the analysis and the USAR. The affects of this USAR change will only help the reader of the USAR with a clearer understanding of the material presented.
Attachment II to ET 99-0003 Page 211 of 217 In conclusion, the affect of the subject changes to the WCNOC-4 and USAR documents are summarized as clarification type changes to the existing analysis and the USAR. These changes will have no adverse affect to the '
safe operation of the plant.
The proposed change to the USAR text clarifies the treatment of the polar crane load blocks / hooks and load cell linkage. This clarification to the USAR text does not conflict with any other USAR statements, descriptions, controls, procedures, et cetera. The USAR text sections that describes the polar crane and the reactor vessel head drop accident are not affected by the proposed change. The proposed change to the heavy loads report does not affect the conclusions of the Licensing Basis contained in Supplement #5 to the Wolf Creek SER. The conclusion being that it complies with the guidelines of NUREG-0612.
The changes to the analysis and the USAR are associated with the polar crane usage in the containment building. In USAR Section 9.1 the reactor vessel head drop is evaluated. This heavy load drop accident has not been impacted by the subject changes as it does not alter any conditions or assumptions of this accident. The dropping of a fuel assembly in the containment building evaluated in USAR Section 15.7.4 has been reviewed.
These USAR accidents are not impacted by this change. The review of USAR Chapters 2, 3, 6, 9, and 15 has concluded that no USAR accidents have been !
impacted as a result of the proposed changes. :
I The clarification of the treatment of the polar crane blocks / hooks and l load cell linkage is not a change to the analysis or USAR because the '
original analysis inferred no restrictions on these polar crane features except over the open reactor vessel with fuel in it. This clarification does not create any new or unique accidents because it simply clarifies the treatment of these polar crane features in the analysis and USARs The change to the analysis to enable hook reorientation by trollying over the open vessel with fuel in it with or without the load cell linkage attached is viewed as an infrequent action. This action remains bounded within the acceptable limits of the original analysis that allowed for the blocks / hooks to be over the open vessel with fuel in it for needed reactor vessel servicing activities. These servicing activities include manipulation of the hooks that the change does not allow when reorienting I the hooks, if needed.
It is not credible to assume a drop of the polar crane blocks / hooks and load cell linkage due to their design, structural integrity, inspections !
and testing requirements. The potential to damage the fuel or potentially l crush it into a configuration such that Keff could become larger than 0.95 by failure of the polar crane load blocks / hooks and or load cell is not a credible accident to consider. Thus, the fuel assembles are not affected by the change and will retain their integrity, ensuring no release of fission product gases. Thus, no credible accidents have been created as a result of the changes.
Attachment II to ET 99-0003 Page 212 of 217 i
The polar crane, its blocks / hooks, and load cell linkage are important to ,
safety. Their use and control remains as designed and are not subject to any new or unique conditions as a result of the changes. The polar crane and its hooks are controlled from its' bridge-mounted cab, a portable cab, i or a portable radio control unit. These control features are not affected by the subject changes. It is concluded that no credible malfunctions can ,
be identified as a result of the proposed changes.
Wolf Creek Generating Station Technical Specification 3/4.9, " Refueling ,
Operations" and Technical Requirements Manual Section 16.9, " Refueling Operations," were reviewed and no margins were observed as affected by the changes.
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Page 213 of 217 ;
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Safety Evaluation: 59 98-0148 Revision 0 '
Radiological Emergency Response Plan Revision to Convert to Administrative Procedure Format -
This Unreviewed Safety Question determination provides an evaluation of ;
changes to the Wolf Creek Generating Station " Radiological Emergency Response Plan (RERP) , " now AP 06-002, Revision O. The RERP has been i reformatted into the Wolf Creek Generating Station (WCGS) standard j procedure numbering system with this revision. The previous Emergency i Action Levels (EAL) forms are superseded by the new EAL form, APF 06-002- !
01, Revision O. l r
Changes to the RERP are summarized below. These changes provide for a ;
more efficient implementation of the RERP by the emergency response !
organization. Certain of these changes were determined by Wolf Creek to f require NRC approval prior to their implementation. These changes were submitted to the NRC by letter dated September 18, 1998, and subsequently l approved by the NRC by letter dated November 17, 1998.
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The RERP has been reformatted, consistent with the format requirements of j the Wolf Creek Generating Station, " Procedure Writers Guide," AP 15C-004. i The previous RERP has Leen superseded by procedure AP 06-002, I
" Radiological Emergency Response Plan," (RERP), Revision O. Information provided in several sections of the previous RERP has been consolidated into the new RERP. The generic summaries and overviews have been deleted. The new RERP format contains a Table of Contents, Attachments,
-Figures, and procedure section for contents, which follow the recom;..endations of NUREG 0654.
The major. changes to the content of the RERP are as follows:
Position titles changed throughout the RERP for new Emergency Response Organization (ERO) implementation. The title changes better define the function for each position when discussing events with off-site agencies. j Positions performing the same functions in the Technical Support Center i (TSC) and Emergency Operations Facility (EOF) have been given the same title and will be cross trained to allow personnel to fill positions in either facility.
Certain functions were moved to positions better suited to perform the functions. The required functions are still being performed, and the l changes will make the organization more efficient in their implementation I of the RERP.
l The Emergency Classification level for activation of the EOF was changed from "SAE or higher" to " Alert or higher." Concurrent activation of the facilities eliminates concerns with deployment of teams and dose assessment, and eliminates the need for duplication of certain functions. )
For example, the Emergency Notifications System (ENS) Communicator '
position at the EOF was eliminated, since the redundant position remains l
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Page 214 of 217 l
i in the TSC, the TSC facility is activated at the same time as the EOF, and i plant related data originates in the TSC. Similarly, the Heath Physics l Network (HPN)' communicator position located in the TSC was eliminated, !
since the radiological data processing originates with the EOF, and the EOF HPN position is activated concurrently. The functions of these two ,
communicator positions are being performed as recommended in NUREG 0654 l
and are located in the facilities that can best provide necessary j information to the NRC in the event of an emergency.
Dose Assessment has been moved from the TSC to the LOF because the off-site monitoring is performed and controlled from that facility. The Dose Assessment function is started and performed from the EOF. This change reduces the number of persons required for the ERO specific positions and ,
allows for.better use of experienced persons in the locations needed for !
other radiation monitoring.
The title of Wolf Creek Lake (WCL) was changed to Coffey County Lake (CCL) in the RERP. )
i Vice President Plant Operations was changed to Plant Manager in the RERP '
due to earlier organization changes. This change provides consistency between the RERP and the Updated Safety Analysis Report. !
l Emergency Action Levels (EALs) have been relocated to form APF 06-002-01, Revision 0, associated to the RERP. To assure that proper regulatory ;
evaluations continue, a commitment has been included in the RERP ,
commitment.section to require an Unreviewed Safety Question Determination !
and a 50.54 (q) evaluation to be performed for each change to the EAL form. l The EAL flow charts, Form APF 06-002-01, have four minor changes: )
Box 1-RER5 deleted a reference to the Nuclear Plant Information System 'j display and changed from information located on a form to information located in the EAL bases, l
- Box 2-SGTF9, changed the "NO result".from '"No Action This Category" to )
"NUS." This' change' ensures a small Steam Generator (S/G) tube leak will I be declared on the SGTR chart. !
Box 3-LRCB1, last asterisk added "(except S/G tube leakage)." The !
statement is added to separate S/G tube leakage and Reactor Coolant (RCS) leakage, as S/G leakage should be declared on the SGTR chart.
Box 10-FR2 added " Hot Machine Shop." The Hot Machine Shop is added because it is connected to the Auxiliary Building, and the NUMARC EAL bases states connected buildings should be identified in the EAL chart.
None of these changes reduce the effectiveness of the EALs. The Bases part of the form was revised to reflect procedure number changes as a result of writer's guide requirements and ERO title changes due to new ERO implementation.
Technical Logkeepers and Communicators titles were changed to Administrative Assistants.
Attachment II to ET 99-0003 Page 215 of 217 The Wolf Creek Spokesperson title was changed to the Pubic Information Officer (PIO), and the former PIO position was changed to Public Information Manager (PIM). The new PIO has overall responsibility for the Public Information Organization and the PIM ensures the facilities are ,
activated and function in accordance with procedure.
This revision did not change the wording of the four non-delegatable duties of the Emergency Managers. The Emergency Managers may not delegate the duties; however, these duties may be divided between the Emergency Managers.
Attachment II to ET 99-0003 ,
Page 216 of 217 [
I Safety Evaluation: 59 98-0158 Revision 0 i
Essential Service Water Piping to Containment Coolers ;
This change to the Updated Safety Analysis Report (USAR) revises USAR !
Tables 6.2.1-6, 6.2.1-7, 6.2.1-59, and 6.2.1-60 to document information !
discovered in the evaluation of Generic Letter 96-06, " Assurance of ,
Equipment' Operability and Containment Integrity During Design-Basis !
Accident Conditions." !
In the evaluation of the events described in Generic Letter, it was ;
discovered that the ESW piping to the containment coolers may not !
completely refill and pressurize until 65 seconds following a design basis !
accident (DBA) coincident with loss of offsite power. The calculation for ;
peak containment temperature and pressure originally assumed that the containment fan coolers would start removing heat from the containment at ,
60 seconds following the initiating event. While this assumption would ;
probably still be valid, since heat would be removed while the piping system is filling, the exact amount of heat removal would be difficult to !
determine. Therefore, the calculation was revised, using 70 seconds as I the assumption of the time delay to start of heat removal by the ,
containment coolers. This change of the time delay assumption needs to be !
reflected in the USAR.
The results of the revised calculation show that there is negligible difference in the peak temperature and pressure of the containment by assuming heat removal at 70 seconds versus 60 seconds. For this reason, new traces of containment precsure and temperature will not be generated. ;
In addition, since there was negligible change in the limiting cases of both LOCA and MSLB, other cases were not re-calculated using'the longer time.
There are no changes to the results of any calculatione or evaluations as a result of assuming 70 seconds time delay for the containment coolers to begin removing heat from the containment. Therefore, there are no procedures, activities, administrative controls or sequences of plant operation which are affected. . Tables 6.2.1-6, 6.2.1-7, 6.2.1-59, and 6.2.1-60 detail the assumptions that go into the calculation for peak containment pressure and temperature following a DBA. One of the assumptions has changed, and that necas identified in the tables. The results of the calculation did not change.
There are no new types of credible accidents that the new assumption can create. There is no change to the plant or to operation of the plant.
The results of the analyses remain the same. This is only a change to one assumption which is input to the calculation for peak containment pressure and temperature following a DBA.
There are no credible malfunctions of equipment important to safety which
w Attachment II to ET 99-0003 Page 217 of 217 l
are affected. While it has been identified that no credit can be taken for containment cooler heat removal for 70 seconds, rather than 60 seconds, the containment coolers will still function. Peak containment pressure and temperature are negligibly affected by the extra time delay ,
assumed in the revised calculation. Containment integrity is not affected. l
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The only acceptance limit found which is applicable is safety design basis ,
number seven of USAR Section 6.2.2.2.1.1. This basis states that the i Containment Cooling System (CtCS), in conjunction with the Containment i Spray System, is capable of removing sufficient heat energy and subsequent !
decay heat from the containment atmosphere following the LOCA or MSLB !
accident to maintain the containment pressure below design values, USAR ,
Table 6.2.1-2 gives the design value for containment pressure as 60 psig. '
Calculation SA-90-067 and USAR Table 6.2.1-2 show that the peak calculated
_ pressure following a MSLB is 48.9 psig. Calculation SA-90-064 and USAR :
Table 6.2.1-2 show that the peak calculated pressure following a LOCA is [
47.3 psig. These values are based on the assumption that the containment ;
coolers begin heat removal at 60 seconds. Wolf Creek Safety Analys...e has !
recalculated the containment pressure and temperature for the limiting l cases of Large Break LOCA and MSLB, considering the additional delay time ]
before the containment coolers begin to remove heat (Calculations AN !
004 and 005). The results show that the peak calculated containment ;
pressure is negligibly affected. This is explained by the fact that the l peak calculated containment pressure is controlled mainly by the containment spray system, and the containment coolers have minimal ,
effect. The results show that the CtCS will still meet this Safety Design 1 Basis. ,
I Based on the above discussion, this revision will not increase the '
probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. This revision does not create a possibility for an accident or malfunction of a differ (nt type than any evaluated previously in the safety analysis report. The margin of safety, as l defined in technical specifications, is not reduced by this revision.
Therefore, this revision does not involve any unreviewed safety question.
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Attachment III to ET 99-0003 Page 1 of 1 1.IST OF COMMITMENTS The following table identifies those actions committed to by Wolf Creek Nuclear Operating Corporation (WCNOC) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be commitments. Please direct questions regarding these commitments to Mr.
Michael J. Angus, Manager Licensing and Corrective Action, at Wolf Crets Generating Station, (316) 364-4077 COMMITMENT Due Date/ Event None N/A 0
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