ML20213D576
| ML20213D576 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/07/1986 |
| From: | Koester G KANSAS GAS & ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20213D580 | List: |
| References | |
| KMLNRC-86-213, NUDOCS 8611120149 | |
| Download: ML20213D576 (14) | |
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8 4 KANSAS GAS AND ELECTRIC COMPANY T>E ELECTAsC COPAPANY OLENN L MOESTER WeCE PHESsOENT. NUCLEAR Novenber 7,1986 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comnission Washington, D.C. 20555 KMLNRC 86-213 Re: Docket No. STN 50-482 Subj Revision to Technical Specifications 3.5.1 and 3.5.2
Dear Mr. Denton:
he girpose of this letter is to transmit three original and 40 conformed copies of an amendnent to Facility Operating License No.
NPP-42 for Wolf Creek Generating Station, Unit No. 1. Se application for amendnent revises Wolf Creek Generating Station, Unit No. 1, Technical Specifications 3.5.1 and 3.5.2. % e proposed changes to the Technical Specifications are provided in Attachnent III. A con 1plete Safety Evaluation and Significant Hazards Consideration are provided in Attachnent I and II respectively.
In accordance with 10 CFR 50.91, a copy of the atplication with attachnents is being provided to the designated Kansas State Official. Enclosed is a check for the $150.00 application fee required by 10 CFR 170.21.
We proposed nodification of the Wolf Creek Generating Station Technical Specifications will be fully implemented within 30 days of formal Nuclear Regulatory Comnission approval.
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201 N. Market - WicMa, Kansas - Mail Address: Po. Box 208 i WicMa, Kansas 67201 - Telephone: Area Code (316) 2616451
Mr. H. R. Denton November 7,1986 10DRC 86-213 Page 2 1
If you have any questions coteerning this matter, please contact me or Mr. O. L. Maynard of my staff.
Very truly yours, h )
Glenn L. Koester b Vice President - Nuclear GLK wbb Enclosure Attachnents: I-Safety Evaluation II-Significant Hazards Consideration III-Proposed Technical Specification dianges oc P0'Connor JCununins GAllen FJohnson
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STATE OF KANSAS )
) SS CITY OF WICHITA )
s I, John A. Bailey, of lawful age, being first duly sworn upon oath, do depose, state and affirm that I am Director Engineering and Technical Services of Kansas Gas and Electric Company, Wichita, Kansas, that I have i signed the foregoing letter of transmittal for Glenn L. Koester, Vice President -
Nuclear of Kansas Gas and Electric Company, know the content thereof; and that all statements contained therein are true.
By John A. Bailey /
Director Engindering and Technical Services SUBSCRIBED and sworn to before me this y 4 day of< , , 1986. g.9 gh . , . , :
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s ATTACHMENT I .
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Mr. H. R. Denton Novenber 7, 1986 Attachment I to IU D BC 86-213 Page 1 i
l Safety Evaluation A change in the aligmaant of the anergency Core Cooling System (E03) has been proposed for the Wolf Creek Generating Station (NOGS) during ,
Reactor Coolant System (RCS) heatup and pressurization after '
terminating Residual Heat Removal (RHR) heat renoval operations in l l Mode 3. The change would permit closing of valve EJ HV-880% an4/or FJ HV-8809B in the lines from the disdarge of the RHR pmps to the BCS l cold legs. With either valve 880 % or 8809B closed, flow would be ,
isolated from the RRR paps to two of the four RCS cold legs.. With ,
both valves 8809A and 8809B closed, flow would be isolated from the RHR pumps to all four RG cold legs. The closing of these valves would not affect the safety injection flow from the centrifugal charging pays or safety injection pungs. The Technical Specifications require two F subsystems of EOCS pays to be operational in Mode 3. Safety injection flow would be available from at least one centrifugal darging puup and one safety injection pump following a safety injection signal, even with both valves 880 m and 8809B closed and assuming a single failure.
If only one of the valves is closed, safety injection flow from at ,
least one RHR pap would also be available for injection into two of '
the RG cold . legs. It is noted that valves 880 R and 8809B can be reopened by a single manual action from the control room to restore l
, full safety injection flow from the RHR pumps. For RG pressures above 1000 peig, the accumulators would also. be available for injection in the event of a IDCA.
Another pK W 3 change in the EOCS alignment is also being considered during the Ras heatup and pressurization in lede 3 when the pressure is above 1000 peig. This change would permit closing of one of the
===$1ator discharge isolation valves EP HV-8808A, B, C, or D. With !
one of the valves closed, acctmulator injection would be limited to three out of the four a m =ilators. The closure of valves 880R or 8809B in the RHR discharge lines discussed above would not be performed coincident with the closure of the a m =$1ator discharge isolation valve. Since two couplete subsystems of EOCS pungs are available in i
Mode 3, safety injection flow would be available from at least one centrifugal charging pung, one safety injection pmp, and one RHR punp, and also from three sectr 11ators. The closed accueulator isolation l
valve could also be opened by manual action from the control room. {
An evaluation has been performed to assess the acceptability of l isolating valves 880% an4/or 8809B in the RRR pmp discharge L (1)
I lines and (2) isolating one of the accumulator discharge isolation i v'.lves 8808A, B, C, or D during RG heatup and pressurization operations as described above.
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i Mr. H. R. Denton November 7,1986 l Attaciunent I to 10f2RC 86-213 Page 2 -l I
It has been determined that low pressure shutdown and startup operating conditions are so far below the conditions for which the RG has teen designed that. a large IOCA is not credible and for all practical purposes can be asstaned not to . occur. With the equipnent status described above, it is concluded _that operator actions can be taken for a credible IOCA to avoid exceeding the E03 performance criteria. The supporting information for these statenents is presented below.
A rupture in the RG pressure boundary piping greater than 6 inches in-nominal diameter is considered to be highly unlikely even at normal operating pressure. Engineering studies and operating experience have shown that through wall cradts in the RG dana 1 pressure boundary i piping greater than 6 inches in nominal pipe diamater are highly unlikely. A leak-before-break analysis for the RCS loop piping has been performed for the NOGS and approved by the Nuclear Regulatory Comunission. In addition, leak-before-break analyses have been performed for the RG pressure boundary piping down to 10 inches in nominal pipe diameter for the McGuire and Catawba plants. It is expected that a
, similar analysis for the RG pressure boundary piping down to 10 inches in diamater for the NOGS would yield results comparable to those for the McGuire and Catawba plants.
- The results of the leak-before-break analyses demonstrated that even if a through wall crack is postulated at normal operating pressure, ,
i RCS pressure boundary leakage would be detected with existing leak '
detection systems and the crack will remain stable (i.e.,not propagate to a pipe _ rupture) . The mart == size leak which could occur in the piping down to 10 inches in diameter without being detected less than 1 inch in equivalent diameter) .
would be very small (i.e.,
Since there is no 8 inch nominal diamater piping in the RG pressure boundary for the NOGS, the next smaller piping size to be considered is
- 6 inches in nominal diameter (5.187 inches inside diameter) . Thus, the
=a vi= = size pipe which will not be subjected to a leak-before-break analysis- is the 6 inch piping. If a rupture of the 6 inch piping is asstuned to occur, the maxbmsn size IOCA would be 5.187 inches in 1 diameter.
Below the RG normal operating pressure, a rupture in the RG pressure boundary piping greater than 6 inches in nominal diamater is considered even more unlikely. Normal operation at 2000-2250 psig serves as a more severe condition which demonstrates that pipe ruptures below normal operating pressures are highly unlikely since additional margins m -- . - - , , , - , , , , , , . , , , - - - - ,e..r-- ~-.vr . _ . , - - .
Mr. H. R. Denton November 7, 1986 Attachment I to KMIJEC 86-213 Page 3 of safety exist at the lower pressures. 'the condition which could lead to a pipe rupture, a large through wall crack, would be identified during operation._ However, even with the presence of such a crack, the piping systen would remain stable and a piping rupture would be unlikely at the reduced RG pressure. 'therefore, based on the above information, it is concluded that the maxista credible IOCA to be considered for the RG pressure boundary during shutdown operations is 5.187 inches in diameter, corresponding to the rupture of a 6 inch PiPe.
Por a credible IOCA, the RG break flow rate and depressurization rate is significantly less than for a design basis large break IOCA. Por startup conditions, the break flow and depressurization rates would be further reduced due to the lower initial RG pressure and temperature.
In addition, the initial _ fuel rod tenperatures and decay heat level would be significantly less than for full power since the reactor would have been shut down for a period of time. Automatic Safety Injection (SI) actuation may not occur since the pressurizer pressure SI signal -
is blocked and the high contairrait pressure SI signal may not be operable or may not-be reached for lower initial RCS temperatures.
Operator action would be relied upon to initiate sufficient SI flow to maintain sufficient reactor vessel inventory for adequate core cooling.
'the indications available to the operator that there is a small IOCA in ,
progress would be the following:
- 1. Ioss of pressurizer level ;
- 2. Decrease of RCS pressure j
- 3. Ioss of RG subcooling
- 4. Radiation alarms inside conhi m t i Other potential indications include an increase in contairunent pressure and stmp water level. However, the reduced break flow rate and reduced energy in the break flow for small breaks at low initial RCS temperatures may not noticeably increase the containment pressure.
If a credible IOCA should occur during the RG heatup and pressurization operations in Mode 3, the operators -would be relied
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upon to manually initiate SI. In Mode 3, safety injection flow would be available initially from at least one centrifugal charging planp and one
Mr. H. R. Denton November 7,1986 Attactunent I to KMDIRC 86-213 Page 4 safety injection ptmp. When the RG pressure is greater than 1000 peig, injection flow would also be provided by the acctmulators without any operator action. For the case when one of the acetunulator discharge isolation valves 8808A, B, C, or D is closed in Mode 3, with '
the RHR ptmp discharge valves 8809A and 8809B open, safety injection flow would be available from three of the four a m - dators without !
operator action and from at least one centrifugal charging pung, one l safety injection pump, and one RRR ptmp following SI actuation. 'Ibe '
fourth acctmulator could also be made available by operator action from the control room to open the closed am=Bator isolation valve.
Based. on a shutdown IOCA scoping study for a typical Westinghouse 3411 PMt four-loop plant at two hours after. reactor shutdown, it is concluded that at least 20 minutes are available to initiate SI from a centrifugal charging ptmp to prevent significant oore uncovery for breaks up to three inches in diameter. Initiation of SI flow within this time may not preclude core uncovery, but it is expected to limit the fuel cladding heatup to less than the full power small break IOCA results provided in the FSAR. For breaks larger than three inches and up to six inches in diameter, operator action to initiate SI from a centrifugal charging punp is estimated to be required within approximately 10 minutes. Additional operator action may be required, t depending upon break size, within one hour of the event initiation to start an additional centrifugal charging pung or safety injection ptmp or to depressurize the RCS using the steam generators and to initiate '
i flow from a RHR pimp.
For the proposed changes in the ECCS aligriment during RCS heatup and pressurization in Mode 3 for WOGS, the available injection capability
- of the EOCS plunps would exceed the flow from one centrifugal charging
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ptmp which was used for the scoping study. It should be noted that the scoping study was based on the IOCA occurring at two hours after shutdown, whereas the proposed change for WOGS would be applicable during startup. If a IOCA should occur during startup, the initial fuel rod tenperature and decay heat levels would be less than those in the scoping study. 'Ihis would result in longer times before operator action is required to initiate safety injection than the times
- determined from the scoping study. After initiation of safety injection, the operators can open the affected valves by manual action '
from the control room to restore the E03 configuration to normal.
Pollowup operator actions can also be taken to provide additional safety injection capability if required.
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0 Mr. H. R. Denton Novenber 7,1986 3 Attachment I to IG D BC 86-213 Page 5
- h indications noted previously will alert the operators to a IOCA so that they can perform the required manual actions. It is ===+ad that the operators can cmplete the manual actions to actuate safety injection and restore the E03 ar=d==9t to operable status as required such that the level of protection for a IOCA & ring RG heatup and pressurization in Mode 3 with the proposed EOCS configurations will be equivalent to that during power operation.
An evaluation was also performed to assess the Japact of the proposed E03 aligriment changes on the steam generator tube rupture (SGTR) and other non40CA accident analyses presented in the FSAR. ;'Jhe closing of valves 8809A an4/or B in the RHR ptmp discharge lines would not affect the SGTR or other non-IOCA accident. analyses since the delivery of SI
' flow fran the RRR ptmps is not modeled in the analyses. Amumlator .
injection'also does not occur in the SGIR and non-IOCA accident analyses with the exception ~of the hypothetical steamline'. break accident, and accisnulator injection only occurs after the limiting accident condition is predicted for this case, h refore, the closing of a = =dator isolation valve 88084, B, C, .or D would not inpact the SGTR or non-IOCA accident analysis conclusions presented in the FSAR.
9 *==y =nt to formal approval of the pK--wl anendnent but prior to. ,
inplementation at WOGS, NGEE will assure tint the appropriate eriergency proce&res are available and ~ that adarymte. training is provided to i
support the above statement on the operators ability to perform the-required actions.
'Ihis Amendnent request revises the existing Wolf Creek Generating Station Technical Specification requirements concerning the anergency Core Cooling Systen. 'Ibe proposed Amendnent would allow a change in the aligrunent of the E03 during RCS heatup and pressurization in EDE - 3. Based on the aforementioned discussion, NG&E has concluded that the proposed revisions to Technical Specifications 3.5.1 and 3.5.2 do not adversely affect or endanger the health or safety of the general public or involve an unreviewed safety question.
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.e ATTACHMEN? II a
Mr. H. R. Denton Noven1ber 7,1986 Attachment II to 100mc 86-213 Page 1 SIGNIFICANT HAZARDS (IMSIIERATICE h proposed change to Technical Specification 3.5.1.a would allow closure of one E005 a m =mlator isolation valve in MODE 3 above 1000 psig: during startup, while performing Surveillance Requirement l
4.4.6.2.2. This will only be done providing neither FJ HV-880% anWor - '
B are closed and the closed isolation valve is capable of being reopened. The proposed change to Technical Specification 3.5.2.e would
-allow closure of FJ HV-8809A anWor B in MODE 3 during startup, while performing Surveillance Requirement 4.4.6.2.2. _This will only be done providing the closed valve (s) is (are) capable of being reopened and l pressurizer pressure.is below 1000 psig , or if above 1000 peig, no '
E03 acctmulator isolation valve is closed. ;
h Kansas Gas and Electric Company (IOG&E) has determined that this proposed change to- Technical Specification 3.5.1 does not involve a significant hazards consideration because operation of Wolf Creek Generating Station in accordance with this change would not: !
- 1) Involve a_ significant increase in the- probability or consequences of an accident previously evaluated. It has been
. determined that- low pressure shutdown and startup operating conditions are so far below the conditions for which the RCS has been designed, that a large Ioss of Coolant Accident (IOCA) is not credible and for all' practical purposes can be ,
assumed not to occur. It has been concluded that the mav h = !
credible IOCA to be considered for the RCS pressure boundary during shutdown operation would be a 6 inch pipe break. For a ,
credible IOCA, the Reactor Coolant System (RCS) break flow rate ;
and depressurization rate is significantly less tharr for a '
design basis large break IOCA. Pbr startup conditions, the break flow and depressurization rates would be further reduced ,
due to the lower initial RC3 pressure and temperature. In addition, the initial fuel rod tenperature and decay heat level would be significantly less than for full power since the reactor would have been shutdown for a period of time. With this longer depressurization time and lower decay heat levels, there is anple time available for operator action to open the closed acctmulator valve.
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated. This Technical Specification change pertains to IOCA's and to how much Bnorgency Core Cooling System (E03) flow is available inmediately and after operator action. The possible slight delay in initiating full EOCS flow does not effect any other kind of accident.
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Mr. H. R. Denton November 7, 1986 l Attachment II to KMLNRC 86-213 Page 2 !
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- 3) Involve a significant reduction in a margin of safety. %e - 1 10 CFR 50.59 Safety Evaluation for this Technical Specification amendnent has concluded that the mari== credible IOCA during heatup is the rupture of a 6 inch pipe.. During the l period of time when this valve may.be closed per the proposed i Amen &nent, RG pressure will be above the pressure at which the a m - dators can inject water. % erefore it.is concluded that during the depressurization of the RCS following a IOCA, the operators. will recognize the condition and will be able to reopen the closed acetanulator valw and prevent any significant l fuel heatup. S e valve will be able to be reopened from the main control room & ring this period.
Se Kansas Gas and Electric Coupany-(RGEE) has determined that this prcpceed change to Technical Specification -3.5.2 does not involve a significant hazards consideration because operation of Wolf Creek Generating Station in accordance with this change would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated. . It has been determined that low pressure shutdown and startup operating conditions are so far below-the conditions for which the RCS has been designed, that a large Ioss of Coolant Accident (IOCA) is not credible.and for all practical purposes can be asstuned not to occur. It has been concluded that the maxinun credible IOCA to be considered for the ~R 3 pressure boundary during shutdown operation would be a 6 inch pipe break. For a credible IOCA, the Reactor Coolant System (R3) break flow rate and depressurization rate is significantly less than for a design. basis large break IOCA. For startup conditions, the break flow and depressurization rates would be further reduM5 die to the lower initial RG pressure and temperature. In addition, the initial fuel rod temperature and decay heat level would be significantly less than for full power since the reactor would have been shutdown for a period of time. With
, this longer depressurization time and lower decay heat levels, there is ample time available for operator action to open the closed BHR valve or valves.
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated. Wis Technical Specification change pertains to IOCA's and to how much Emergency Core Cooling System (EOCS) flow is available innediately and after operator action. %e possible slight delay in initiating full EOCS flow does not effect any other kind of accident.
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i Mr. H. R. Denton. November 7, 1986 ,
Attachment II to KMLNRC 86-213 Page 3
- 3) Involve a significant reduction in a margin of safety. The
-10 CPR~ 50.59 Safety Evaluation for this Technical Specification amendnent has concluded that the maximisa credible -
I K A during heatup is the rupture of a 6 inch pipe. During the ;
period of time when FJ HV-880% anWor B may be closed _per the proposed Amendnent, RG pressure _will be above the pressure at which the RRR puqs can inject water. 'therefore it is concluded that during the depressurization of the RCS.following aIRA, the operators will recognize the condition and will be able to reopen the respective closed valve (s) and prevent any significant fuel heatup. '1he valve (s) will be able to be reopened from the main control' room during this period.
During startup, the low pressure safety injection signal is blocked until RCS pressure exceeds 1970 psig. '1herefore should a IEA occur below 1970 peig, operator action would be required
, to initiate any EOCS flow. When this occurs, the operator will-also open any valves that had been closed. If above 1970'psig,-
the . safety injection sigral will be unblocked. Should a IiXA
,- occur,.the safety injection signal will start both centrifugal charging pumps, both safety injection ptmps, both RHR ptsups and '
1 if any a m = 1ator valve is closed, it will automatically open it. Since the RHR pumps cannot inject water until RCS pressure drops to approximately 190 psig, two charging pungs, two safety injection ptmps and four acetanulators will be injecting or will have injected into the core before the RCS has depressurized to-190 peig. This allows adequate time for the operator to open FJ HV-880% an#0r B, before the ptmps are needed to inject. i Based on the above discussions and those presented in Attachment I, it -
'has_ been determined that the requested Technical Specification
! revisions do not involve a significant increase in the probability or consequences of an accident or other adverse condition over previous evaluations; or create the possibility of a new or different kind of accident or condition over previous evaluations; or involve a significant reduction in a naggin of safety. '1herefore the requested license amendnent does not involve a significant hazard consideration.
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ATTACMENT III ~
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