ML20140C083
| ML20140C083 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 01/20/1986 |
| From: | Koester G KANSAS GAS & ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20140C086 | List: |
| References | |
| KMLNRC-86-007, KMLNRC-86-7, NUDOCS 8601270223 | |
| Download: ML20140C083 (16) | |
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KANSAS GAS AND ELECTRIC COMPANY TPE ELECTN COMPANY GLENN L MOESTER
= m.= 1. . u. January 20, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comnission Washington, D.C. 20555 KMLNRC 86-007 Re: Docket No. STN 50-482 Ref: KMLNRC 86-008, from G.L. Koester, KG&E, to H.R. Denton, NNC, dated 1/20/86 Subj: Revision to Technical Specification 4.6.1.2
Dear Mr. Denton:
The purpose of this letter is to transmit three or*ginal and 40 conformed copies of an application for Amendment to Facility Operating License No.
NPF-42 for Wolf Creek Generating Station, Unit No.1.
This application requests that Section 4.6.1.2.d of the Wolf Creek Generating Station Technical Specifications be revised to allow a one-time only deferment of the Type C local leak rate testing (LLRT) for the penetrations delineated in Table 4.6.1.2-1 until startup following Refuel L The proposed changes to the Technical Specifications are provided as Attachment IV.
ni Section 4.6.1.2.d requires performance of LLRTs on an interval of at least S $' once per 24 months which is based on the requirements of 10 CFR 50, Appendix
$@Q- J, Section III.D.3. The Reference provides a request for exemption to the oo two year surveillance interval of 10 CFR 50 Appendix J, Section III.D.3
$$ pursuant to the requirments of 10 CFR 50.12.
Ob In granting a similar request to Carolina Power & Light Company for oO Brunswick Steam Electric Plant, Unit No. 1, the Staff stated that, " The 24 7
month interval requirement for Type B and C penetrations is to be often enough to prevent significant deterioration from occurring and long enough
@$ to permit the LLRTs to be perforned during plant outages. Leak testing of m a- the penetrations during plant shutdown is preferable because of the lower radiation exposures to plant personnel. Moreover, scme penetrations, because of their intended functions, cannot be tested at power operation.
For penetrations that cannot be tested during power operation or those that, AD - J. Knight (1tr only)
EB (BALLARD) l EICSB (ROSA) 4 PSB (CAMMILL) I kI RSB (BERLINGER)
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201 N. Market -Wichits, Kansas - mar Address: RO. Box 208 i Wichita, Kansas 67201 - Telephone: Area Code (316) 261-6451
Mr. H. R. Denton l KMLNRC 86-007 Page 2 Jrnusry 20,-1986 4
if tested during plant operation would cause a degredation in the plants l overall safety (e.g., the closing of a redundant line in a safety systen) , l the increase in confidence of containment integrity following a successful test is not significant enough to justify a plant shutdown specifically to ,
perform the LLRTs within the 24 month time period...".
Refuel 1 is the next scheduled shutdown and is currently scheduled to begin in October, 1986. In toe event of any unanticipated delays in the refueling schedule a unit shutdown to perform these surveillances will be initiated by October 31, 1986. Werefore this request entails an approxbate seven month l If an unschedu?ed shutdown of extension in the most limiting case.
l sufficient duration and resulting in appropriate plant. conoitions occurs I prior to Refuel 1, Kansas Gas and Electric Conpany intends to perform any f surveillances that can be completed. Since refueling outages normally occur approximately every twelve to eighteen months, extensions beyond the 24 month maxinaan interval required by the Technical Specifications and 10 CFR 3
50 Appendix J,Section III.D.3 are not usually necessary. However, due to j
the extended plant startup program and plant performance during Cycle 1, a
Kansa Gas and Electric Company is requesting the aforementioned Technical
! Specification revisions. Without the requested revisions Wolf Creek j Generating Station would be forced into an outage for the purpose of surveillance testing.
W e equipment involved with this extension request' is safety related and highly reliable. The requested revisions -would not have a significant impact on safe operation of the plant when compared to the transients
- associated with a shutdown for the purpose of performing these surveillances. A conplete Safety Evaluation and Significant Hazards i Consideration are provided as Attachments I and III respectively.
In accordance with 10 CFR 50.91, a copy of this application, with Attachments, is being provided to the designated Kansas State Official.
- Enclosed is check (No. 67589) for the $150.00 application fee required by 10 CFR 170.21.
! If you have any questions concerning this matter, please contact me or Mr.
! O.L. Maynard of my staff.
Very truly yours, c
I Glenn L. Koester i Vice President - Nuclear GLK:see
- Enclosure l Attachments
- I-Safety Evaluation II-Type C LLRT Test Data i III-Significant Hazards Consideration IV-Proposed Technical Specification Changes cc: PO'Connor (2) w/a i JCannins w/a l
GAllen w/a i FJohnson w/a
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STATE OF KANSAS )
) ss CITY OF WICHITA )
Glenn L. Koester, of lawful age, being first duly sworn upon oath says that he is Vice President - Nuclear and an Officer of Kansas Gas and Electric Company; that he has read the foregoing document and knows the contel.t thereof; that he has executed the same for and on behalf of said Company with full Power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
By _ ~hj m f yh GlenEL'. Koester Vice President - Nuclear SUBSCRIBED and sworn to before me this 20th day of January , 1986 .
k4dC AnE Notary Public
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Mr. H. R. Denton Attachment I to KMLNRC 86-007 Pago 1 of 2 SAFETY EVALUATION This application requests a revision to the Wolf Creek Generating Station, Unit No. 1, 'Ibchnical Specifications Section 4.6.1.2.d to allow a one-time only defere nt of the specified 24 month maximum surveillance interval on the below listed containment penetrations for the first operating cycle only (Cycle 1); provided these surveillances are performed prior to startup following the first refueling outage (Refuel 1) . These surveillances cannot be safely performed at POWER OPERATION. The proposed changes to the Technical Specifications are provided as Attachment IV.
Refuel 1 is the next scheduled shutdown and is currently scheduled to begin in October, 1986. In the event of any unanticipated delays in the refueling schedule a unit shutdown to perform these surveillances will be initiated by October 31, 1986. Therefore this request entails an approximate seven month extension in the most limiting case. If an unscheduled shutdown of sufficient duration and resulting in appropriate plant conditions occurs prior to Refuel 1, Kansas Gas and Electric Company intends to perform any surveillances that can reasonably be completed. Since refueling outages normally occur approximately every twelve to eighteen months, extensions beyond the twenty four month maximum interval allowed by the Technical Specifications are not usually necessary. However, due to the extended plant startup program and plant performance during Cycle 1, Kansas Gas and Electric Ccrnpany is requesting the' aforementioned Technical Specification Revisions. Without the requested revisions Wolf Creek Generating Station would be forced into an outage for the purpose of surveillance testing.
Penetration Valve Surveillance Due Date 1
P-41 . . . . . . . . . . . . . BB-Vll8 .... 12/12/l6 BB-HV8351A .... 07/12/86 P-22 . . . . . . . . . . . . . BB-V148 .... 07/06/86 BB-ws351B .... 07/07/86
, P-39 . . . . . . . . . . . . . BB-V178 .... 07/09/86 BB-HV8351C .... 07/11/86 P-40 . . . . . . . . . . . . . BB-V208 .... 07/09/86 BB-HV8351D .... 07/09/86 P-24 . . . . . . . . . . . . . BG-V135 .... 03/01/86 BG- ws100 .... 03/01/86 BG-HV8112 .... 08/25/86 P-74 . . . . . . . . . . . . . EG-V204 .... 07/19/86 7 DG-HV58 .... 12/04/86 y EG-HV127 .... 12/04/86 P-8 0 . . . . . . . . . . . . . BG-8 381 .... 04/13/86 Bc- ws105 .... 07/18/86 i P-23 . . . . . . . . . . . . . BG-HV8160 .... 04/19/86 BG-w8152 .... 04/22/86 P-71 . . . . . . . . . . . . . EF-HV31 .... 07/12/86 P-73 . . . . . . . . . . . . . EF-HV45 .... 06/09/86 EF-W47 .... 06/09/86 P-28 . . . . . . . . . . . . . EF-W32 .... 09/25/86 P-29 . . . . . . . . . . . . . EF-HV46 .... 09/25/86 l
1* Note these valves included only for completeness
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Mr. H. R. Denton i Attachment I to KMLNRC 86-007 Page 2 of 2 For each of the penetrations considered in the request for extension of the surveillance interval, it can be shown that one or both of the following design parameters apply:
- 1. A closed ASME piping systen (inside and/or outside contairment) precludes the release of containment atmosphere to the environs.
- 2. Should an isolation valve leak slightly when closed, the pressurized fluid seal within the pipe precludes release of containnent atmosphere to the environs.
Specific details for each containnent penetration are included as Attachment II and III. In each case, possible degradation of a containment leakage barrier does not result in a release of radioactive materials from the containment atmosphere.
Therefore, the proposed amendment does not increase the probability or consequences of an accideat previously evaluated.
The proposed change does aat involve hardware changes nor a change in the manner in which the isolation syttems at each penetration functions.
Therefore, the possibility of creating a new or different kind of accident from any accident previously evaluated does not exist.
The proposed change does not alter the leakage rate limitations established in the Technical Specifications. 'Iherefore, the margin for possible deterioration of the containment integrity during the service intervals between tests has not been re'.md. Additionally, the integrity of the closed systems has been verified by the Containnent Integrated Lea *: Rate Test (ILRT). The inpact of extending the testing interval of penetrations bounded by these closed systens is reduced by satisfactory ILRT results.
The requested revisions would not have a significant inpact on safe operation of the plant when compared to the transient associated with a shutdown for the purpose of performing these surveillances. Based on the above and the considerations presented in Attachment III, the aforementioned proposed revisions to the Wolf Creek Generating Station, Unit No. 1, Technical Specifications do not adversely affect or endanger the health or safety of the general public or involve a significant safety hazard.
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Mr. H. R. Denton Attachment II to KMLNRC 86-007 Page 1 of 2 l l
TABLE 1A TYPE C LLRT TEST DATA (AIR TEST)
% of Current Measured Total Test Leak Rate Measured Valve Penetration Valve (1) Date scem(air) Leakage (2) Type (3)
P-41 BB-Vll8 12/12/84 8.59 00.0142 1 BB-ws351A 07/12/84 4.32 00.0071 3 P-22 BB-V148 07/06/84 211. 00.3485 1 BB-wa351B 07/07/84 10.37 00.0171 3 03/24/85 6.25 00.0103 3 P-39 BB-V178 07/09/84 0.30 00.0005 1 BB-HV8351C 07/11/84 4.40 00.0073 3 P-40 BB-V208 07/09/84 89.5 00.1478 1 BB-HV8351C 07/09/84 165.5 00.2734 3 P-24 BG-V135 02/29/84 14.43 00.0238 1 BG-HV8100 02/29/84 5.64 00.0093 2 BG-HV8112 08/25/84 2.93 00.0048 2 P-74 D V204 07/19/84 1383. 02.2843 1 EG-HV58 12/04/84 1276. 02.1075 2 EG-HV127 12/04/84 - -
2 P-80 BG-8381 04/13/84 2.35 00.0039 1 BG-HV8105 07/18/84 28.8 00.0476 2 P-23 BG-HV8160 04/19/84 30.7 00.0507 4 BG-HV8152 04/22/84 66.1 00.1092 4 Total 3310.18 05.4673 Notes: (1) Valves EG-HV58 and EG-HV127 tested simultaneously (2) % of current total measured leakage value = (current
- measured valve leak rate / current combined leakage rate (60,545 secm)) x 100 i (3) Valve Types
- 1. Check Valve t 2. Motor-Operated Gate Valve i 3. Motor-Operated Globe Valve
- 4. Air-Operated Globe Valve
Mr. H. R. Denton Attachment II to KMTERC 86-007 Page 2 of 2 TABLE IB TYPE C LLRF TESF DATA (MkTER TEST)
Measured Acceptance Test Leak Rate Criteria (2)
Penetration Valve (l) Date ml/hr ml/hr P-71 EF-W31 07/12/84 0 420 EF-W33 07/12/84 0 420 P-73 EF-W45 06/09/84 0 420 EF-W47 06/09/84 0 300 EF-W49 06/09/84 0 420 P-28 EF-W32 09/25/84 0 420 EF-W34 09/25/84 0 420 l p-29 EF-W46 09/25/84 0 420 EF-W48 09/25/84 0 300 EF-W50 09/25/84 0 420 Notes: (1) All valves are motor operated butterfly valves (2)The acceptance criteria is based on the criteria contained in ASME Section XI, Section IW-3426 which is 30(d) ml/hr where d= valve diameter I
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9 ATrg21MENF III 1
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Mr. H. R. Denton Attachment III to KMLNRC 86-007 Page 1 of 5 SIGNIFICANT HAZARDS CONSIDERATION This application requests a revision to the Wolf Creek Generating Station, Unit No. 1, Technical Specifications Section 4.6.1.2.d to allow a one-time only deferment of the specified 24 month maximum surveillance interval for the first operating cycle only (Cycle 1); provided these surveillances are performed prior to startup following the first refueling outage (Refuel 1) .
These surveillances cannot be safely performed at POWER OPERATION. The proposed changes to the Technical Specifications are provided as Attachment IV.
Refuel 1 is the next scheduled shutdown and is currently scheduled to begin in October, 1986. In the event of any unanticipated delays in the refueling schedule a unit shutdown to perform these surveillances will be initiated by October 31, 1986. Therefore this request entails an approxis te seven month extension in the most limiting case. If an unscheduled shutd~.;u of sufficient duration and resulting in appropriate plant conditions occurs prior to Refuel 1, Kansas Gas and Electric Cortpany intends to perform any surveillances that can reasonably be completed. Since refueling outages normally occur approximately every twelve to eighteen months, extensions beyond the twenty four month eximum interval allowed by the 'Ibchnical Specifications are not usually necessary. However, due to the extended plant startup program and plant performance during Cycle 1, Kansas Gas and Electric Coupany is requesting the aforementioned Technical Specification Revisions. Without the requested revisions, Wolf Creek Generating Station would be forced into an outage for the purpose of surveillance testing.
Penetration P-24 Reactor Coolant Pump Seal Water Return This penetration provides a return path for Reactor Coolant Puup No. 1 leak-offs and a flow path for excess letdown when excess letdown is directed to the suction side of the centrifugal charging pump. (See FSAR Figure 6.2.4.1, Page 20)
The piping from penetration P-24 is quality group B (ASME III, Class 2),
Seismic Catagory I and forms a closed loop outside contairrnent by way of the seal water heat exchanger to the suction side of the charging pumps in the energency core systen (ECCS) cooling recirculation loop. Isolation valves BG-HV-8100 and BG-HV-8112 close on receipt of a CIS-A signal.
Any back leakage from the contairrnent through the penetration would be returned to containnent by way of the energency core cooling systen recirculation loop.
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Mr. H. R. Denton Attachment III to KMLNRC 86-007 Page 2 of 5 Penetration P-80 Normal Charging This penetration provides the normal charging flow path to the reactor coolant system by way of the regenerative heat exchanger. (See FSAR Figure 6.2.4-1, Page 57)
The piping both inside and outside of containment is Quality Groups A and B (ASME III, Class 1 and 2), Seismic Category I.
The outside containment isolation valve BG-lW-8105 closes on recipt of a Safety Injection Signal (SIS) . In addition, valve BG-lW-8106 located outside containment closes on a SIS, providing additional isolation.
Fol' wing a postulated accident, the Auxiliary Building side of Penetration P-80 will be pressurized by the centrifugal charging pump while operating in the ECCS mode which precludes back-leakage from contairmnent through this penetration.
In addition, there are two (2) check valves inside containment located in the Quality Group A portion of the piping system, between the reactor coolant system (RCS) and the regenerative heat exchanger which provides an additional boundary against leakage from containment.
Penetration P-23 CVCS Normal Letdown This penetration provides a flow path for reactor coolant letdown from the RCS to the chemical and volume control systen (CVCS) outside contairunent.
(See FSAR Figu re 6.2.4-1, Page 19)
The piping inside the containment from the RCS is Quality Group A (ASME III, Class 1) , Seismic Category I, through two air-operated RCS isolation valves BG-LCV-459 and BG-ILV-460. These valves isolate on low pressurizer level and are designed to fail close. The piping between these isolation valves and the contairinent penetration is Quality Group B (ASME III, Class 2),
Seismic Category I. In addition, letdown orifice isolation valves BG-lW-8149 A, B, and C, located inside containment, isolate on low pressurizer level and are designed to fail close. The contairunent isolation valves BG-IW-8160 and BG-lW-8152 close on receipt of a CIS-A signal.
The automatic isolation valves BG-ILV-459, BG-IEV-460 and BG-lW-8149 A, B, C combined with the design of the piping system between these valves and penetration, provide an additional boundary against leakage from containment.
Mr. H. R. Denton Attachment III to KMLNRC 86-007 Page 3 of 5 Penetrations P-28, P-29, P-71, P-73 Essential Service Water (ESW) to and from contalment air coolers.
These penetrations provide an ESW flow path to and from the contaiment air coolers. The piping system inside containment is Quality Group C (ASME III, Class 3), Seismic Category I, and closed loop. (See FSAR Fig. 6.2.4-1 Pages 24, 25, 50 & 51)
The ESW piping and the containment air cooler inside contaiment serve an essential safety-related function. These penetrations are classified as essential penetrations with their associated isolation valves opening on receipt of a SIS.
With the ESW Systen operating under postulated accident conditions, the pressure at the penetrations will be greater than containment pressure.
Considering the conservative single failure assumption i.e. only one train in operation, no flow would exist to the two penetrations in the non-operating train.
The ESN piping inside contaiment, by the nature of it being a closed loop, provides an additional boundary to prevent contaiment teakage. In addition, outside contaiment, the water filled system provides a water seal to preclude leakage.
Penetrations P-22, P-39, P-40, P-41 RCP Seal Water Supply These penetrations provide a flow path from the CVCS charging pumps to the reactor coolant pump seals. (See FSAR Fig. 6.2.4-1 Pages 18, 29, 30 & 31) i These penetrations are classified as essential penetrations which are open i following a postulated accident. These penetrations direct water from the l centrifugal charging pumps to the Reactor Coolant Systen (RCS) .
l Outside contalment, the system is Quality Group B (ASME III, Class 2),
Seismic Category I. These penetrations are pressurized by a centrifugal charging pump. These penetrations perform an EOCS function and therefore do not require automatic isolation.
. In addition, there are two (2) check valves located inside containment I between the RCS &nd its associated penetration which provide an additional boundary against leakage from containment.
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Mr. H.:R. Denton
' Attachment III to KMLNRC 86-007 Page 4 of 5 3 i Penetration P.-74 Component Cooling Water to the Reactor Coolant Pimps
! 'Ihis penetration provides a flow path for wWnt cooling water to the reactor coolant pmp auxiliaries and other safety-related conponents. (See i FSAR Fig. 6.2.4-1 Page 52)
The Component Cooling Water System (CCWS) is Quality Group B and C (ASME III, Class 2 and 3), Seismic Category I, and a closed loop both inside and j outside cor..ainment.
Penetration P-74 is classified as an essential penetration with isolation not required until receipt of a CIS-B signal. Also the component cooling water surge tank, located in the Auxiliary Building, is provided with a radiation monitor located on the vent line which isolates the vent valve on i high radiation. Therefore, a closed system outside of containment is
- assured.
I j The CCNS piping inside contairmnent, by the nature of it being a closed loop, provides an additional boundary against leakage from containment. When the CCWS is in operation following a postulated accident, the pressure at the i i
penetration will be greater than containment pressure._ Considering the conservative single failure assmption, i.e. only one train in mperation and
- flow to the penetration isolated, leakage would be contained by the closed
! loop outside contairunent.
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a i Generically, each of the subject penetrations is from the Contairmnent into j the Auxiliary Building. Leakage through the recective isolation valves would be to a fluid filled system. Penetrations P-22, P-39,-P-40, P-41 and P-80 are at a pressure higher than the maximtun containment postulated peak accident pressure. Except for penetrations P-23 (normal letdown) and P-71,
! 73, 28, 29 (ESW supply and return), the systems associated with the
- 4 remaining penetrations are totally contained within the Auxiliary Building.
l Leakage from these systems into the Auxiliary Building would be filtered by ,
! the automatically actuated, safety-related Bnergency Exhaust Systen. It
- should be noted that the . dose calculations for accidents presented in l Chapter 15 of the FSAR are based on 100 percent unfiltered leakage from l contairunent,- i.e. no specific credit has been taken for filtered leakage j from contairmnent through the Auxiliary Building.
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Mr. H. R. Denton Attachment III to FMLNRC 86-007 Page 5 of 5 The contairunent leakage rate specified by 10 CFR Part 50, Appendix J and the WXIS Technical Specifications is a combined leakage rate of less than 0.60 La for all penetrations and valves subject to Type B and C tests, when pressurized to Pa. For WOGS 0.60 La is 252,028 scan at a Pa of 48 psig.
The present combined total measured leakage rate (Im) is 60,545 scem, which is less than 25 percent of the allowable. Attachment II provides a sununary of the LLRT Type C test Data for the subject penetrations. The total leakage rate associated with these penetrations is small when ecmpared to the total measured leakage rate Im for WCGS and therefore, if higher leak rates were to exist the effect would be insignificant.
The requested surveillance interval extension applies only to Type C LLRT requirements. All surveillances associated with denonstrating the operability of the containment isolation valves are current and will continue to be performed as required. Therefore, the above assessment combined with the assured operability of the valves demonstrates that the safety-related function of these penetrations is maintained.
Based on of the above discussions and those presented in Attachment I, it has been determined that the requested Technical Specification revision does 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 margin of safety. Therefore the requested license amendment does not present a significant hazard.
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. O ATTA09Efff IV
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