ML20138B901
| ML20138B901 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 04/23/1997 |
| From: | Muench R WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20138B905 | List: |
| References | |
| ET-97-0028, ET-97-28, NUDOCS 9704290302 | |
| Download: ML20138B901 (11) | |
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WP,JLF CREEK NUCLEAR OPERATING CORPORATION Richard A. Muench Vice President Engineering April 23, 1997 ET 97-0028 U. S. Ncclear Regulatory Commission ATTN: Document Control Desk Mail Station F1-137 Washington, D. C. 20555
Subject:
Docket No. 50-482: Proposed Revision to Technical l Specification 3/4.9.4, Containment Building i Penetrations l
l Gentlemen:
l The following is an application for amendment to Facility Operating I.icense
[ No. NPF 42 for Wolf Creek Generating Station (WCGS). This request proposes to revise Technical Specification 3/4.9.4, Containment Building Penetrations, and its associated Bases section, to allow selected containment isolation valves to be opened under administrative controls during periods of core alterations or movement of irradiated fuel inside containment.
A Safety Evaluation is provided in Attachment I, and Attachment II provides a No Significant Hazards Consideration Determination. Attachment III is the Environmental Impact Determination, and the marked-up technical specification pages for this request are provided in Attachment IV.
In accordance with 10 CFR 50.91, a copy of this revision to our original application, with attachments, is being provided to the designated Kansas State Official. This proposed revision to the WCGS Technical Specifications will affect outage-related operations. Therefore, to allow this amendment to be implemented prior to the start of the WCGS ninth r6 fueling outage, WCNOC requests this amendment be approved by August 15, 1997 08)b i
9704290302 970423
P.O Box 411/ Burhngton, KS 66839 i Phone. (316) 364-8831 :
j An Equal Opgx>rtunity E mployer M F HC VET i
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l ET 97-0028 !
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l If you have any questions concerning this matter, please contact me at (316) 364-8831, extension 4034, or Mr. Richard D. Flannigan, at extension 4500.
Very truly yours, l
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l Richar A. Muench i
RAM /jad Attachments I - Safety Evaluation II - No Significant Hazards Consideration Determination l III - Environmental Impact Determinati(.,n IV - Proposed Technical Specification Char.ge l-cc: V. L. Cooper (KDHE), w/a ,
E. W. Merschoff(NRC), w/a I l W. D. Johnson (NRC), w/a J. F. Ringwald (NRC), w/a ,
J. C. Stone (NRC), w/a i
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STATE OF KANSAS )
) SS COUNTY OF COFFEY ) .
Richard A. Muench, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Oper at ing Corporation; that he has read the foregoing document and knows the content thereof; 'that i he has executed that same for and'on behalf of said Corporation with full '
power and authority to do so; and that the facts therein stated are true and ;
correct to the best of his knowledge, inforration and belief.
y ll$ N Richard . Muench Vice Pre.sident Engineering SUBSCRIBED and sworn to before me this O day of Apr[/ , 1997.
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JUUE A. DALE Notary PubEc. Stape of K,anses Notarguntic A Appt. Expires / /U JD /'] E / /
Expiration Date /O/20/ N ,
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Attachment I to ET 97-0028 Page 1 of 4 4
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. ATTACHMENT I SAFETY EVALUATION l 1 e
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l l Attachment I to ET 97-0028 Page 2 of 4 Safety Evaluation Proposed Change This license amendment request proposes to revise the Wolf Creek Generating Station (WCGS) Technical Specification 3/4.9.4. This revision adds a new item d to the limiting condition of cperation to state that Penetration P-63 (Service Air valves KA V-039 and KA V-118) and Penetration P-98 (Breathing Air valves KB V-001 and KB V-002) may be opened under administrative controls.
The new item d would allow these penetrations to be unisolated during CORE ALTERATIONS and movement of irradiated fuel assemblies within containment providea that specified administrative controls are employed. The associated Bases section is also revised to reflect this change and describ^ the administrative controls to be used.
Evaluation l The requirements for containment penetration closure ensure that a release of l fission products within containment will be restricted from escaping to the
( environment. Allowing penetration flow paths that provide direct access from l the containment atmosphere to the outside atmosphere to be opened under
- administrative control raises the concern that radioactivity could be released l through the unisolated flow paths and vented to the outside environment should i
accidents that involve radioactivity release occur during core alterations or movement of irradiated fuel. During plant outages service air and breathing air must be provided to the containment building to support various outage activities. Normally, Penetration P-63 (Service Air valves KA V-039 and KA V-118) and Penetration P-98 (Breathing Air valves KB V-001 and KB V-002) would I not provide direct access from containment atmosphere to the outside atmosphere, due to the pressurization of the service air and breathing air lines from their respective air compressors when the manual isolation valves are open. However, the possibility exists that direct access could be established through Penetrations P-63 and P-98 if an air compressor was turned off or failed with its respective isolation valves open. Thus, administrative l controls must be established during periods of core alterations and movement j of irradiated fuel inside containment to ensure that the manual Service Air l valves KA V-039 and KA V-ll8 and manual Breathing Air valves KB V-001 and KB l V-002 are closed when their respective air compressors are not running.
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t Accidents that could result in a release of radioactive material through the l
potential leak paths include a fuel handling accident that results in breaching of the fuel rod cladding, and a loss of residual heat removal (RHR) cooling event that leads to core boiling and uncovery. The impact of the alternate leak path on the radiological consequences of accidents that involve j
a release of fission product radioactivity within containment is discussed below.
Fuel Handling Accident:
During movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a fuel handling accident. The l fuel handling accident is a postulated event that involves damage to l irradiated fuel. Fuel handling accidents include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies, In Operational Mode 6, Refueling, it is highly unlikely that a fuel handling accident could result in pressurization of the containment building.
Therefore, the majority of the radioactive material released from a fuel handling accident during refueling would be held up inside containment, since 4
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Attachment I to ET 97-0028 j ~ Page 3 of 4 4 l
. without containment pressurization there would be no motive force to drive the ~
! radioactive material through the unisolated flow paths. However, even if a j release were to take place through these alternate leak paths, the dose 2
consequences of this release would be less than that which would occur through
- the open containment personnel airlock doors. These doors are allowed to be
- open under administrative controls during core alterations and movement of
- irradiated fuel (per WCGS Technical Specification 3.9.4, " Containment Building
!_ Penetrations," Amendment No. 95). The potential release through the {
4 unisolated containment penetrations is bounded by the fuel handling accident '
l analysis performed to provide justification for containment personnel airlock i j doors revision (WCGS Technical Specification Amendment 95). That analysis 4 assumes that all of the fuel rod gap activity is released from the damaged j rods and all the gaseous effluent escaping from the refueling pool is released directly to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> through the open personnel airlock i doors. In addition, no credit is taken for mixing of the gaseous effluents j with the surrounding building atmosphere, nor the removal of any iodine by the j atmosphere filtration system filters. The potential dose consequences from a j simultaneous release of the gaseous effluents through the unisolated i penetration flow paths and the open personnel airlock doors will not be different from that analyzed for the open containment personnel airlock doors.
i That is because that analysis assumes all radioactive material from the fuel I handling accident is released to the environment through the open personnel
} airlock doors. Therefore, allowing penetration flow paths to be unisolated J during core alterations or movement of irradiated fuel would not invalidate j the conclusion, presented in USAR Section 15.7.4, that the potential dose i
consequences from a fuel handling accident will be well within 10 CFR 100 limits.
Loss of RHR Cooling:
Release of radioactive materials would be insignificant as a result of core boil-off.due to a loss of RHR cooling, provided the event does not continue for an extended period of time, which would then result in core uncovery and subsequent core damage. Core boil-off could cause pressurization of containment which would establish a driving force for the containment atmosphere to be released, via the unisolated penetration flow paths, to the outside atmosphere. If core boil-off continues past the point where the core is uncovered, subsequent core damage would release radioactive material to containment atmosphere, which could then be released to the outside atmosphere through the open penetration (s). However, the consequences of this release of radioactivity due to core boil-off with no consideration for core uncovery and core damage is expected to be significantly lower than those from a fuel handling, accident. This is because the total coolant activities (corresponding to 1% fuel defect) during a loss of RHR cooling event occurring 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> following reactor shutdown (the earliest time fuel offloading may commence) are less than the total gap activities released from the damaged fuel rods from a fuel handling accident.
A review of a typical calculation performed for outage risk assessment reveals that the time to core boil would be greater than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> should a loss of RHR cooling event occur at the beginning of fuel of floading, based on the normal water level maintained in the refueling pool. Technical Specification 3/4.9.8 requires that corrective actions be taken immediately to restore RHR cooling (by having one RHR loop operable and in operation) as soon as possible if RHR loop requirements are not met. It also requires that all containment penetrations that are open to the outside atmosphere be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
If RHR cooling, or an alternative method of core cooling, is restored within the 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, the scenario involving core boiling and subsequent containment pressurization would not occur and thus radioactivity release from the reactor core through the unisolated penetration flow paths would not be a concern.
. 1 Attachment I to ET 97-0028 1
Page 4 of 4 i Adding the proposed item d to the Technical Specification 3/4.9.4 LCO is I acceptable based on administrative controls that would be used in conjunction l
, with the note. These controls include written procedures that require j
! designated personnel to be informed of the open . status of the valves in 1
, question and specified persons to be designated and readily available to
{ isolate the open penetration in the event of a fuel handling accident. These l administrative controls provide protection equivalent to that provided by the j j administrative controls used to establish containment closure for a l containment personnel airlock. The NRC staff recently approved (WCGS 2 l
{' Technical Specification Amendment 95) changes 'o the requirements for the containment personnel airlock that allow both deors of the airlock to be open l during CORE ALTERATIONS and during movement of irradiated fuel inside j containment, provided that administrative controls are in place to quickly
! close one door and establish containment closure.
l Based on the above discussions and the no significant hazards consideration i determination presented in Attachment II, the proposed change does not j increase the probability of occurrence or the consequences of an accident or i malfunction of equipment important to safety previously evaluated in the i safety analysis report; or create a possibility for an accident or malfunction
! of a different type than any previously evaluated in the safety analysis l report; or reduce the margin of safety as defined in the basis for any i technical specification. Therefore, the proposed change does not adversely 4
affect or endanger the health or safety of the general public or involve a ,
significant safety hazard.
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l l Attachment II to ET 97-0028 i Page 1 of 3 l ,
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l ATTACHMENT II NO SIGNIFICANT RAZARDS CONSIDERATION DETERMINATION l
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Attachment II to ET 97-0028 Page 2 of 3 No Significant Hazards Consideration Determination This license amendment request proposes to revise the Folf Creek Generating Station (WCGS) Technical Specification 3/4.9.4. This revision adds a new item d to the limiting condition of operation to state that Penetration P-63 (
(Service Air valves KA V-039 and KA V-118) and Penetration P-98 (Breathing Air l valves KB V-001 and KB V-002) may be opened under administrative controls.
The new item d would allow these penetrations to be unisolated during CORE 1 ALTERATIONS and movement of irradiated fuel assemblies within containment provided that specified administrative controls are employed. The associated Bases section is also revised to reflect this change and describe the administrative controls to be used.
l Standard I - Involve a Significant Increase in the Probability or I Consequences of an Accident Previously Evaluated The proposed change involves changes to the Technical Specification i requirements for containment closure which is an accident mitigating feature. )
The changes would not affect the likelihood of occurrence of any accidents The proposed change t(oes not involve any hardware or previously evaluated. l plant design changes. The containment leakage value is not assumed to be an initiator of any analyzed event. Containment isolation valves and temporary closure devices serve to limit the radiological consequences of accidents.
The proposed change would ensure the service air and breathing air manual l isolation valves will perform their required containment closure function and l will serve to limit the consequences of a fuel handling accident as described in the USAR, such that the results of the analyses in the USAR remain bounding. In considering the consequences of a design basis fuel handling accident inside containment, the assumptions in the analysis take no credit i for the containment as a barrier to prevent the postulated release of radioactivity. For events that could occur during CORE ALTERATIONS or movement of irradiated fuel assemblies, containment closure is considered a defense-in-depth boundary to prevent uncontrolled release of radioactivity. i Additionally, the proposed change does not impose any new safety analyses ;
limits or alter the plant's ability to detect and mitigate events. ,
Therefore, this change does not involve a significant increase in the I probability or consequences of an accident previously evaluated.
Standard II - Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated The proposed change involves reliance on manual actuation of containment penetration valves (Service Air valves KA V-039 and KA V-ll8 and Breathing Air valves KB V-001 and KB V-002 are manual valves) to block the unimpeded flow of I the containment atmosphere to the environment under certain conditions. The proposed change would not necessitate a physical alteration of the plant l features that provide core cooling or subcriticality (no new or different type of equipment will be installed) or changes in parameters governing plant )
operation during CORE ALTERATIONS or movement of irradiated fuel in containment. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. ;
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Standard III - Involve a Significant Reduction in the Margin of Safety i
The proposed change is similar to the use of administrative controls to isolate an open containment airlock door. The use of administrative controls ;
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Attachment II to ET 97-0028 Page 3 of 3 in this manner has been approved by the NRC (WCGS Technical Specification Amendment 95) for plant operations that would not require the containment to maintain a pressure boundary. This scenario is applicable during plant shutdown for refueling when CORE ALTERATIONS and movement of irradiated fuel assemblies in the containment occur. Accidental damage to spent fuel during these operations is classified as a fuel handling accident. The proposed change has been developed considering the importance of the containment boundary in limiting the consequences of a design basis fuel handling accident. The proposed change allows for protection equivalent to that provided by previously approved methods of containment closure. Considering the probability of an event that would challenge the containment boundary, the alternative protection provided by this change, and the . operational requirements to occasionally open these penetrations, the proposed change is acceptable and any reduction in the margin of safety is insignificant.
Based on the above discussions, 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 i previous evaluations; or create the possibility of a new or diffarent kind of accident or condition over previous evaluations; or involve a significant reduction in a margin of safety. The requested license amendment does not involve a significant hazards consideration.
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1 Attachment III to ET 97-0028 Page 1 of 2 l
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ATTACHMENT III ENVIRONMENTAL IMPACT DETERMINATION l
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4 l Environmental Impact Determination i
i 10 CFR 51.22(b) specifies the criteria for categorical exclusions from the i
requirement for a specific environmental assessment per 10 CFR 51.21. This j amendment request meets the criteria specified in 10 CFR 51. 22 (c) ( 9) as j specified below:
l (i) the amendment involves no significant hazards consideration i
- As demonstrated in Attachment II, the proposed change does not involve any j
! significant hazards consideration. !
f (ii) there is no significant change in the types or significant increase in i
the amounts of any effluents that may be released offsite j The proposed change does not involve a change to the facility or operating
- procedures that would cause an increase in the amounts of effluents or create j new types of effluents. l (iii) there is no significant increase in individual or cumulative occupational radiation exposure The proposed change would require individuals to be designated and readily available to close the service air and breathing air valves f*11owing an evacuation that would occur in the event of a fuel handling aculdent. One method of accomplishing this would be to post individuals at or near the location of the outboard isolation valves for the two penetrations. The two locations in question are in low radiation areas. Therefore, 'the two individuals that would be assigned to these locations would receive slight radiation exposure. However, this exposure would be comparable to, if not less than, the exposure these individuals would receive in most work areas outside the reactor bioshield area. Thus, the proposed change will not result in a significant increase in individual or cumulative occupational radiation exposure.
Based on the above it is concluded that there will be no impact on the environment resulting from this change and the change meets the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of '
10 CFR 51.21 relative to requiring a specific environmental assessment by the Commission. j l
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