ML20195G021

Application for Amend to License NPF-42,revising TS 3.7.1.6 & Associated Bases.Proposed Change Would Modify LCO to Require Four ARVs to Be Operable;Eliminate Use of Required in Action Statements;Add New Action for Three or More ARVs

ML20195G021
Person / Time
Site:	Wolf Creek
Issue date:	06/11/1999
From:	Muench R WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20195G025	List: ML20195G021 ML20195G033
References
ET-99-0006, ET-99-6, NUDOCS 9906150193
Download: ML20195G021 (13)

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. . WapLF CREEK NUCLEAR OPERATING CORPORATION Richard A. Moench Vee President Engineering JWI111999 ET 99-0006 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station Pl-137 Washington, D. C. 20555

Subject:

Docket No. 50-482: Proposed Revision to Technical Specification 3.7.1.6, Steam . Generator Atmospheric Relief Valves Gentlemen:

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This letter transmits an application for amendment to Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). This request proposes to revise Technical Specification 3.7.1.6, Steam Generator i Atmospheric Relief Valves, and associated Bases. The proposed changes would:

modify the Limiting Condition for Operation (LCO) to require four ARVs to be OPERABLE; eliminate the use of " required" in the Action Statements; add a new ACTION for three or more ARVs inoperable; and limit the LCO 3.0.4 exception to one inoperable ARV.

A safety evaluation is provided in Attachment I. A No Significant Hazards Consideration Determination is provided in Attachment II. Attachment III is the related Environmental Impact Determination. Marked up pages are provided in Attachment IV (for . current Technical Specifications and Bases) and in Attachment V. (for improved Technical Specifications and Bases as approved by Amendment No. 123). Attachment VI provides a listing of commitments made in /

this submittal. The ACTIONS for multiple inoperable ARVs are revised i consistent with the improved Technical Specification (ITS) submittal of Diablo Canyon Power Plant and approved ITS for Comanche Peak Steam Electric Station.

Additionally, a similar change was recently approved for the Callaway Plant in Amendment No. 131.

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In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Kansas State Official. This proposed revision to the WCGS current Technical Specifications will be fully implemented within 60 days of formal NRC approval.

g 9906150193 990611 PDR ADOCK 05000482 p PDR .

P.O. Dox 411/ Burhngton, KS 66839 / Phone: (316) 364-8831 An Equal Opportunny Employer M FHCVET f

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f If you have any questions concerning this matter, please contact me at (316) 364-4,034, or Mr. Michael J. Angus, at (316) 364-4077.  ;

Very truly yours, I

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Richar A. Muench j RAM /rir Attachments: I - Safety Evaluation l II - No Significant Hazards Consideration Determination III - Environmental Impact Determination ,

IV - Proposed Current Technical Specification Changes i ll - Proposed Improved Technical Specification Changes VI - List of Commitments cc: V. L. Cooper (KDHE), w/a W. D. Johnson (NRC), w/a E. W. Merschoff (NRC), w/a K. M. Thomas (NRC), w/a Senior Resident Inspector (NRC), w/a l l

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i STATE OF KANSAS )

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COUNTY OF COFFEY ) l Richard A. Muench, of lawful age, being'first duly sworn upon oath says that ,

he is Vice President Engineering of Wolf Creek Nuclear Operating 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 l power and authority to do so; and that the facts therein stated are true and l correct toithe best of his knowledge, information and belief.

By Richard g. Muench j Vice Predident Engineering I 1

SUBSCRIBED and sworn to before me this /[ day of 3tOL , 1999, bDn A # CA$ OW Notary Public ' II NLN l Expiration Date -

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Attachment I to ET 99-0006 j Page 1 of 5 '

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SAFETY EVALUATION l

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, Attachment I to ET 99-0006 Page 2 of 5 Safety Evaluation Proposed Changes This license amendment request proposes to revise Wolf Creek Generating Station (WCGS) current Technical Specification 3. 7.1. 6, Steam Generator Atmospheric Relief Valves and associated Bases. The associated changes to the improved Technical Specifications, as approved by Amendment No. 123, are provided in Attachment V.

The amendment request proposes the following changes to Technical l Specification 3.7.1.6- 1 l

• Revise LCO 3.7.1.6 to require four ARVs to be OPERABLE; j e Eliminate the use of " required" in the Action Statements; l

• Adds a new ACTION for three or more ARVs inoperable; and ,

• Limit the LCO 3.0.4 exception to one ARV inoperable. l l

Background

An ARV is installed on the outlet piping from each steam generator (SG). The four valves are installed to provide for controlled removal of reactor decay heat during normal reactor cooldown when the main steam isolation valves are closed or the turbine bypass system (steam dump valves to the condenser) is not available. This is done in conjunction with the auxiliary feedwater system providing cooling water to the steam generators. The valves will pass sufficient flow at all pressures to achieve a 50'F per hour cooldown rate.

The total capacity of the four valves is 15 percent of rated main steam flow at steam generator no-load pressure. The ARVs also assure that subcooling can be achieved to facilitate equalizing pressure between the Reactor Coolant System (RCS) and the ruptured steam generator following a postulated steam generator tube rupture (SGTR) event, and that cooldown of the RCS to Residual Heat Removal (RHR) System entry conditions can be accomplished in a timely manner.

As noted in the Bases for Technical Specification 3. 7.1. 6, the SGTR event is the limiting analysis for defining ARV OPERABILITY requirements. The Updated Safety Analysis Report (USAR) discusses two explicit SGTR thermal-hydraulic analyses, SGTR with Failure of Ruptured SG Auxiliary Feedwater (AFW) Control Valve (Section 15.6.3.1), SGTR with Postulated Stuck-open ARV (Section 15.6.3.2) and a separate section on the SGTR Radiological Consequences (Section 15.6.3.3). The SGTR with failure of the Ruptured SG AFW Control Valve is referred to as the " margin-to-overfill" analysis, and assumes a single failure of the AFW control valve which maximizes the filling of the SG and thereby minimizes margin to SG overfill. The SGTR with Postulated Stuck-open ARV is typically referred to as the "offsite dose" analysis, and assumes a single failure of the ruptured SG ARV which maximizes calculation of offsite dose and thereby minimizes margin to offsite dose limits. In general, the

" margin-to-overfill" thermal-hydraulic analysis is referenced in the USAR to demonstrate that SG overfill will not occur, and the thermal-hydraulic results from the "offsite dose" analysis (i.e., mass and energy release history) are used in the offsite dose calculation.

However, at WCGS, although margin to SG overfill was predicted, it was decided to force SG overfill during the licensing process of the SGTR analysis in order to conservatively analyze the radiological consequences of a SGTR event.

Specifically, following SGTR overfill, the consequential failure of the

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. Attachment I to ET 99-0006 Paqe 3 of 5 ruptured SG safety valve to reseat due to water relief is assumed to result, thereby resulting in continued release of activity from the ruptured SG until the *RCS is cooled to RHR entry conditions to terminate the primary-to-secondary break flow. This continued release of activity from the ruptured SG l maximizes the radiological consequences of the SGTR event, and thus the I offsite dose following a SGTR event was calculated assuming the thermal-

! hydraulic results from the WCGS " margin-to-overfill" analysis. As noted in USAR Section 15.6.3.3.2, "The radiological consequences have been based on a worst case scenario, i.e., forced steam generator overfill with a stuck open safety valve."

Thus, the WCGS SGTR radiological consequences identified in USAR Section 15.6.3.3 are based on the thermal-hydraulic result of the SGTR " margin-to-overfill" analysis: SGTR with Failure of the Ruptured SG AFW Control Valve (USAR Section 15.6.3.1). As identified in USAR Section 15.6.3.3, the calculated radiological consequences are within the 10 CFR 100 acceptance limits. Since this forced SG overfill case is the basis for the WCGS SGTR offsite dose results, this is the specific SGTR analysis which will be referenced in the subsequent discussion.

The SGTR analysis assumed that two intact SG ARVs were available to perform RCS cooldown. For the forced SG overfill case, RCS cooldown is performed prior to SG overfill as a specific action to achieve RCS subcooling, a precursor to performing a RCS depressurization and subsequently terminating primary-to-secondary breakflow. RCS cooldown is also performed following SG overfill, cooling the primary system to RHR entry conditions to terminate the primary-to-secondary break flow. The single failure for this case is assumed to be the AFW control valve; with one ARV potentially out-of-service as allowed by current Technical Specification 3.7.1.6 and the ARV on the ruptured SG assumed to be unavailable for RCS cooldown, the ARVs on the remaining two intact SGs remained available to perform the RCS cooldown. However, the single failure of an intact SG ARV could be postulated for this case; with this assumption, although the failure of the AFW control valve need not be considered, there would be only one intact SG ARV available for RCS cooldown.

Consequently, WCNOC personnel performed a sensitivity analysis to determine the impact of the "new" single failure assumption of an intact SG ARV. As noted above, since current Tech Spec 3.7.1.6 requires three ARVs to be j OPERABLE, the single failure of an intact SG ARV coupled with one ARV out of I service and the ruptured SG unavailable results in only one intact SG ARV being available for RCS cooldown. A sensitivity analysis was performed to determine if SG overfill would be precluded following a SGTR " margin-to-overfill" case with only one intact SG ARV available for RCS cooldown. If the analysis showed margin-to-overfill, then it could be concluded that the current SGTR analysis was bounding, and that the plant could operate with only three OPERABLE ARVs. The results indicated that RCS cooldown with only one intact SG ARV could not be performed in time to preclude SG overfill. As j such, the results of the SGTR analysis presented in the USAR Sections 15.6.3.1 and 15.6.3.3 were not bounding, and additional action would be required. To ensure two intact SG ARVs are available for RCS cooldown, as currently modeled in the SGTR analyser, four OPERABLE ARVs are required to ensure that at least two intact SG ARVs are available to perform the necessary RCS cooldown. This amendment request revises current Technical Specification 3.7.1.6 to require four rather than three OPERABLE ARVs. Until this amendment request is l approved, administrative controls are in place to require four OPERABLE ARVs.

Evaluation l

The proposed change revises current Technical Specification LCO 3.7.1.6 to require at least four steam generator ARVs be OPERABLE, rather than three ARVs. The OPERABILITY of four ARVs ensures that reactor decay heat can be

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, Atiachment I to ET 99-0006 Page 4 of 5 dissipated.to the atmosphere in the event of a SGTR event coincident with a

. loss of offsite power, that subcooling can be achieved within assumed time constraint's and in accordance with single f ailure assumptions used in the SGTR analysis, : and that the RCS can be cooled down to the point of RHR System operation following SG overfill.

The proposed change revises the required allowed outage time for restoration when there are two inoperable ARVs due'to causes other than excessive seat leakage. With two ARVs inoperable, the allowed outage time for restoration of l- all but one ARV to OPERABLE status is changed from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The l existing specification allows one valve to be inoperable indefinitely and with I

one required ARV inoperable, the allowed outage time for restoration is seven days. By modifying the LCO to require four ARVs to be OPERABLE, an allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is more restrictive than the existing specification.

Therefore, revising the allowed outage time from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is l

acceptable based on a more restrictive allowed outage time from the existing specification and the low probability of an event requiring decay heat removal occurring during the restoration period that would require the ARVs. With respect to RCS cooldown for SGTR accident mitigation, the increase in time is acceptable based on the-low probability of a SGTR event occurring during the restoration period and the low probability of a SGTR event in conjunction with the failure of the turbine bypass system (i.e., loss of offsite power).

Additionally, this allowed outage time is consistent with the current Diablo Canyon Power Plant and Comanche Peak Steam Electric Station Technical Specifications. A similar change was recently approved for the Callaway Plant in Amendment No. 131.

The proposed change revises the ACTIONS to delete the use of " required." This change is consistent with NUMARC 93-03, " Writer's Guide for the Restructured Technical Specifications," Section 4.1.6, which states: " Occasionally an LCO requires OPERABILITY of only some of the components of a particular function which could be used to satisfy the requirement (i.e., two offsite circuits when three are installed and available). In this event, the Conditions, Required Actions and Surveillances which refer to the item (s) required by the LCO are preceded by " required." Since the LCO is revised to require four SG ARVs be OPERABLE and since the design of the system only provides four SG l

ARVs, the use of " required" is unnecessary.

The proposed change adds a new required ACTION for three or more ARVs inoperable and specifies an, allowed outage time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to return all but two ARV lines to an. OPERABLE status. This change is consistent with NUREG-1431, Rev.1, " Standard Technical Specifications, Westinghouse Plants," (as revised -by traveler TSTF-100) which indicates that with two or more atmospheric relief valve lines inoperable, the Required Action is to restore all but one to an OPERABLE status within a Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is based upon the low probability of an event occurring that would require the atmospheric relief valve lines to function.

The proposed change limits the exception to Specification 3.0.4 to new ACTIONS a and d.' This change is more restrictive than current Technical Specification requirements in that the exception would allow entry into an OPERATIONAL MODE or other specified condition with only one ARV inoperable. This change is consistent with NUREG-1431, Rev. 1.

Amendment No. 123, dated March 23, 1999, converts the current WCGS Technical l Specifications to the improved Technical Specifications. Attachment V to this letter provides proposed changes to the improved Technical Specifications l

(which are scheduled to be implemented by December 31, 1999). The proposed changes in Attachment V adopt the current Technical Specification ACTIONS based on excessive ARV seat leakage.

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Conclusion The propos'ed changes impose more restrictive requirements than currently exist in the WCGS Technical Specifications. These changes are consistent with the SGTR analysis assumptions, and ensure the radiological consequences of a SGTR event, as presented in the WCGS USAR, remain valid.

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Attachment II to ET 99-0006 Page 1 of 3 ATTACHMENT II NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION i

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, . Attachment II to ET 99-0006 Page 2 of 3 l No Significant Hazards Consideration Determination Proposed Changes -

This license - amendment . request proposes to revise Wolf Creek Generating

-Station (WCGS) current -Technical Specification 3. 7.1. 6, Steam Generator Atmospheric Relief Valves and associated Bases. The associated changes to the improved Technical Specifications, as approved ' by Amendment . No. 123, are provided in Attachment V.

The amendment request . proposes the following changes to Technical  ;

Specification 3.7.1.6:

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• Revise.LCO 3.7 1.6 to require four ARVs to be OPERABLE;

'o Eliminate the use of " required" in the Action Statements;.

• - Adds a new ACTION for three or more ARVs inoperable; and

• Limit the LCO 3.0.4 exception to one ARV inoperable.

Application of Standards The following Standards identified in 10 CFR 50.92 have been used to determine whether the proposed changes involve a Significant Hazards Consideration.

Each of the identified proposed changes is evaluated against the three Standards.

Standard I - Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated  ;

Revising the LCO to require four ARVs to be OPERABLE rather than three eliminating " required" from the Actions; adding a new ACTION for three or more ARVs inoperable; and limiting the LCO 3.0.4 exception to one ARV inoperable  ;

constitute more restrictive changes from the current Technical Specifications.

The- proposed changes do not affect initiating mechanisms or mitigation capabilities associated with SGTR events analyzed in Chapter-15 of the Updated Safety Analysis Report. The- proposed changes impose more stringent requirements to ensure that ARV OPERABILITY is maintained consistent with the safety analysis and licensing basis, and also to address all potential single  ;

failure scenarios. Therefore these changes do not involve a significant

. increase in the probability or consequences of an accident previously evaluated.

With two ARVs inoperable, .the allowed outage time for restoration of all but J i

one ARV to OPERABLE status is changed from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The existing

! specification allows one valve to be inoperable indefinitely and with one required ARV inoperable, the allowed outage time for restoration is seven ,

days. By. modifying the . LCO to require four ARVs to be OPERABLE, an allowed

. -outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is more restrictive than the existing specification.

! -Therefore, . revising the allowed outage time from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is

l. acceptable based on a more restrictive allowed outage time from the existing l -specification and the low probability of an event requiring decay heat removal occurring during the restoration period that would require the ARVs. With respect to Reactor Coolant-System cooldown for SGTR accident mitigation, the increase in time is acceptable based on the low probability of a SGTR event occurring during the restoration period and the low probability of a SGTR

. event in conjunction with the failure of the turbine bypass system (i.e., loss of offsite power). Therefore, this change in allowed outage time does not P

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' prevlously result .in analyzed a significant increase in the probabilit y or consequences of accidents.

Standard II - Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated There are no hardware changes nor are there any changes in the method by.which any safety related plant system performs its safety function. Revising the LCO . to require . four ARVs to be OPERABLE rather than three; eliminating i

" required" from the Actions; adding a new ACTION for . three or more ARVs inoperable; and limiting.the LCO 3.0.4' exception to one ARV inoperable will

-not impact the normal method of plant operation. The proposed changes ensure '

' operation of the plant. remains consistent with analysis assumptions. No new-accident scenarios, transient precursors, ' failure mechanisms, or limiting

.. single failures are introduced as'a result of the proposed changes. Based on the above discussion, the proposed ~ change does not create the possibility of a e new or different kind of accident from any previously evaluated.

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Standard III - Involve a Significant Reduction in the Margin of Safety The proposed changes do not affect the acceptance criteria for any analyzed

. event. There will be no effect on the manner in which safety limits or limiting safety system settings are-determined nor will there be any affect on those plant systems necessary to assure the accomplishment of protection functions. The proposed changes ensure operation of the plant consistent with the analysis assumptions. Therefore, there will be no impact on any margin of safety.

Conclusions Based 'on the above discussions, it has been determined that the requested f technical specification revisions do not involve a significant increase in the i probability of consequences of an accident or other adverse conditions 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 involve a significant hazards consideration. '

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ATTACHMENT III ENVIRONMENTAL IMPACT DETERMINATION l

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! Environmental Impact Determination )

This' license amendment request proposes to revise Wolf Creek Generating Station (WCGS) current Technical Specification 3.7.1.6, Steam Generator l' Atmospheric Relief Valves and associated Bases. The associated changes to the improved Technical Specifications, as approved by Amendment No. 123, are provided in Attachment V.

The amendment -request proposes the following changes to Technical )

l Specification 3.7.1.6:

• Revise ' LCO 3.7.1.6' to require four ARVs to be OPERABLE;

• Eliminate the use of " required" in the Action Statements;

• Adds a new ACTION for three or more ARVs inoperable; and

• Limit the LCO 3.0.4 exception to one ARV inoperable.

10 CFR 51.22(b) specifies the criteria for categorical exclusions from the requirement for a specific environmental assessment per 10 CFR 51.21. This amendment request meets the criteria specified in 10 CFR 51. 22 (c) ( 9) as specified below:

(1) the amendment involves no significant hazards consideration As demonstrated in Attachment II, the proposed changes do not involve any significant hazards consideration.

(ii) there ,is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite None of the proposed. changes involves a change to the facility or operating procedures that would cause an increase in the amounts of effluents or create new types of effluents. The proposed changes ensure that the plant is operated consistent with the analysis assumptions and that . the analysis results remain bounded.

(iii) there is no significant increase in individual or cumulative occupational radiation exposure i

The proposed changes will require'four ARVs to be OPERABLE rather than three; i eliminating " required" from the Actions; add a new ACTION for three or more ARVs inoperable; and limit the LCO 3.0.4 exception to one ARV inoperable.

These changes have no relation to occupational radiation exposure, either l individual or cumulative.

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.  !

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