ML20247F845
| ML20247F845 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/30/1989 |
| From: | Withers B WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20247F847 | List: |
| References | |
| WM-89-0096, WM-89-96, NUDOCS 8904040058 | |
| Download: ML20247F845 (9) | |
Text
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LF CREEK W8) NUCLEAR OPERATING 8",$M
March 30,1989 Chief Executive Offcer WM 89-0096 U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk Mail Station F1-137 Washington, D. C. 20555
References:
- 1) KMLNRC 86-210, dated November 7, 1986, from l
G. L. Koester, KG&E, to H. R. Denton, NRC
- 2) Letter dated March 16, 1989, from J. A. Calvo, NRC, to B. D. Witeers, WCNOC Subj ect :
Docket No. 50-48?.: Proposed Technical Specification Supporting Steam Generator Tube Rupture Analysis Gentlemen:
The purpose of this letter is to modify the proposed technical specification supporting the steam generator tube rupture analysis for Wolf Creek Generating Station (WCGS).
The original proposed technical specification was submitted in Reference 1.
In Reference 2, the NRC requested that changes be made to the technical specifications es proposed, by superseding the original proposed change.
The
Attachment:
to this letter provide a revised Safety Evaluation, 91gnificant Hazards Consideration, and proposed technical specification change.
If you have any questions concerning this matter, please contact me or Mr. O. L. Maynard of my staff.
t Very truly yours, Bart D. Withers President and Chief Executive Officer BDW/j ad Attachments:
I - Safety Evaluation II - Addressing the Standards in 10 CFR 50.92 III - Proposed Technical Specification Changes cc:
G. W. Allen (KDHE), w/a 1
B. L. Bartlett (NRC), w/a
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E. J. Holler (NRC), w/a R. D. Martin (NRC), w/a D. V. Pickett (NRC), w/a k
P.O. Box 411/ Burlington, KS 66839 / Phone: (316) 364-8831 uahtumy Ern@yer M FtCVET ADL k d" S2 P
PDC l
STATE OP KANSAS
)
) SS C0ffMTY OP COPPEY
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Bart D. Withers, of lawful age, being first duly sworn upon oath says that he is President and Chief Executive Officer of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the content thereof; that 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 ;nowledge,-
information and belief.
By
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J Bart D. Withers President and Chief Executive Officer SUBSCRIBED and sworn to before me this 0 0 d.2y of N S4cdf 1989.
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Notary Public
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4 ATTACHMENT I SAFETY EVAT.fitTLON 1
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' Attachment I to WM 89-0096 Page 1 of 3 SAFETY EVALUATION Introduction Wolf Creek License Conditicn 2.c(11) states:
" Prior to restart following the first refueling outage, KG&E shall submit for NRC review and approval an analysis which demonstrates that the steam generator tube rupture (SGTR) analysis presented in the FSAR is the most severe case with respect to the release of fission products and calculated doses.
Consistent with the analytical assumptions, the licensee shall propose all necessary changes to Appendix A to this license".
The SGTR analysis has been completed and submitted to the NRC for review (References 1, 2, and 3).
This analysis takes credit for Atmospheric Relief Valve (ARV) operation for mitigating the consequences of a SGTR and states that a Technical Specification amendment request will be submitted concerning the operability of the ARV's in a SGTR event.
Proposed Technical Specification requirements were developed in order to support the SGTR analysis and su*. ltted to the NRC in Reference 4.
After
. review of the proposed Technical Specification, the NRC identified two areas which needed resolution in Reference 5.
These areas are:
(1) the time inte rval allowed for inoperability of one of the required three ARV's, and (2) operability with an ARV which has been isolated due to excessive seat leakage.
In order to resolve these issues, the proposed technical specification changes have been modified to reflect what the NRC has indicated would be acceptable.
The new proposed requirements state that three of the four steam generator atmospheric relief valves shall be operable in Modes 1,
2, and 3.
If only two ARV's are operable, this specification allows 7 days to return one of the two inoperable ARV's to operable status.
If only one or no ARV is operable, this specification allows only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to return to at least two operable ARV's.
If one or more of the required ARV's are inoperable due to excessive seat leakage with the associated block valve closed, the specification allows 30 days to restore the required number of ARV's to operable status.
An action statement is also provided which allows the plant to start up if the Limiting Condition of Operation (LCO) is not met.
- Finally, the proposed Technical Specification would not increase the surveillance requirements for the ARV's beyond those addressed in existing Specification 4.0.5.
Evaluation The purpose of the license amendment request is to incorporate technical specification LCO and surveillance requirements for the steam generator ARV's into the Wolf Creek Ooerating License to assure the availability of mitigating equipment assumed in the SGTR analysis.
The technical specification requirements constitute additional limitations on facility operations.
No requirements on ARV operability have been included in the existing Wolf Creek Technical Specifications because the ARV's have not been required in the mitigation of postulated accidents and transients.
l
httachment 1 to 1M 89-0096 Pa$c 2 of 3 The Wolf Creek Generating Station design includes four ARV's, one ARV for each steam genertor.
The ARV's are safety grade and have Class lE electrical control circuits (Reference 6).
The operability of at least three of the four ARV's ensures that reactor decay heat can be dissipated to the atmosphere in the event of a SGTR coincident with a loss of offsite power and that the Reactor Coolant System (RCS), can be cooled down to the point of Residual Heat Removal System (RHR) operation. In the initial phases of a SGTR event, the ARV's are relied upon for achieving proper RCS subcooling prior to equalizing pressures between the RCS and the faulted steam generator.
The proposed number of operable ARV's assures that subcooling can be. achieved within assumed time constraints and in accordance with single failure assumptions used in the SGTR analysis.
For the RCS cooldown to RHR initiation, only one ARV is needed for the heat removal required.
Three operable ARV's are adequate, assuming that one of the operable valves is on the faulted steam generator, and thus is not available for heat removal, and that one ARV fails to function in accordance with accident assumptions.
The proposed Technical Specifications are applicable in plant operational Modes 1, 2, and 3.
Since the purpose of the ARV's is to provide for removal of reactor decay heat during the initial phases of the SGTR event and during cooldown to KHR initiation, and since the RHR system is available in Modes 4, 5,
and 6,
this Technical Specification is only applicable in Modes 1, 2, and 3.
Each ARV is equipped with a manual block valve, located near the ARV in the auxiliary building, to provide a positive shutoff capability should an ARV develop leakage.
Closure of the block valve; of all ARV's because of excessive seat leakage does not endanger Cne reactor core because decay heat can be dissipated with the steam line safety valves. Also, consistent with SGTR analysis assumptions, a block valve can be used to control release of steam to the atmosphere.
The restoration time periods provided in the proposed LCO Action statement are based on the low likelihood of having a SGTR event coincident with a loss of offsite power during the time period that one or more of the required ARV's is out of service.
The allowed time period (7 days) for the case of one inoperable ARV is longer than the time period (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) allowed when more than one ARV is inoperable.
The surveillance requirements of the Wolf Creek inservice valve testing program for ASME Code valves are considered appropriate and sufficient for the ARV's.
Therefore, the surveillance requirements in the proposed Technical Specification refer to the provisions of existing Specification 4.0.5.
Testing of the ARV's at power which involves stroking the valves is i
performed with the block valves in the closed position.
This precludes a plant transient associated with inadvertent failure of an ARV in the open i
position.
I
Attachment I to WM 89-0096 Page 3 of 3 j
Conclusions The proposed ARV Technical Specification constitutes new LCO and surveillance requirements which govern operability of existing equipment in the Wolf Creek design.
It was concluded that the proposed Technical Specifications would not adversely affect the conclusions in existing plant transient and accident analyses and that the propased requirements provide acceptable operability limitations for ARV's consistent with the assumptions used in the SGTR analysis.
Therefore, the level of plant safety described in the Wolf Creek FSAR is not adversely affected by the proposed amendment.
References 1.
SLNRC 86-01, dated 01/08/86, Steam Generator Tube Rupture Analysis - SNUPPS 2.
SLNRC 86-03, dated 02/11/86, Steam Generator Tube Rupture Analysis - SNUPPS 3.
SLNRC 86-08, dated 04/01/86, Steam Generator Tube Rupture Analysis - SNUPPS 4.
KMLNRC 86-210, dated 11/07/86, Proposed Technical Specification Supporting Steam Generator Tube Rupture Analysis 5.
Letter, dated 03/16/89, from J. A. Calvo, NRC, to B. D. Withers, WCNOC 6.
Wolf Creek Generating Station Final Safety Analysis Report, Section 10.3
l I
ATTACHMENT II ADDRESSING THE STANDARDS IN 10 CFR 50.92 w___-____---_____________
. 1 to WM 89-0096 Page 1 of 2 ADDRESSING THE STANDARDS IN 10 CFR 50.92 t
The proposed Technical Specification for steam generator Atmospheric Relief l
Valves (ARV's) incorporates new requirements into the Wolf Creek Operating License.
These requirements constitute additional limitations on facility operations to aneure operability of the ARV's and to impose operating restrictions if less than the required number of ARV's is available.
The proposed specification does not involve a significant hazards consideration because operation of Wolf Creek Generating Station (WCGS) in accordance with the proposed requirements would not:
1.
involve a significant increase in the probability or consequences of an accident previously evaluated. The ARV's are not relied upon for mitigation of accidents or transients, other than the Steam Generator Tube Rupture (SGTR) event, previously evaluated in the WCCS Final Safety Analysis Report (FSAR).
The Limiting Conditions for Operation (LCO) do not alter the manner in which ARV operation was considered in previous accident and transient analysis.
Surveillance testing of ARV's, which involves stroke cesting, is performed (in accordance with existing surveillance test program procedures) with the block valves closed. Thus, a plant transient is precluded.
To assure that the ARV's are available for mitigation of a postulated SGTR event, the proposed specification establishes surveillance requirements and restrictions on ARV operability consistent with the assumptions used in the SGTR event analysis submitted to the NRC in accordance with WCGS License Condition 2.c(ll). Therefore, the proposed Technical Specification requirements would neither increase the probability of an accident nor increase the consequences of accidents previously evaluated.
2.
create the possibility of a new or dif ferent kind of accident from any previously evaluated.
Establishing new LCO and surveillance requirements for the existing WCGS ARV's does not result in the possibility of new or different types of accidents.
3.
involve a significant reduction in a margin of safety.
The proposed Technical Specification assures that the margin of safety established in the current SGTR analysis is maintained.
As discussed in items 1 and 2 above, the proposed LCO and surveillance requirements do not alter the margins to safety established in previous accident and transient analysis or in established surveillance test p rog rams.
. 1 to WM 89-0096 Page 2 of 2 The.NRC has established guidance concerning the determination of whether a significant hazards consideration exists by providing certain examples (51 FR 7751) of amendments not likely to involve a signficant hazards consideration.
The proposed ARV Technical Specification conforms to NRC example (ii)
"A change that constitutes an additional limitation, restriction or control not presently included in the technical specifications, e.g. a more stringent surveillance requirement."
Based on the above discussion, supplemented by the existing similarity of the proposed Specification requirements to an NRC example of a change not likely to involve a significant hazards consideration, it has been concluded that the proposed ARV Technical Specification does not involve c. significant hazards consideration.
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