ML20086R149
| ML20086R149 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/25/1995 |
| From: | Hagan R WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20086R152 | List: |
| References | |
| ET-95-0068, ET-95-68, NUDOCS 9507310036 | |
| Download: ML20086R149 (12) | |
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W$LF CREEK NUCLEAR OPERATING CORPORATION
- Robert C. Hagan Vice President Engineenng July 25, 1995 ET 95-0068 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, D.
C.
20555
Subject:
Docket No. 50-482:
Revision to Technical Specification 4.0.5,
" Surveillance Requirements for Inservice Inspection and Testing Program" Gentlemen:
This letter transmits an application for amendment to Facility Operating License No. NPF-42 for Wolf Creek Generating Station (WCGS).
This license amendment request proposes revising Technical Specification 4.0.Sa and Bases Section 3/4.4.10 to delete the phrase,
- (g),
except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section
- 50. 55a (g) (6) (i). "
This proposed change is consistent with NUREG-1431, " Standard Technical Specifications - Westinghouse Plants,' and NUREG-1482, " Guidelines for Inservice Testing and Nuclear Power Plants."
This proposed change would allow the implementation of a relief request without prior NRC approval, upon finding an ASME Code requiret.nt impractical because of prohibitive dose rates or limitations in the design, constructio% or system configuration.
This implementation could occur provided the relief request has been (1) acceptably reviewed pursuant to 10 CFR 50.59; and (2) approved by the plant staff in accordance with the administrative process described in the inservice inspection and testing programs administrative procedures; and (3) reviewed and approved by the Plant Safety Review Committee. This proposed change would also alleviate situations where compliance with the technical specifications cannot be achieved for the period between the time of preparation and submittal of a relief request, until the NRC has issued a safety evaluation and granted the relief.
Attachment I provides a description of the proposed change along with a Safety Evaluation.
Attachment II provides a No Significant Hazards Consideration Determination.
Attachment III provides the Environmental Impact Determination.
The specific change to the technical specifications proposed by this request is provided as Attachment IV.
In accordance with 10 CFR 50.91, a copy of this appl'. cation, with attachments, is being provided to the designated Kansas State official.
This proposed revision I
to the WCGS Technical Specifications will be fully implemented within 30 days of r
formal Nuclear Regulatory Commission approval.
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3IvvU' O fi n 9507310036 950725 PDR ADOCK 05000482 ru. tsox 411 i uuningtonN66839 / Phone: (316) 364-8831
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An Equal Oppo6 unity Employer WFMCNET 1
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ET 95-0068 Page'2 of 2 If~ you have any questions concerning this matter, please contact me at (316) 364-8831, extension 4553, or Mr. Richard D. Flannigan, at extension 4500.
Very truly yours, hhl i
Rdbert C. Hagan
-RCH/jra Attachments:
I - Safety Evaluation II - No Significant Hazards Consideration Determination III - Environmental Impact Determination IV - Proposed Technical Specification Change cca G.
W.
Allen (KDHE), w/a L. J.
Callan (NRC), w/a D.
F.
Kirsch (NRC), w/a J.
F. Ringwald (IEC), w/a J. C.
Stone (NRC), w/a i
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1 STATE OF KANSAS
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COUNTY OF COFFEY
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~ Robert C. Hagan, 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 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, informatior sad belief.
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e.$t e of Kansas By I'
Ah-f Wy Appt. Expires f wdy J, /Ff F ;
Robe C. Hagan
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Eng neering SUBSCRIBED and sworn to before me this 28. day of '/leel->
, 1995.
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Notary Olblic
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' Attachment I to ET 95-0068 Page 1 of 3 Q
ATTACHMENT I SAFETY EVALUATION l
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Attachment I to ET 95-0068 Page 2 of 3 Safety Evaluation PIspnand Changs This license amendment request proposes revising Technical Specification 4.0.5a and Bases Section 3/4.4.10 to delete the phrase, " (g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50. 55a (g) (6) (i),"
This proposed change is consistent with NUREG-14 31,
" Standard Technical Specifications - Westinghouse Plants,"
and NUREG-1482,
" Guidelines for Inservice Testing at Nuclear Power Plants."
This proposed change would allow the implementation of a relief request without prior NRC approval, upon finding an ASME Code requirement impractical because of prohibitive dose rates or limitations in the design, construction, or system configuration.
The implementation could occur provided the relief request has been (1) acceptably reviewed pursuant to 10 CFR 50.59; and (2) approved by the plant staff in accordance with the administrative process described in the inservice inspection and testing programs administrative procedures; and (3) reviewed and approved by the Plant Safety Review Committee.
Eyaluation Currently, Technical Specification 4.0.5 specifies, in part, the following requirements:
" Surveillance Requirements for inservice inspection and testing of ASME Code Class 1,
2, and 3 components shall be applicable as follows:
a.
Inservice inspection of ASME Code Class 1,
2, and 3 components and inservice testing of ASME Code Class 1,
2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 5 0. 5 5a (g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50. 55a (g) (6) (1) ;"
This technical specification requirement specifically requires that written relief requests be approved by the Commission prior to implementation of the relief request.
In addition to the technical specification requirement, 10 CFR 50.55a (g) (5) (ii) states the following:
"If a revised inservice inspection program for a facility conflicts with the technical spicification for the facility, the licensee shall apply to the Commission for amendment of the technical specifications to conform the technical specification to the revised program.
The licensee shall submit this application, as specified in 50.4, at least 6 months before the start of the period during which the provisions become applicable, as determined by paragraph (g) (4) of this section."
l Attachment I to ET 95-0068 Page 3 of 3 NUREG-1482,
" Guidelines for Inservice Testing at Nuclear Power Plants,"
specifically addresses the situation in which the technical specifications are in conflict with the regulations of 10 CFR 50.55a.
As discussed in NUREG-1482, the NRC staff recognized that situations could arise which would put the licensee in a condition that is not in strict compliance with Technical Specification 4.0.5 requirements to comply with ASME Section XI "except where specific written relief has been granted."
According to the NUREG, if Technical Specification 4.0.5 was interpreted literally, in the case of the Inservice Testing Program, it would require the licensee to address these situations by shutting the plant down to perform testing.
EmEG-1431, " Standard Technical Specifications - Westinghouse Plants," reflects the NRC staff's position that a licensee may establish and implement the Inservice Inspection and Inservice Testing Programs in accordance with 10 CFR 50.55a, and does not require that relief requests be granted before they are implemented.
Rather, according to the NRC staff, 10 CFR 50. 55a (f) (5) (iv) and 10 CFR 50. 55a (g) (5) (iv) allow a licensee up to a full year after a beginning of the updated interval to inform the NRC of those new Code Requirements which cannot be met and to request relief. The regulations require the licensee to submit relief request within 12 months of the interval start date, or during the interval as it finds specific needs for relief.
As stated in NUREG-1482, the NRC recommends that licensees revise the technical specifications to include the recommendations from the revised standard technical specifications (NUREG-1431) for the inservice inspection and testing prograrns.
With the revisions to the technical specifications, upon finding an ASME Code l
requirement impractical because of prohibitive does rates or limitations in the
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l design, cons t ruction, or system configuration, the licensee can implement the relief request at that time.
This implementation could occur provided the relief request has been (1) acceptably reviewed pursuant to 10 CFR 50.59; and (2) approved by the plant staff in accordance with the administrative process described in the inservice inspection and testing programs administrative procedures; and (3) reviewed and approved by the Plant Safety Review Committee.
Although NUREG-1482 does not specifically address the Inservice Inspection Program, the situation is applicable to both the Inservice Inspection Program and Inservice Testing Programs.
By rulemaking effective September 8, 1993 (Federal Register Vol.
57, 34666), the Nuclear Regulatory Commission established (f) to separate the Inservice Testing Program requirements from the Inservice Inspection requirements in paragraph (g) of 10 CFR 50.55a.
By deleting "(g)",
the reference to 10 CFR 50.55a implies both "(f)" and " (g) " requirements and are applicable as appropriate.
Therefore, reference to paragraph "(g)" of 10 CFR 50.55a should be deleted from Technical Specification 4.0.5a and Bases Section 3/4.4.10.
Based on the above discussions and the considerations presented in Attachment II, the proposed change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report; or create a possibility for an accident or malfunction of a different type that any previously evaluated in j
the safety analysis report; or reduce the margin of safety as defined in the basis for any technical specification.
Therefore, the proposed change does not adversely affect or endanger the health or safety of the general public or I
involve a significant safety hazard.
Attschment II to ET 95-0068 Page 1 of 3
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ATTACHMENT II NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
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Attachment II to ET 95-0068 Page 2 of 3 t
No Significant Hazards Consideration Determination i
This license amendment request proposes revising Technical Specification 4.0. 5a and Bases Section 3/4.4.10 to delete the phrase,
"(g),
except where specific l
written relief has been granted by the Commission pursuant to 10 CFR Part 50, j
Section 50. 55a (g) (6) (i). "
This proposed change is consistent with NUREG-1431,
' Standard Technical Specifications - Westinghouse Plants,"
and NUREG-1482,
" Guidelines for Inservice Testing and Nuclear Power Plants."
f Involves a Significant Increase in the Probability or Consequences j
Standard I of an Accident Previously Evaluated This proposed change would remove the wording
...(g),
except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50. 55a (g) (6) (i)."
The Inservice Inspection and Testing Programs are described in the technical specifications pursuant to 10 CFR 50.55a.
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addition, the prcposed change, in accordance with NUREG-1431 and NUREG-1482,
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would provide relief to the ASME Code requirement in the interim between the time of submittal of a relief request until the NRC has issued a safety evaluation and granted the relief.
The change being proposed is administrative in nature and i
does not affect assumptions contained in plant safety analyses, the physical design and/or operation of the plant, nor does it affect any technical specification that preserves safety analysis assumptions.
Any relief from the j
approved ASME Section XI Code requirements will require a 10 CFR.50.59 evaluation i
to ensure no technical specification changes or unreviewed safety questions exist.
Therefore, operation of the facility in accordance with the proposed change would not affect the probability or consequences of.an accident previously 6
analyzed.
Standard II - Create the Possibility of a New or Different Kind of Accident from
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any Previously Evaluated This proposed change would remove the wording
...(g),
except where. specific
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written relief has been granted by the Commission pursuant to 10 CFR Part 50, l
Section 50.55a (g) (6) (i)."
The Inservice Inspection and Testing Programs are described in the technical specifications pursuant to 10 CFR 50.55a.
In addition, the proposed change, in accordance with NUREG-1431 and NUREG-1482, would provide relief to the ASME Code requirement in the interim between the time of submittal of a relief request until the NRC has issued a safety evaluation and granted the relief.
The change being proposed is administrative in nature and will not change the physical plant or the modes of operation defined in the facility license.
The change does not involve the addition or modification of equipment nor does it alter the design or operation of plant systems. Any relief from the approved ASME Section XI Code requirements will require a 10 CFR 50.59 evaluation to ensure no technical specification changes or unreviewed safety questions exist.
Therefore, operation of the facility in accordance with the proposed change would not create the possibility of a new or different kind of accident from any accident previously evaluated.
Attachmant II to ET 95-0050 Page 3 of 3 i
l Standard III - Involve a Significant Reduction in the Margin of Safety The proposed change would remove the wording
...(g),
except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50. 55a (g) (6) (1)."
The Inservice Inspection and Testing Programs are described in the technical specifications pursuant to 10 CFR 50.55a.
In addition, the proposed change, in accordance with NUREG-1431 and NUREG-1482, would provide relief to the ASME Code requirement in the interim between the time of submittal of a relief request until the NRC has issued a safety evaluation and granted the relief.
The change being proposed is administrative in nature and will not alter the bases for assurance that safety-related activities are performed correctly or the basis for any technical specification that is related to the establishment or maintenance of a safety margin.
Any relief from the approved ASME Section XI Code requirements will require a 10 CFR 50.59 evaluation to ensure no technical specification changes or unreviewed safety questions exist.
Therefore, operation of the facility in accordance with the proposed change would not involve a significant reduction in a margin of safety.
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 previous evaluations; or create the possibility of a new or different kind of accident or condition over previous evaluation; 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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Attcchment.III to ET 9?-0068 1
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I ATTACNMENT III ENVIRONMENTAL IMPACT DETERMINATION l
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m Attachmant III to ET 95-0068 Page 2 of 2 Environstantal Impact Determination 10 CFR 51.22(b) specifies the criteria for categorical exclusions from the requirements for a specific environmental assessment per 10 CFR 51.21.
This amendment request meets the criteria specified in 10 CFR 51.22 (c) (9).
The specific criteria contained in this section are discussed below.
(i) the amendment involves no significant hazards consideration Ac demonstrated in the No Significant Hazards Consideration Determination in Attachment II, the requested license amendment does 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 The requested license amendment involves no change to the facility and does not involve any change in the manner of operation of any plant systems. involving the i
generation, collection or processing of radioactive materials or other types of effluents.
Therefore, no increase in the amounts of effluents or new types of effluents would be created.
(iii) there is no significant increase in individual or cumulative occupational radiation exposure The requested license amendment involves no change to the facility and does not involve any change in the manner of operation of any plant systems involving the generation, collection or processing of radioactive materials or other types of effluents.
Furthermore, implementation of this proposed change will not involve work activities which could contribute to occupational radiation exposure.
Therefore, there will be no increase in individual or cumulative occupational radiation exposure associated with this proposed change.
Based on the above it is concluded that there will be no impact on the environment resulting from this change.
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 specific environmental assessment by the commission.
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l Attachment.
i IV to ET 95-0068 Page 1 of 3
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4 ATTACIOutNT IV PROPOSED TECHNICAL SPECIFICATION CHANGES r
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