ML20213D730
| ML20213D730 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/06/1986 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Rubenstein L Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19292G230 | List: |
| References | |
| 86-477A, NUDOCS 8611120209 | |
| Download: ML20213D730 (13) | |
Text
f VIHOINIA ELECTHIC AND POWEH COMI%NY Hictixonn,VinoisiA 2132 61 W.L. STEWART Vara Pase paar NuctmAm OraRATIOma Mr. Harold R. Denton, Director Serial No. 86-477A Office of Nuclear Reactor Regulation E&C/BSD/KKD/psj:2213N Attn: Mr. Lester S. Rubenstein, Director Docket Nos. 50-338 PWR Project Directorate No. 2 50-339 Division of PWR Licensing-A License Nos. NPF-4 U.S. Nuclear Regulatory Commission NPF-7 Washington, D.C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT NOS. 1 AND 2 PROPOSED LICENSE AMENDMENT - GDC 4 Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests an amendment, in the form of a license condition, to Operating License Nos. NPF-4 and NPF-7 for North Anna Units No.1 and 2.
Proposed wording for the license condition is provided in Attachment 1.
Our discussion of the proposed license condition is provided in Attachment 2.
The proposed amendment would add a license condition stating that the design of the reactor coolant pump and steam generator supports may be revised in accordance with this submittai.
This change would permit 18 large-bore snubbers installed solely to mitigate the effects of pipe rupture events to be removed from each unit's primary. loop and six (6) additional large bore snubbers to be removed and replaced with rigid restraints on each unit's primary loop.
It is our intent to implement this redesign during the 1987 North Anna refueling outages by permanently removing certain snubbers since they are required to be active solely to mitigate the dynamic effects of a reactor coolant system pipe rupture event and by replacing certain other snubbers with rigid struts since these snubbers have small thermal movement.
North Anna Unit 1 is currently scheduled to shutdown May 1,1987 for a 59 day refueling outage.
North Anna Unit 2 is currently scheduled to shutdown August 28, 1987 for a 48 day refueling outage.
The evaluations, which support this request, are included as attachments. is a loading evaluation with twenty-four (24) snubbers removed and six (6) rigid restraints installed. Attachment 4 is an evaluation of the Reactor Coolant System (RCS) leakage detection system. Attachment 5 is a North Anna specific fracture mechanics analysis, Westinghouse Electric Corporation report, WCAP-11163, for primary RCS piping.
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n This request has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control Staff.
It has been determined ~that this request does not involve an unreviewed safety question as defined in 10.CFR 50.59 and does not pose a significant hazards consideration as defined in 10 CFR 50.92. We note, however, that in the Comission's response to Issue 5 in the Supplementary Information accompanying publication of the revised GDC-4 (51FR12504), statements are made that eliminating pipe whip restraints would not involve an unreviewed safety question, but that changing support lo~ad path designs would involve an unreviewed safety question.
Because no additional di:cussion is presented describing the F ses for these conclusioris, it is not apparent where our evaluation differs from the NRC's.
However, because of this apparent conflict, it seems prudent to submit this change to the_NRC for~ review and approval.
Your timely advice on your' disposition towards this request is desired to assist us in planning for the upcoming outages.
Enclosed in support of this application are:
1.
10 copies of WCAP-11163, " Technical Bases for Eliminating Large Primary Loop Pipe Rupture as A Structural Design Basis for North Anna Units 1 &
2", Westinghouse Proprietary Class 2, August 1986.
2.
10 copies of WCAP-11164, " Technical Bases for Eliminating Large Primary Loop Pipe Rupture as a Structural Design Basis for North Anna Units 1 &
2", Westinghouhe Non-Proprietary, August 1986.
(Non-Proprietary).
Also enclosed is a Westinghouse authorization letter, (CAW-86-069), Proprietary Information Notice, and accompanying affidavit.
I Since WCAP-11163 contains information proprietary to Westinghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the owner of the information.
The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the Commission's regulations.
Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations.
Correspondence with respect to the proprietary aspects of the Application for Withholding or the supporting Westinghouse Affidavit should reference (CAW-86-069) and should be addressed to R. A. Wiesemann, Manager, Regulatory and Legislative Affairs, Westinghouse Electric Corporation, P. O. Box 355, Pittsburg, Pennsyl"ania 15230-0355.
We are herewith submitting an Application Fee of $150 for a license amendment pursuant to 10 CFR 170.
Very truly yours, W. L. Stewart
Attachments:
1.
Proposed License Condition 2.
Discussion of Proposed License Amendment 3.
Loading Evaluation 4.
RCS Leakage Detection Evaluation j
5.
Fracture Mechanics Evaluation, WCAP-11163 and WCAP-11164
)
6.
Application Fee 7.
Westinghouse Application for Withholding Proprietary Information cc: Dr. J. Nelson Grace Regional Administrator NRC Region II Mr. L. Reyes, Acting Director Division of Reactor Projects NRC Division II Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station Mr. Leon B. Engle NRC North Anna Project Manager PWR Directorate No. 2 Division of PWR Licensing-A l
115-BSD-2213N-a-1
1-COMMONWEALTH OF VIRGINIA )
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CITY OF RICHMOND
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The foregoing document was acknowledged before me, in and 'for the City and Commonwealth aforesaid, today by W.
L.
Stewart who is Vice President Nuclear Operations, of Virginia Electric and Power Company.
He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this h
day of hekwba, 19 $\\_c My Commission expires:
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r ATTACHMENT 1 PROPOSED LICENSE CONDITION FOR OPERATING LICENSE NOS. NPF-4 AND NPF-7 115-BSD-2213N-5 l
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The following condition should be added to Operating License Nos. NPF-4 and NPF-7.
License Condition 2.F.
"The design of the reactor coolant pump and steam generator supports may be revised in accordance with the licensee's submittal dated
, 1986 (Serial No. 86 - 477A)".
115-BSD-2213N-6
ATTACHMENT 2 DISCUSSION OF PROPOSED LICENSE AMENDMENT 115-BSD-2213N-7
Discussion Based on the recent revision to GDC-4 (Reference 1), Virginia Electric and Power Company requests approval for a redesign of the reactor coolant pump and steam generator supports at North Anna Units 1 and 2.
The new GDC-4 eliminates the need for consideration of postulated breaks in the RCS primary loop piping and its associated dynamic and other effects such as pipe whip, jet impingement, asymmetric pressure loading, and primary component sub-compartment pressurization. Approval of this request will also allow us to eliminate certain snubbers which are now required solely to mitigate a pipe rupture event and to replace with rigid restraints certain snubbers which have minimal thermal movement.
Specifically, approval of this request will allow us to redesign the primary reactor coolant loop equipment support:
(1) Eliminate two snubbers per loop acting in a direction perpendicular to the reactor coolant hot leg at the steam generator upper support ring and replace them with rigid struts.
(2) Eliminate two snubbers per loop which are parallel to the reactor coolant cold leg at the reactor coolant pump support.
(3) Eliminate two snubbers per loop which are parallel to the reactor coolant hot leg at the steam generator lower support.
(4) Eliminate two of four snubbers per loop which are cross over restraints between the reactor coolant pump support and the steam generator lower support.
(5) Eliminate the dynamic effect of postulated primary reactor coolant loop breaks from the remaining reactor coolant pump and steam generator supports.
Granting this request would not affect:
ECCS Design Basis.
Reactor containment building and compartment design basis.
Equipment qualification basis.
Engineered safety feature systems response.
It is our intent to implement these changes during the 1987 North Anna refueling outages.
This request is based upon the use of advanced fracture mechanics technology as applied to primary system piping in Westinghouse Electric Corporation topical reports (References 3, 4 and 5 with a by WCAP 11163 and 11164 (Reference 6)pproval by the USNRC in Reference 2), and prepared specifically for North Anna 1 and 2.
Generic Letter 84-04 provided the NRC staff Safety Evaluation Report for analysis of materials submitted for a group of utilities operating PWR's to 115-BSD-2213N-8
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resolve unresolved safety issue A-2.
The staff evaluation concluded that provided certain conditions were met, an acceptable technical basis exists so that asymmetric blowdown loads resulting from large breaks in main coolant loop piping need not be considered as a design basis.
North Anna Units 1 and 2 were not included with the group of plants for which the Unresolved Safety Issue A-2 was addressed. Therefore, to supplement the fracture mechanics studies performed for the A-2 owner's group, a plant specific fracture mechanics study was undertaken for North Anna Units 1 and 2.
Westinghouse reports, WCAP 11163 and 11164, (Reference 6) document the results of the plant specific study.
These Westinghouse reports, WCAP-11163 and 11164, in association with the other references provide a substantial and adequate basis for eliminating postulated breaks in the North Anna Unit 1 and 2 stainless steel reactor coolant system piping. The analyses demonstrate that the probability of rupturing such piping is extremely low under design basis conditions.
The bases for the request for North Anna Units 1 and 2 are as follows:
1.
Extensive operating experience has demonstrated the integrity of the PWR reactor coolant system primary loop including the fact that there has never been a leakage crack.
2.
Pre-service and in-service inspections performed on the RCS piping minimize the possibility of flaws existing in such piping.
The application of advanced fracture mechanics has demonstrated in other applications that if such flaws exist they will not grow to a leakage crack when subjected to the worst case loading condition over the life of the plant.
3.
If a large through-wall flaw is postulated, large margins against unstable crack extension exist for the stainless steel primary coolant piping even if subjected to the safe shutdown earthquake in combination with the loads associated with normal operation.
4.
Units have adequate leakage detection systems such that a postulated reference flaw would yield detectable leakage with margin when subjected to normal operating load condition.
In addition to WCAP 11163 and 11164, which document the plant specific fracture mechanics study, further information is submitted addressing loading evaluation and leakage detection systems evaluation (Attachments 3 and 4).
Unreviewed Safety Question Evaluation The proposed change does not involve an unreviewed safety question, because operation of North Anna Units 1 and 2 in accordance with this change would not:
(1)
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.
As recognized in the GDC-4 revision, the probability of occurrence of an accident is not increased when leak before break technology is properly applied.
Our application of the GDC-4 revision is in accordance with approved guidance. The consequences of an accident previously evaluated are not increased (other than the dynamic effects of a postulated loop 115-BSD-2213N-9 k
rupture, which are now deleted from the design basis by application of leak-before-break technology) because the revision does not affect the ECCS design basis, reactor containment building and compartment design basis,- equipment qualification basis, or engineered safety feature systems response from that previously evaluated. Malfunctions of equipment may actually be reduced.
Elimination of 24 large bore snubbers per unit reduces the number of snubbers that could potentially fail and also eliminates the corresponding maintenance problems associated with maintaining large bore snubbers. Access to other components for inspection and maintenance purposes is also enhanced and total man-Rem exposure is substantially reduced.
(2) Create a possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report.
It has now been demonstrated that one type of accident - the double ended pipe rupture - has been eliminated from the design basis through application of leak before break technology. More specifically, our submittal establishes that certain snubbers we propose to eliminate are required only to mitigate the effects of pipe rupture. Our analyses demonstrate that with redesigned support system, (a) the loading of the primary loop is enveloped by the loads used in the plant specific fracture mechanics study, and (b) the reactor coolant system equipment, piping, and l
supports continue to have acceptable margins of safety under licensed loading conditions other than the now-eliminated RCS main loop rupture, i
The eighteen snubbers which we propose to eliminate have been demonstrated i
to serve only as pipe whip restraints. The six snubbers which we propose to replace with rigid struts exhibit minimal thermal movement.
(3) Reduce the margin of safety as defined in the basis for any technical specification.
Our loading evaluation with the revised support load path configuration (i.e. snubbers removed) establishes that the piping components and supports are stressed within UFSAR acceptable limits. Adequate safety margins exist in a seismic event (in excess of code allowables) and the maximum forces and moments in the reactor coolant loop piping are acceptable based on WCAP 11163.
This WCAP demonstrates the acceptability of North Anna using methods which extend the analysis previously accepted by in the NRC Safety Evaluation dated February 1,1984, which was applicable only to plants in the USI A-2 Owner's Group.
Therefore, pursuant to the criteria specified in 10 CFR 50.59, we conclude that no unreviewed safety question exists.
No Significant Hazards Consideration Determination The proposed change does not involve a significant hazards consideration because operation of North Anna Units 1 and 2 in accordance with this change would not:
(1)
Involve a significant increase in the probability or consequence of an accident previously evaluated.
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We have determined that the probability or consequences of-an accident 4
previously evaluated are not increased by our application of leak before -
"~
break technology. ' Through proper application, we have demonstrated that:
Ta) advanced fracture mechanics analysis is an acceptable alternative to r
maintaining systems.or features solely to mitigate the consequences of
. postulated pipe. ruptures; b) that high safety margins are maintained for the remaining snubbers and rigid restraints as shown in our loading evaluation to mitigate the effects of all postulated events (except the now eliminated pipe rupture); c) that our leakage detection systems-are
' capable of detecting leakage from postulated through-walls flaws and allow operating personnel to take appropriate responses in a timely fashion, and
'd) the ECCS design basis, reactor containment and compartment design basis, equipment qualification basis, and engineered safety feature system response are unaffected by this change.
4 (2) Create the possibility of a new or different kind of accident from any accident previously evaluated.
It has been demonstrated that one type of accident - a postulated RCS pipe rupture - has been eliminated from the design basis through application of leak before break technology. More specifically, our submittals have established that certain snubbers we propose to eliminate are required only to mitigate the effects of pipe rupture. Our anal demonstrated that with the redesigned support system, (yses havea)theloadingo i
the primary loop is enveloped by the North Anna specific analysis which 3
supplements and extends the generic analyses submitted by Westinghouse on behalf of the USI A-2 Owners Group, and approved by the NRC staff in Generic Letter 84-04; and-(b) the reactor coolant system equipment, piping, and supports continue to have acceptable margins of safety under licensed loading conditions other than the now-eliminated RCS main loop i
. rupture.. The eighteen snubbers which we propose to eliminate have been demonstrated to serve only as pipe whip restraints. The six snubbers which-we propose to replace with rigid struts exhibit minimal thermal movement.
L l
(3)
Involve a significant reduction in a margin of safety.
1
~
Our loading evaluation with the revised support load path configuration (i.e. snubbers removed) establishes that the piping components and F
supports are stressed within UFSAR acceptable limits.
Adequate safety margins exist in a seismic event in excess of code allowables and the forces and maximum moment in the reactor coolant loop piping is within the forces and moment used in a plant specific fracture mechanics analysis, which is an extension to an earlier fracture mechanics study approved by the NRC on February 1,1984, for a group of Westinghouse l
plants.
To better illustrate this determination, the Commission has provided examples (51FR7751) of amendments not likely to involve significant hazards considerations.
Example (iv) states:
"A relief granted upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation was not yet demonstrated.
This assumes that the operating restriction and the criteria to be applied to a request for relief have been established in a prior review and that it is justified in a satisfactory way that the criteria have been met."
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The proposed amendment is similar to the example as follows: The installation of pipe whip restraints and other devices (i.e. snubbers) to protect against the dynamic effects of a postulated pipe break is equivalent to the operating restriction in the example. Acceptable operation had not yet been demonstrated.
With the advent of leak before break technology, acceptable operation without this restriction can be demonstrated. The criteria for accepting leak before break technology have been previously accepted by the NRC in their February 1,1984, Safety Evaluation as documented in Generic Letter 84-04, as well as the analyses referenced in the supplementary information which supports the Commission action to modify General Design Criterion 4 to eliminate this " operating restriction" from the design basis.
Therefore, pursuant to the standards in 10 CFR 50.92 we have determined that this change involves no significant hazards consideration.
115-BSD-2213N-12
References:
1.
Modification of General Design Criterion 4, requirements for protection against Dynamic Effects of Postulated Pipe Ruptures; 51 FR 12502.
2.
Generic Letter 84-04, February 1,1984, " Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."
3.
WCAP-9558, Revision 2 (May 1981) " Mechanistic Fracture Evaluation of Reector Coolant Pipe Containing a Postulated Circumferential Through-wall Crack."
4.
WCAP-9787 (May 1981) " Tensile and Toughness Properties of Primary Piping Weld Metal for use in Mechanistic Fracture Evaluation."
5.
Letter Report NS-EPR-2519, E.P. Rahe to D. G. Eisenhut (November 10,1981)
Westinghouse Response to Questions and Comments Raised by Members of ACRS Subcommittee on Metal Components During the Westinghouse Presentation on September 25, 1981.
6.
WCAP 11163 Westinghouse Proprietary Class 2 and WCAP 11164 Westinghouse Proprietary Class 3; Technical basis for eliminating large primary loop pipe rupture as a structural design basis for North Anna Units 1 and 2.
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