ML20080K546
| ML20080K546 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/23/1995 |
| From: | Wadley M NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20080K538 | List: |
| References | |
| NUDOCS 9503010165 | |
| Download: ML20080K546 (11) | |
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UNITED STATES NUCLEAR REGUIATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISIAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED February 23, 1995 Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Prairie Island Operating License, Appendix A as shown on the attachments labeled Exhibits A, B, and C. Exhibit A describes the specific exemption, proposed changes, reasons for the changes, and the supporting safety evaluation /significant hazards determination. Exhibit B contains the current Prairie Island Technical Specification page marked up to show the proposed changes. Exhibit C contains the revised Technical Specification page.
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPA.NY By sf$A(/}ds ALl M. D. Wadley Plant Manager Prairie Island clear Generating Plant On this h y of ore me a notary public in and for said County, personally appeared M . Wadley, Plant Manager, Prairie Island Nuclear Generating Plant, and eing first duly sworn acknowled F ed that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.
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LICENSE AMENDMENT REQUEST DATED February 23, 1995 Specific Exemption from 10CFR50, Appendix J
.Tvoe A Periodic Retest Schedule Reauirements for Unit II EXHIBIT A Description of the Specific Exemption, Proposed Changes, The Reasons for Requesting the Changes, and the Supporting-Safety Evaluation /Significant Hazards Determination Pursuant to 10 CFR Part 50, Sections 50.12, 50.59 and 50.90, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following specific exemption from these regulations and changes to the Facility Operating i Licenses and Appendix A, Technical Specifications:
INTRODUCTTON The purpose of this request is for two separate but related licensing actions.
A license amendment is request in accordance with the provisions of 10CFR50.90 would revise the wording in the Prairie Island Technical Specifications to allow implementation of exemptions to the schedule requirements of 10CFR50, Appendix J.
i The request for specific exemption in accordance with the provisions of l 10CFR50.12 is to request temporary relief for Prairie Island Nuclear Generating Plant Unit 11 from the requirements of 10CFR 50, Appendix J.
Exemption is requested from Section III, D,1(a) of Appendix J which states, "After the preoperational leakage rate tests, a set of three Type A tests shall be performed, at approximately equal intervals during each 10-year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year plant inservice inspections."
LICENSE AMENDMENT REOUEST
Background
This requested change to the Prairie Island Technical Specifications is administrative in nature and will allow Prairie Island to implement exemptions to Appendix J schedule requirements which may be granted by the Nuclear Regulatory Commission. As presently written, the Technical Specifications require compliance with Appendix J as written without allowance for any exemptions.
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Pronosed'Channes and Reasons for Channes ..
The proposed changes to Prairie Island Operating License Appendix A, Technical- ~f Specifications are described below, and the specific wording changes are shown
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in Exhibits B and C, t SURVEILIANCE REOUIREMENTS . Pararraoh 4.4. A. 5: Add the phrase, "and all j approved exemptions." after "/gpendix J". ;
Justification: This wording addition is necessary for the Technical !
Specifications to allow implementation of exemptions from the testing j schedule requirements of 10CFR50 Appendix J Section III,.D,'1 (a).
1 Safety Evaluation ~!
i When Appendix J to 10CFR50 was written into the regulations,-the nuclear i industry had limited experience with containment performance and testing. The j ratesting schedular requirements of Appendix J were based on the best ,
judgement at that. time for providing assurance that containment integrity would be maintained. Since then many nuclear plants have been built and I operated for many years and maty containment tests have.been performed.'This i experience provides the basis f understanding how containment integrity will l be assured from test to test ans allow adjusting test schedules to meet ,
specific needs.
This request for license amendment does not in and of itself change any test' ;
schedules, it is only an administrative change that would allow schedules to j be changed when the Nuclear Regulatory Commission grants an exemption. Each 6 request for specific exemption would require evaluation on its own merits to i determine its safety impact. Therefore, this amendment does not introduce any l safety concerns.
Determination of Sirnificant Hazards considerations !
The proposed changes to the. Operating License have been evaluated to determine l whether they constitute a significant hazards consideration as required by 10 -l CFR Part 50, Section 50.91 using the standards provided in Section 50.92. This 1 analysis is provided below
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- 1. The proposed amendment will not involve a significant increase in '
the erobability or conseauences of an accident creviousiv evaluated The proposed amendment is an administrative change which allows l implementation of approved exemptions to the regulations and by itself j does not change any retest schedules. !
1 Therefore, the probability or consequences of an accident previously !
evaluated are not affected by the proposed amendment. I Page 2 i
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- 2. The proposed ameridsent will not create the possibility.of a !
new or different kind of accident from any accident previously- j analyzed '
The proposed amendment is'an administrative change which allows l implementation of approved exemptions to the regulations and by_itself '!
does not change any ratest schedules. - _l t Therefore,-the possibility'of a new or different kind'of accident from i any accident previously evaluated would not be created by the' proposed-amendment. !
- 3. : The proposed amendment will not involve a. significant reduction in f the marrin of safety _!
The proposed amendment is an administrative change which allows
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4 implementation of approved exemptions to the regulations and by itself does not' change any retest schedules. (
.. i Therefore, a significant reduction in the margin of safety would notL ~!
be involved with.the proposed amendment.
l Based on the evaluation described above, and pursuant to 10 CFR Part 50, Section 50.91, Northern States Power Company has determined that operation the Prairie Island Nuclear Generating = Plant in accordance'with the proposed- 4 license amendment request does not involve any significant hazards. :
considerations as defined by Nuclear Regulatory Commission regulations in-10' li 4 CFR Part 50, Section 50.92.
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REQUEST FOR SPECIFIC EXEMPTION l Backaround .l Prairie Island Nuclear Generating _ Plant Unit II is scheduled to shut._down for.
refueling May 12, 1995. Currently the third Type A test'of the second 10-year' j service period is scheduled to be performed concurrent with the completion of ,
the 10-year inservice inspections. This request for specific exemption seeks i relief from performance of the Type A test during 1995 and in lieu Northern I States' Power Company proposes to perform the Type A test during the next Unit i II refueling outage.' As demonstrated below, this. proposed relief from the j regulation schedular requirements will allow the underlying purpose of the- _ ';
regulation to be achieved as required by 10CFR50.12(a)(2)(ii). Postponing the -l 1995 Type A may, as also discussed below, benefit Northern States Power- _
l Company's stakeholders through reduced radiation exposure, reduced probability .;
of valve misalignment,.more efficient use of plant resources and' increased, ;
generation of low cost electricity, '{
The containment vessel completely encloses the reactor coolant system.to ensure no leakage of radioactive materials to the environment in the unlikely Page 3
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event of a. loss of coolant accident. The containment system incorporates a l' free-standing containment vessel surrounded by a. low leakage reinforced =
! concrete shield building Limitations on containment leakage rates ensure that
. 'the total containment leakage volume will not exceed the value assumed in the
~ safety analyses at peak post accident containment pressure (P ). ;
Safety Evaluation l Prairie Island Nuclear Generating Plant Technical Specification 4.4.A.1.d' specifies that the containment leakage rate (Lg) shall be 0.25 weight percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at pressure P.. To, assure that the Prairie Island containments i meet these Technical Specifications, Type A integrated leakrate tests have l been performed prior to startup of the plant in compliance with the technical j
. and schedular requirements of 10CFR50 Appendix J. Appendix J allows testing l at a' reduced pressure, Pe which is not less than one-half of P., A
- corresponding maximum allowable leakage rate, Lg is also defined as the.
' acceptance. criteria at the test pressure,-Pg. As an added conservatism the ;
measured overall integrated leakage rate test acceptance criteria for as-left -! '
tests is further limited to less than or equal to 0.751s.
The Prairie Island Unit II preoperational Type A test was performed August }
1974. For the reduced pressure test, the leakage characteristics of the i preoperational Type A test yielded an allowable measured leakage rate, Ig., as 0.176784/ day for subsequent ratests at Ps of 23 psig. ,
t The Type A retests at Prairie-Island have been conducted at the end of the designated outage just prior to plant startup. These "as-left" tests give an added level of safety in that the condition of the containment is known to be ,
leak tight as the plant starts its operating cycle.
Prior to the Type A test, Type B and C tests are conducted each refueling outage which demonstrate the "as-found" leakage of penetrations and seals.
Since 1985 Prairie Island has been using the Type B and C as-found test results in conjunction with the as left Type A test results to determine backfit "as-found" Type A test results.
As a matter of management philosophy throughout the plant's operating history, comprehensive repairs or modifications have been performed on valves, penetrations or seals contributing significantly to Type B or C test leakage results. Any valve, penetration or seal which is worked upon during the outage is retested to establish an as-left leakage rate. The subsequent Type A test then benefits from the maintenance or modifications which have been performed.
The Prairie Island containments have an excellent Type A test performance history that provides substantial justification for the proposed schedular relief. The three as-found Type A results determined since 1985 have consistently been less than 28% of the allowable test leakage rate, Lg.
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. 1 The following is a brief history of Prairie Island Nuclear Generating Plant Unit II Type A testing:
The first periodic retest was successfully completed in December 1977.
A least squares fit to the 1977 test data yielded a leakage rate of 0.0583t/ day and 95% upper confidence level as-left Lu, of 0.0628 t/ day.
No containment modifications affecting containment integrity were performed subsequent to the previous test.
The second periodic retest was successfully completed in March 1981. A least squares fit to the 1981 test data yielded a leakage rate of 0.0147%/ day and a corresponding 95% upper confidence level as-left is, of 0.0206%/ day. Two electrical containment penetrations affecting containment integrity were installed in accordance with the site modification control process subsequent to the previous Type A test, prior to this test.
The third periodic retest was successfully completed in October 1985.
A least squares fit to the 1985 test data yielded a leakage rate of 0.0367t/ day and a corresponding 95% upper confidence level ac-left L.
of 0.0402t/ day. The as-found Type A test results were determined to be 0.0482%/ day. Various modifications which affected containment integrity were performed subsequent to the previous Type A test, prior to this test. These modifications, described in Northern States Poser Company report to the Nuclear Regulatory Commission entitled, Reactor Containment Building Integrated Leak Rate Test, October 1985, were performed in acecraance with the site modification control process. A couple of these Lc41fications were made specifically to improve containment integrity.
The fourth periodic ratest was successfully completed in April 1989. A least squares fit to the 1989 test data yielded a leakage rate ci 0.0266t/ day and c srresponding 95% upper confidence level Is, of 0.0272%/ day. The as-found Type A test results were determined to be 0.03689t/ day. w: containment modifications affectin5 containment integrity wtrr, r rformed subsequent to the previous test.
The fifth and last periodic retest to date was successfully completed in January 1993. A least squares fit to the 1993 test data yielded a leakage rate of 0.01236/ day and a corresponding 95% upper confidence level Lun of 0.0158t/ day. The as-found Type A test results were determined to be 0.0307t/ day. No containment modifications affecting containment integrity were performed subsequent to the previous test.
It. surmary, all Prairie Island Unit II Type A tests were completcd successfully without any significant incidents and the highest as-found leakage rate was determined to be less than 28% of 1s. The as-left results of the first Type A retest was less than 36% of Ls. Through improvements in valves, penetrations and seals, the as-left Type A leakage rate has been ;
consistently reduced to less than 23% of 1s. l 1
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. . i Many modifications have been performed on the containment pressure boundary, l however they were properly controlled by the site modification process and !
subsequent Type A tests demonstrated they did not adversely affect containment j integrity. Many of these modifications were made specifically to improve containment leak tightness based on Type B and C test results. These modifications included replacement of valves and blanking penetrations with blind flanges to remove valves from consideration as sources of containment !
leakage. '
Unit I Type A test results were also evaluated for performance characteristics in support of this exemption request. Overall, Prairie Island Type A test ;
leakage ratca for both units have been consistent and very low. Likewise Type ;
B and C test results for both units have been consistently low.
Factors affecting leak tightness of the containment may be categorized as active components such as valves and senis which are leak rate tested by Type A and B or C tests or passive components which constitutes the containment r structure and are tested only during the Type A test. ,
Active components such as valves and seals are tested during refueling outages ,
in accordance with the appropriate Type B or C test. These components are again leak rate tested when containment is subjected to a Type A test.
Industry experience indicates that the failures associated with Type A test are generally found on active components which receive Type B or C testing.
Prairie Island has aggressively pursued modifications to reduce containment penetration leakage based on Type B and C test results. Therefore, continued overall leak tightness of the active containment components can be assured by the Plant's reliable Type B and C testin5 Program.
Two mechanisms could adversely affect the passive structural capability of '
containment. First, modifications could be made to the structure which, if not carefully controlled, could leave the structure with reduced capability.
Second, deterioration of the structure may be caused by pressure, temperature, radiation, chemical or other effects.
Prairie Island will not be making any major modifications to the containment structure itself during the 1995 refueling outage. Modifications to close two penetrations with seal welded caps may be made to the passive structure during ;
1995. The penetrations under consideration have historically contributed significantly to the Type C test leakage. If performed, the modifications will be controlled by the Prairie Island modification process and subjected to l 1eakage rate testing prior to acceptance of the modification. The i effectiveness of this process has been demonstrated by past modifications t which did not adversely affect any Type A test results.
Absent actual accident conditions, structural deterioration is a gradual phenomenon which requires periods of time well in excess of the approximately four year (January 1993 to January 1997) untested interval proposed by this request for specific exemption. We are unaware of any relatively quick acting degradation mechanisms which could adversely affect containment integrity in a relatively short four year interval. Other than accident conditions which have never occurred at Prairie Island, the only significant pressure challenge to Page 6
the containment structure is the Type A test itself. Postponing this test ;
could therefore lessen the potential for adverse pressure effects. Prairie i Island has previously operated over a 55 month period between Type A tests j without any apparent degradation. ;
t 10CFR50 Appendix J Section V.A. requires a general inspection of the l' accessible interior and exterior surfaces of the containment structures and components to be performed prior to any Type A test to uncover any evidence of l structural deterioration which may affect either the containment structural ;
integrity or leak tightness. At Prairie Island there has been no evidence of l structural deterioration that would impact structural integrity or leak tightnesa. l Also at issue in this request for specific exemption is decoupling the Type A test from the 10-year Inservice Inspection Program. The 10-year Inservice Inspection program is a series of inspections performed every 10 years in j accordance with Section XI of the ASME Boiler and Pressure Vessel Code and -
Addenda as required by 10CFR50.55a. j There is no benefit in coupling the requirements of the 10-year Inservice Inspection program with those for performing Type A leakage rate tests. Each of these two test programs, that is the containment Type A test and the 10-year Inservice Inspection program, is independent of the other and provides assurances of different plant characteristics. The Type A test assures the 7 required leak tightness for the containment to demonstrate the leakage rate is less than assumed in accident analyses demonstrating compliance with 10CFR100.
The 10 year Inservice Inspection program provides assurance of integrity of !
the plant structures, systems, and components in compliance with 10CFR50.55a.
There is no safety related concern necessitating their coupling to the same refueling outage. .
l Exemption from testing Prairie Island Unit II Type A test in 1995 is based on: l l
Prairie Island Unit II containment has consistently had excellent Type A tcst results at a fraction of the allowable leakage rate. ;
The Type B and C testing will be performed in the Unit II 1995 outage ;
and will detect primary sources of containment leakage that would i compromise containment integrity.
The exemption will allow a Type A test interval of approximately 48 months which is less than previous test intervals as allowed by i Appendix J.
Problem penetrations have been aggressively modified based on Type B f and C test results to eliminate sources of leakage.
The containment passive structure will remain structurally sound. i t
Modifications affecting containment integrity will be controlled under the Prairie Island modification process and subjected to leakage rate ;
testing. l I
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The Appendix J 1eakage rate testing and Inservice Inspection programs test different components and there is no technical reason to couple the two programs to the same refueling outage.
Compliance with 10CFR50.12 Specific Exemotions The following information demonstrates compliance with 10CFR50.12 which provides guidance for granting exemptions from the regulations:
- 1. Paragraph (a)(1): The exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security.
The proposed changes are not in violation of any applicable law, will not present an undue risk to the public health and safety and would be consistent with the common defense and security.
As discussed above, the proposed exemption does not present undue risk to the public because 1.) Type B and C tests will be performed which will test the containment components most likely to cause Type A test failure; 2.) modificatione affecting containment integrity will be controlled and tested in accordance with the site modification program; and 3.) there is no known mechanism for failure of the basic containment structure during the test proposed interval.
Additionally, the proposed exemption does not present any undue risk because the test interval proposed, approximately 48 months, is less than or comparable to previous test intervals at Prairie Island. In fact Prairie Island has previously tested Unit II containment on a 55 month interval (March 1981 to October 1985) without any detrimental effects.
As discussed previously, there is no technical benefit to performing the containment Type A test during the same refueling outage as the 10-year Inservice Inspection since these test programs test different independent components.
The Prairie land Nuclear Generating Plant is one of Northern States Power Company's lowest cost generators of electricity. Postponin5 the Type A test will enhance the common defense and security by allowing Prairie Island Nuclear Generating Plant to shorten the 1995 outage by at least three days and resume generation of low cost, environmentally clean electricity sooner.
- 2. Paragraph (a)(2)(ii): The underlying purpose of the regulation is achieved.
The underlying purpose of 10CFR50, Appendix J is achieved.10CFR50 Appendix J states that the leakage test requirements set forth in Appendix J provide for periodic verification by tests of the tight integrity of the primary reactor containment. The appendix further states that the purpose of the tests are to assure that leakage Page 8
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through the primary reactor containment shall not exceed the allowable i leakage rate values as specified in the technical specifications or !
associated bases, j As discussed above, the Prairie Island Unit II containment leakage j rate has historically been very low and a Type A test has never 'l failed. Type B and C test results have also been consistently low. The j success of the Type A test program is attributable-to (1) the Type B -[
. and C testing program and associated corrective actions and (2) the t modification control program when modifications have been made which !
7 affect containment integrity, j
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The leak tightness of containment will continue to be met if an !
exemption is granted through the Type B and C. tests, the modification
. -l control program and the integrity of the containment structure itself. l The test interval allowed by this exemption,'approximately 48 months, l is less than the Unit II 55 month test interval from March 1981 to !
October 1985 allowed under Appendix J. The as-found containment j leakage rate after this interval was determined to be less than 284 of -
the maximum allowable leakage.
- 3. Pararrach (a)(2)(iii): Comoliance will result in undue hardshin. l l
Performance of a Type A test on Prairie Island Unit II containment in l 1995 will require at least three days of critical path outage time ;
which results in three days of lost electricity production. !
In order to perform the Type A test, the containment is closed and all containment work activities and equipment operations are suspended. .;
Generally the outage personnel working on containment tasks cannot be ;
quickly and efficiently utilized on other tasks during the three or ~;
more days the test is being performed. This inefficient use of l resources would be eliminated by postponing the test. !
As discussed above, a Type A test is unnecessary to assure the leak tightness of containment. Performance of an unnecessary test resulting ,
in unnecessary costs represents an undue hardship on Northern States ;
Power Company. l 4 Paragraph (a)(2)(iv): Exemption would benefit the public health'and [
safety. compenssting for any decease in safety that may result. i i
Postponing the Prairie Island Unit II 1995 Type A containment leakage ;
rate test does not decrease the safety of the plant. The resulting ;
interval of 48 months is comparable to some of the previous test !
intervals and is less than the longest interval of 55 months which i exhibited no adverse containment leak tightness. -l Granting this exemption will benefit the public health and safety. [
Ultimately postponing the 1995 Type A test may result in less tests ;
being performed over the remaining life of the plant. Therefore. l Page 9 i
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postponing this Type A test may eliminate the radiation exposure associated with valve line-ups, inspections and other outage '
activities performed during the test. Also, there is some small probability for operator error concomitant with realigning the ;
containment isolation valves after the test is complete, postponing of j the test may reduce this probability further.
- 5. Pararraoh (a)(2)(v): Temporary relief from avolicable rerulation.
This request for specific exemption proposes temporary relief from the l requirement to perform a Type A test on Prairie Island Unit II in 1995 i in accordance with the requirements of 10CFR50 Appendix J, Section III, D and in lieu perform the test at the next scheduled refueling. ;
outage. ,
Based on the evaluation described above, and pursuant to 10 CFR Part 50, Section 50.12, Northern States Power Company has determined that-Prairie i Island Nuclear Generating Plant Unit II complies with the requirements for i granting specific exemption for temporary relief from 10CFR50 Appendix J, Section III, D, 1(a) as it applies to the scheduled 1995 Type A retest. l ENVIRONMENTAL ASSESSMENT Northern States Power Company has evaluated the proposed changes and exemption f and determined that-P
- 1. The changes and exemption do not involve a significant hazards ,
consideration, or j i
- 2. The changes and exemption do not involve a significant change in the types or significant increase in the amounts of any effluents that may be .;
released offsite, or ;
- 3. The changes and exemption do not involve a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51 Section 51.22(c)(9).
Therefore, pursuant to 10 CFR Part 51 Section 51.22(b), an environmental assessment of the proposed changes is not required. ,
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