3F0597-21, TS Change Request 212,Rev 1 Replacing Prescriptive Requirements of 10CFR50,App J,Option a w/performance-based Approach to Leakage Testing Contained in 10CFR50,App J, Option B
| ML20138H865 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/01/1997 |
| From: | Richard Anderson FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20138H870 | List: |
| References | |
| RTR-REGGD-01.163, RTR-REGGD-1.163 3F0597-21, 3F597-21, NUDOCS 9705070308 | |
| Download: ML20138H865 (12) | |
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Florida Power CORPORATION NEa5 May 1, 1997 3F0597-21 i
U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C. 20555-0001
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Subject:
Technical Specification Change Request Notice No. 212, Revision 1
Reference:
A.
FPC to NRC letter, 3F0297-09, dated February 17, 1997 B.
FPC to NRC letter, 3F0595-02, dated May 19, 1995 C.
NRC to FPC letter, 3N0995-12, dated September 29, 1995
Dear Sir:
The purpose of this letter is to submit Revision 1 to Technical Specification Change Request Notice No. 212 requesting amendment to Operating License fD. DPR-72. The initial request (Reference A) was submitted to propose changes to the Crystal River Unit 3 (CR-3)
Technical Specifications (TS) to replace the prescriptive requirements of 10 CFR 50, Appendix J, Option A with the performance-based approach to leakage testing contained in 10 CFR 50, Appendix J, Option B.
Florida Power Corporation intends to apply Option B methods to Types A, B and C testing.
TSCRN No. 212, Revision 1 is being provided to include in the Administrative Controls definition of the Containment Leakage Rate Testing Program the numeric value of the calculated design basis loss-of-coolant accident peak containment pressure P, for performing leak rate testing. This revision also provides the leakage rate acceptance criteria for Types A, B and C testing in the Administrative Controls definition. This information is provided to make the proposed Technical Specifications consistent with the Technical Specification Model published by the NRC in a letter to the Nuclear Energy Institute (NEI) dated November 2, 1995.
This revised submittal contains additional editorial changes which have been made in Technical Specifications SR 3.6.1.1, SR 3.6.2.1, SR 3.6.3.2, SR 3.6.3.6 and several Bases paces. The attached TSCRN identifies all changes in the Summary of Changes list which precedes the proposed Technical Specification pages. This revised submittal replaces Revision 0 in its entirety.
for performing the leak rate testing and the leakage rate Adding the numeric value of P*inistrative Controls definition of the Containment Leakage acceptance criteria to the Adrn Rate Testing Program is a clarification of information previously submitted and does not change the intent of the original submittal. The " Proposed Determination of No
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Significant Hazards" and the " Environmental Impact Evaluation" previously provided with I
Revision 0 have been reviewed and remain valid with no changes, ly
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t U. S. Nuclear Regulatory commission 3F0597-21 Page 2 The CR-3 Containment Leakage Rate Testing Program, described in proposed Technical Specification (TS) 5.6.2.20, will be in accordance with methods approved by Regulatory Guide (RG) 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.
FPC is not proposing any deviations from the guidance in the RG.
Revision 0 of the submittal explained that the testing frequency for the 48 inch containment purge valves with resilient seals will not change. The commitment to endorse the guidance of RG 1.163 without exception is better described in this revised submittal for clarification.
Furthermore, following the guidance of RG 1.163, the interval between Type C tests of the 6 inch Post-Accident Hydrogen Purge Valves identified by FPC tag numbers LRV-70, LRV-71, LRV-72 and LRV-73, will be limited to a maximum of 30 months.
FPC is requesting your review and approval of this TSCRN by July 7,1997 to support i
timely closure of Restart Issue R-10. FPC also requests that the implementation date for this Technical Specification amendment be 90 days from the date of approval.
t Sincerely,
<f<
-Ro A. Anderson Se ior Vice President Nu lear Operations l
1 RAA/LVC Attachment xc:
Regional Administrator, Region 11 Senior Resident Inspector NRR Project Manager t
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U. S. Nuclear Regulatory commission
.3F0597-21 Page 3 STATE OF FLORIDA COUNTY OF CITRUS j-Roy A. Anderson states that he is the Senior Vice President, Nuclear Operations for Florida Power Corporation; that he is authorized on the part of said company to sign and i
file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
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/fEoy Anderson i Sen or Vice President g
ear Operations Sworn to and subscribed before me this l5fday of May,1997, by Roy A. Anderson, who is personally known to me.
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'O Signature of Notary Public State of Florida l(~
LYNNE S. SMITH
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$l EXPtfE!B: December 18,1900 gij g;# sammam num my Stainp comniissioned Name of Notary Public i
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U. S. Nuclear Regulatory commission
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3F0597-21' 1
Page 4
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION LIN THE MATTER
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DOCKET N0. 50-302 FLORIDA POWER CORPORATION
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CERTIFICATE OF SERVICE deposes and says that the following has been served on the Designated State Representative and Chief Executive of Citrus County, Florida, by deposit in the United States mail, addressed as follows:
- Chairman, Administrator, Board of County Commissioners Radiological Health Services of Citrus County Department of Health and-Citrus County Courthouse Rehabilitative Services Inverness, FL 34450 1323 Winewood Blvd.
Tallahassee, FL 32301 l
A copy of Technical Specification Change Request Notice No. 212, Revision 1.
FLORIDA POWER CORPORATION 1
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oy A. Anderson Sen'or Vice President Nuc' ear Operations Sworn to and subscribed before me this day of May, 1997, by Roy A. Anderson, who is personally known to me.
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--m Signature of Notary Public State of Florida l
l LMS. SMITH MYCOMemB840Nf 00514300 k..
EXPNB: December 18,1000 MIF seedmummencumwwe Stanp commissionea wame of Notary Public
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l U. S. Nuclear Regulatory commission 3F0597-21 Page 5 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 REQUEST NO. 212, REVISION 1 CONTAINMENT LEAKAGE RATE PROGRAM 10 CFR 50, APPENDIX J, OPTION B LICENSE DOCUMENT INVOLVED:
Technical Specifications j
PORTIONS:*
Technical Specification 3.6.1 Technical Specification 3.6.2 Technical Specification 3.6.3 Technical Specification Section 5.0 Technical Specification BASES B3.0.2 Technical Specification BASES B3.6.1 Technical Specification BASES B3.6.2 Technical Specification BASES B3.6.3
- See
SUMMARY
OF CHANGES for specific Technical Specification changes DESCRIPTION OF REQUEST:
The proposed changes to the Crystal River Unit 3 (CR-3) Technical Specifications (TS) are to adopt 10 CFR 50, Appendix J, Option B for Types A, B and C containment leakage testing.
This option gives the licensee the opportunity to voluntarily change from prescriptive testing requirements to performance-based testing based on the leakage rate testing history of the containment and components.
REASON FOR REQUEST:
The Nuclear Regulatory Commission (NRC) amended 10 CFR 50, Appendix J to include Option B, " Performance-Based Requirements."Section V.B.2 of Option B requires licensees who wish to voluntarily replace the prescriptive testing requirements (0ption A) with performance-based testing requirements (Option B), submit to the NRC a request for revision to the Technical Specifications.
i EVALUATION OF REQUEST:
The proposed changes to the CR-3 TS are to adopt 10 CFR 50, Appendix J, Option B for Types A, B and C containment leakage testing.
Compliance with 10 CFR 50, Appendix J, Option B provides assurance that the primary containment, and those systems and l
components which penetrate the primary containment, do not exceed the allowable leakage rate values specified in the Containment Leakage Rate Testing Program.
The allowable leakage rate is determined so that the leakage assumed in the safety analyses for a design basis accident is not exceeded.
As described in 10 CFR 50, Appendix J, Option B, deviations to the testing requirements, detailed in Appendix J and from methods endorsed by Regulatory Guide (RG) 1.163, must be justified. FPC does not intend to deviate from the requirements in 10 CFR 50, Appendix J, Option B or from the methods endorsed by RG 1.163. The current exemption, approved by the NRC (References B and C), was a schedular exemption and, upon approval of this proposed TS amendment, this exemption will no longer be necessary.
Consistent with the guidance in RG 1.163, this TS amendment is not intended to change the testing frequency for the 48 inch containment purge valves with resilient seals. The TS
U. S. Nuclear Regulatory commission 3F0597-21 Page 6 commitments for surveillance and testing of these valves is separate from Appendix J required tests.
The current surveillance requirements in the TS will remain valid for the 4B inch. containment purge valves. The frequency for the Type C testing of the 6 inch Post Accident Hydrogen Purge valves identified by FPC tag numbers LRV-70, LRV-71, LRV-72, and LRV-73, will be limited to a maximum of 30 months.
BASES FOR THE TECHNICAL SPECIFICATION CHANGE REQUEST:
CR-3 has historically demonstrated satisfactory Type A, B, and C leakage rate tests on the containment and components. These results demonstrate leak-tightness and provide a high degree of confidence that the containment will not significantly degrade due to extended test intervals.
Experience demonstrates that essentially all containment leakage can be detected by local leakage rate tests (LLRTs) Type B and C.
As noted in NUREG-1493, " Performance-Based Leak Rate Testing Program," of 180 ILRTs reported, which covered 110 individual reactors and i
approximately 770 years of cumulative operating history, cnly 5 integrated leakage rate test (ILRT) failures were found that LLRTs could not detect.
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This TS' amendment does not affect the methodology of performing Type A, B and C testing at CR-3 and would not contribute to the degradation of containment integrity.
The leakage rates observed at CR-3 during Type B and C tests have been consistently below the allowable leakage rates as described in Appendix J. The combined "as-found" leakage in each test (from 1990 to 1996) was less than the allowable combined Type B and C leakage rate of 0.6 La (266,431 SCCM) at the calculated maximum peak containment pressure.
The "as-left" leakage rates were also significantly below the allowable limits.
Since 1987 testing, the "as-left" leakage rate has trended downward.
This downward trend highlights FPC's commitment to repair and maintain components that may affect the overall containment integrity.
FPC has already invoked administrative limits on valve leakage rates based on valve opening size.
If valves fail to meet those administrative limits, then the valves are repaired or replaced and re-;ested.
Repairs and replacement are in accordance with Appendix J and CR-3's Inservice Testing and Repair ana Replacement Program, as described in the American Society of Mechanical Engineers (ASME) Code,Section XI.
STRUCTURAL CAPABILITY OF THE CONTAINMENT:
The CR-3 containment is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow dome roof. The cylinder wall is pre-stressed, with a post-tensioning system in the vertical and horizontal directions.
The dome roof is pre-i stressed using a three way post-tensioning system. The inside surface of the containment is a carbon steel liner. The liner plate nominal thickness is 3/8 inch for the cylinder and dome and 1/4 inch for the base. The liner plate is anchored to the concrete. Piping penetrations have been designed to ensure that the liner will not be breached due to the rupture of the piping. During Refuel 10, a visual inspection of the containment and containment liner was performed by FPC. There were no findings of' deterioration that could affect the leak tightness or structural integrity of containment.
RISK IMPACT ASSESSMENT:
The purpose of containment leakage rate testing is to detect an unacceptable trend in containment leakage before a design basis accident occurs. Containment leakage, caused by the degredation of sealing material within containment penetrations and containment
O. S. Nuclear Regulatory commission 3F0597-21 Page 7 isolation components, will continue to be effectively measured by the Type B and C testing performed in accordance with Option B.
The only potential failure not covered by Type B and C testing is a failure of the containment due to structural deterioration caused by pressure or temperature excursions between tests.
Under normal conditions, there are no significant environmental or operational stresses which could contribute to the degradation of the containment structure. Additionally, visual inspections performed in accordance with Appendix J, Option B, provide the opportunity to detect structural deterioration that may be present.
The results of a sensitivity study exploring the risk impact of several alternate leak rate testing schedules are found in NUREG-1493. The NUREG concludes that the impact on risk due to decreasing Type A, B and C testing frequency is very low and that a decrease of ILRT frequency to once per ten years would not significantly increase the risk of containment failure.
This TS amendment submittal is bounded by the analyses in NUREG-1493.
Based on the historical data, the risk increase due to the adoption of Appendix J, Option B, at CR-3, is negligible.
BENEFITS:
ALARA The estimated radiation exposure for performing each ILRT is approximately 4 ManREM. The estimated dose savings by not performing the additional ILRTs is 16 ManREM, over the remaining life of the plant.
The estimated dose for performing each set of LLRTs is approximately 2 ManREM. The dose savings for the reduction of LLRTs is 20 ManREM, over the remaining life of the plant.
i Cost i
The anticipated cost savings for implementing Appendix J, Option B, is attributed to reducing the number of remaining ILRTs from six to two and LLRTs from twenty to ten. The cost to perform each set of LLRTs is approximately $60,000.
The savings over the remaining life of the plant is expected to be $600,000. The approximate cost to perform one ILRT is $2 million, which over the remaining life of the plant, would be a savings j
of $ 8 million.
U. S. Nuclear Regulatory commission 3F0597-21 Page 8 d
DETERNINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION (SHOLLY):
An evaluation of the proposed TS amendment has been performed in accordance with 10 CFR 50.91(a)(1) regarding significant hazards considerations, using the standards in 10 CFR 50.92(c).
Criterion 1 Does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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The TS amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated, i
The proposed changes to the TS are to implement Option B of 10 CFR 50, Appendix J, at CR-3.
The proposed changes will result in increased intervals between containment leakage i
tests based on the leakage rate testing history. The proposed changes do not involve a change to the plant design or operation and does not change the testing methodology.
NUREG-1493, " Performance-Based Containment Leak-Test Program," provides the technical basis of 10 CFR 50, Appendix J, Option B.
NUREG-1493 contains a detailed evaluation of the expected leakage from containment and the associated consequences.
The increased risk due to increasing the intervals between containment leakage tests was also evaluated. The NUREG used a statistical approach to determine that the increase in the expected dose to the public due to decreasing the testing frequency is extremely low.
NUREG-1493 also concluded that a small increase is justifiable in comparison to the benefits from decreasing the testing frequency. The primary benefit is in the reduction in occupational radiation exposure.
Criterion 2 Does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The TS amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed TS amendment incorporates the performance-based testing approach authorized by 10 CFR 50 Appendix, J, Option B.
Decreasing the testing frequency allowed by this change does not involve a change to plant design or operation. Safety related equipment 1
and safety functions are not altered as a result of this change. Decreasing the testing frequency does not affect testing methodology. As a result, the proposed change does not affect any of the parameters or conditions that could contribute to the initiation of any accidents.
Criterion 3 Does not involve a significant reduction in the margin of safety.
This TS amendment does not involve a significant reduction in the margin of safety.
The proposed TS amendment does not change the methodology of the containment leakage rate testing program or program acceptance criteria. The proposed TS change does affect the frequency of containment leakage rate testing. With an increased interval between tests, a small possibility exists that an increase in leakage could go undetected for a longer period of time.
Based on the operational experience at CR-3, it has been demonstrated that the leak-tightness of the containment building has consistently been significantly
U. S. Nuclear Regulatory commission 3F0597-21 Page 9 below the allowable leakage limit.
Adequate controls are in place to ensure that required maintenance and modifications are performed.
ENVIRON 11 ENTAL INPACT EVALUATION:
10 CFR 51.22 (c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment.
A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not; (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) result in an increase in individual or cumulative occupational radiation exposure.
Florida Power Corporation (FPC) has reviewed this TS amendment and believes it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22 (c)(9). Pursuant to 10 CFR 51.22 (c), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the proposed TS amendment. The basis for this determination is as follows:
1.
The proposed TS amendment does not involve a significant hazards consideration as described previously in the significant hazards consideration evaluation.
2.
The proposed TS amendment does not result in a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite. The proposed TS amendment does not introduce any new equipment nor does it require any existing equipment or systems to perform a different type of function than they are presently designed to perform.
The proposed TS amendment does not alter the testing methods or testing acceptance criteria associated with primary containment testing.
The primary containment ILRTs and LLRTs being performed will continue to ensure that the consequences of any previously evaluated accident does not increase. FPC has concluded that there will not be a significant increase in the types or amounts of any effluents that may be released offsite and does not involve irreversible environmental consequences beyond those already associated with normal operation.
3.
The proposed TS amendment does not change the duration, methods or acceptance criteria for the performance of ILRTs and LLRTs.
Since the number of ILRTs and LLRTs will be decreased, the amendment does not increase the individual or cumulative occupational radiation exposure.
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L U. S. Nuclear Regulatory commission 3F0597-21 Page 10
SUMMARY
OF CHANGES:
PAGE NUMBER SECTION CHANGE 3.6-2 SR 3.6.1.1 Change the reference to 10 CFR 50, Appendix J for a reference to the Containment Leakage Rate Testing Program.
Deleted the note "SR 3.0.2 is not applicable."
3.6-3 Actions Add the letter "s" to the word Notes.
3.6-7 SR 3.6.2.1 Change the reference to 10 CFR 50, Appendix J for a
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reference to the Containment Leakage Rate Testing Program.
Deleted the note "SR 3.0.2 is not applicable."
3.6-8 Condition A Added "48 inch" clarification to purge valves.
q 3.6-10 Action B1 Corrected typographical error, Added "48 inch" clarification to purge valves.
3 3.6-11 Condition D Added "48 inch" clarification to purge valves.
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3.6-12 SR 3.6.3.2 Renamed 6 inch purge valve to 6 inch post accident l
hydrogen purge valve.
3.6-13 SR 3.6.3.6 Change the reference to 10 CFR 50, Appendix J for a 3
reference to the Containment Leakage Rate Testing Program.
5.0-23A 5.6.2.20 Addition of the Containment Leakage Rate Testing l
Program requirements.
The program definition l
includes the value of the peak calculated internal pressure for the design basis loss-of-coolant j
accident and leakage rate acceptance criteria.
B 3.0-16 SR 3.0.2 Deleted sentences referring to 10 CFR 50, Appendix J as an example where SR 3.0.2 does not apply because of the requirements of the regulations taking precedence over TS.
Added clarification about the Containment Leakage i-Rate Testing Program as an exception where SR 3.0.2 does not apply because the program already includes provisions for the extension of test intervals.
B 3.6-1
Background
Addition of 10 CFR 50, Appendix J,
Option B reference.
B 3.6-2
- Applicable SA Added Option B.
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U.'S. Nuclear Regulatory commission i
3F0597-21 Page 11 I
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SUMMARY
OF CHANGES:(continued)
PAGE NUMBER SECTION CHANGE B 3.6-3 LCO Change the references to 10 CFR 50, Appendix J for a reference to the Containment Leakage Rate Testing Program.
B 3.6-4 SR 3.6.1.1 Change the references to 10 CFR 50, Appendix J for a reference to the Containment Leakage Rate Testing Program.
B 3.6-5~
References Correct Reference 1 to add " Option B."
Correct FSAR reference.
Add Reference 6, NEI 94-01, Revision 0, "Inom ry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J" Add Reference 7,
ANSI /ANS-56.8-1994, "American National Standard for Containment System Leakage 1
Testing Requirement" B 3.6-6
Background
Addition of 10 CFR 50, Appendix J,
Option B reference.
j B 3.6-8 LCO Added (Ref.5) in the text.
B 3.6-13 Actions C1,C2,C3 Change the reference to 10 CFR 50, Appendix J for a reference to the Containment Leakage Rate Testing i
Program. Added clarification of airlock tests.
SR 3.6.2.1 Change the references to 10 CFR 50, Appendix J for a reference to 10 CFR 50, Appendix J, Option B.
1 B 3.6-14 SR 3.6.2.1 Change the references to 10 CFR 50, Appendix J for a reference to the Containment Leakage Rate testing program.
B 3.6-14A' References Correct Reference I to add " Option B."
Correct FSAR references and include ANSI /ANS 56.8-1994.
B 3.6-16
Background
Change name of minipurge valves to 6 inch post i
accident hydrogen purge valves.
B 3.6-18
- Applicable SA Add "48 inch" as a clarification to purge valves statements.
B 3.6-21 Actions B1 and B2 Added 48 inch clarification to purge valves.
B 3.6-24 Note D1
. Added 48 inch clarification to purge valves.
l U.'S. Nuclear Regulatory commission 3F0597-21 Page 12 SUNNARY OF CHANGES:(continued)
PAGE NUNBER SECTION CHANGE B 3.6-25 SR 3.6.3.2 Change name of minipurge valves to 6 inch post accident hydrogen purge valves.
B 3.6-27 SR 3.6.3.6 Added reference to the Containment Leakage Rate Testing Program.
B 3.6-28
. References Correct FSAR references.
- Applicable Safety Analyses (Applicable SA)
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