ML18065A798
| ML18065A798 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/28/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18065A797 | List: |
| References | |
| NUDOCS 9607010308 | |
| Download: ML18065A798 (11) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE THIRD JO-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN REQl!lfSTS FOR RELIEF NOS. PR-02 AND PR-04
- 1. 0 INTRODUCTION fQR CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NUMBER: 50-255 The Technical Specifications for Palisades state that the 1nservice inspection (ISi) of the American Society of Mechanical Engineers (ASHE) Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of qua1ity and safety or (ii) compliance with the specified requirements would resu1t in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASHE Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components,* to the exbmt practical within the.1 imitations of design, geometry, and materials of construction of the components.
The regulations require that inservice examination of components and system pressure tests conducted during the first IO-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASHE Code for the Palisades Plant third 10-year ISi interval is the 1989 Edition. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Convnission approval.
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,,I In a letter dated March 11, 1996, Consumers Power Company submitted to the NRC its third IO-Year Interval ISi Program Pl an Requests for Relief Nos. PR--02 and PR-04 for Palisades Plant, seeking prior NRC approval for proposed alternatives.
2.0 EVALUATION AND CONCLUSIONS The staff, with technical assistance from its contractor, the Idaho National Engineering Laboratory (INEL), has evaluated the information provided by the licensee in support of its third IO-Year Interval ISi Program Plan Requests for Relief Nos. PR-02 and PR-04 for Palisades Plant.
Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the Technical Letter Report attached. The staff has concluded that for Request for Relief PR-02 to require the licensee to perform the code-required VT-2 visual examination of the reactor pressure vessel and associated components, accessible only within the reactor cavity, would result in a hardship without a compensating increase in the level of quality and safety.
To perform the code-required examination an examiner enters the restricted area beneath the reactor pressure vessel under high temperature, radiation, and other potentially hazardous conditions.
The licensee's proposed alternative to monitor for leakage based on Technical Specification leakage limits and the staff's provision to perform a VT-2 visual examination of the reactor cavity during a cold shutdown following a run cycle to verify that there is no evidence of leakage will provide reasonable assurance of operational readiness of the reactor pressure vessel within the reactor cavity. Therefore, the alternative contained in Request for Relief No. PR-02 is authorized pursuant to 10 CFR 50.55a(3)(ii), provided that the licensee performs a VT-2 visual examination for evidence of leakage in the reactor pressure vessel cavity during cold shutdown following a run cycle.
in addition, the ~taff has determined that the licensee's proposed alternative to use Code Case N-522, "Pressure Testing of Containment Penetration Piping, 11 contained in Request for Relief No. PR-04 provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative to use Code Case-N-522 in lieu of the code-required pressure tests is authorized pursuant to 10 CFR 50.55a(a)(3)(i), provided that the licensee performs the leak test at the peak calculated containment pressure and uses a test procedure that provides for detection and location of through-wall leakages in the pipe segments being tested. The use of alternatives contained in Code Case N-522 are authorized for the current interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147._ At that time, if the licensee intends to continue to implement this COO'e Case, the licensee is to follow all provisions in code case N-522 with limitations issued in Regulatory Guide 1.147, "lnservice Inspection of Code Case Acceptability, ASME Section XI, Division l," if any.
Attachment:
INEL Technical Letter Report Date:
June 28, 1996
TECHNICAL LETTER REPORT ON THE THIRD 10-YEAR INSERVICE INSPECTION INTERVAL REQUESTS FOR RELIEF PR-02 AND PR-04
. FOR PALISADES PLANT CONSUMERS POWER COMPANY DOCKET NUMBER:
50-255
1.0 INTRODUCTION
By letter dated March 11, 1996, Consumers Power Company submitted Requests for Relief PR-02 and PR-04.
The Idaho National Engineering Laboratory (INEL) staff has evaluated the subject requests for relief in the following section.
2.0 EVALUATION The Code ~f record for the Palisades Plant, third IO-year inservice inspection (ISI) interval, which began in August 1995, is the 1989 Edition of the American Society of Hechanical Engineers (ASHE) Boiler and Pressure Vessel Code,Section XI.
The information provided by the licensee in support of the requests for relief from Code requirements has been evaluated and the bases for disposition are documented below.
A.
Request for Relief PR-02. Table IWB-2500-1. Examination Category 8-P. VT-2 Visual Examination of the Reactor Pressure Vessel Code Requirement:
Section XI, Table IWB-2500-1, Examination Category 8-P requires a VT-2 visual examination in conjunction with system leakage and hydrostatic tests specified in IWB-5221 and IWB-5222.
Licensee's Code Relief Request: The licensee requested relief from performing the Code-required VT-2 visual examination of the reactor pressure vessel.
Licensee's Basis for Requesting Relief (as stated):
2 "Pursuant to 10CFR50.55a(a}(3}(ii}, an alternative to the visual VT--2 examination specified by the requirements is requested on the.
basis that they would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety.
Also, per 10CFR50.55a(a}(3}(ii}, the proposed alternate examination will provide an acceptable level of quality and safety.
"The area under the reactor vessel is extremely hazardous when the Plant is at Hot Shutdown conditions for system leakage testing.
Radiation levels are expected to be 2.5 rem/hr (on contact}, which is the maximum measured during cold shutdown.
General dose is 1.5 to 2 rem/hr in the area.
Assuming 2 persons in this area at one-hal f hour per person, a total of 1.5 to 2 rem of dose would be received.
"In addition to radiation concerns, access to the area under the vessel poses various industrial hazards.
Of primary concern is confined spaces and heat stress. Ambient air temperatures with the Primary Coolant System at full pressure and temperature are expected to be approximately 300 degrees. Access under these conditions would require significant ventilation for cooling.
The access tube to this area is only 30 inches in diameter. This size limits the amount of ventilation possible while allowing personnel access."
Licensee's Proposed Alternative Examination (as stated}:
"Palisades Nuclear Plant shall determine leakage from piping and components in the area under the reactor vessel in accordance with paragraph IWA-5244 "Buried Components" of ASME Section XI, 1989 Edition, no Addenda.
This requirement will be satisfied by conducting Palisades' System Operating Procedure SOP 1, "Primary Coolant System (PCS}," which completes the PCS leak rate calculation. Plant Technical Specification 3.1.5 states, "If the primary coolant system leakage exceeds 1 gpm and the source of the leakage is not identified, reduce unidentified primary coolant system leakage to less than 1 gpm within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or place the reactor in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." Technical Specification 3.l.5d invokes a more restrictive leakage limit of 0.6 gpm during startups. Technical Specification Table 4.2.2, Item 7 requires this leak rate determination on a daily basis.
These limits are approved as documented in Palisades Facility Operating License DPR-20, through Amendment No 161 and are applicable at all times when the Primary Coolant System is greater than cold shutdown conditions."
Evaluation:
The licensee requested relief from performing a VT-2 visual examination of the reactor pressure vessel and associated
3 components, accessible only from within the reactor cavity, due to high radiation and the hazardous environment at operating pressure.
To perform the subject VT-2 visual examination, examiners would be required to enter the reactor cavity area under the reactor pressure vessel via the 30 inch diameter access tube.
Considering that this area could be categorized as a confined space, additional personnel would be required to support this activity for safety purposes and for monitoring radiation exposure.
The radiation levels in the reactor pressure vessel cavity are estimated to be 1.5 to 2 REM per hour.
As such, examiners required to enter this area to perform the examination would likely receive a dose of 1.5 to 2 REM for the subject examination.
In addition, to perform the subject examination at operating pressure, the reactor coolant system would be at an elevated temperature (approximately 300°F).
This would result in the examiners having to utilize cooling systems to perform the required examination.
As a result of the harsh environment and the high radiation area in which examiners would be required to work, the INEL staff believes that attention to details for performing a quality examination may be compromised.
As an alternative to the VT-2 visual examination, the licensee proposes to monitor the primary coolant system for leakage.
The licensee will comply ~ith Technical Specifi~ations for leakage limits to.maintain the system within safe operating parameters.
Plant Technical Specification 3.1.5 states, "If the primary coolant system leakage exceeds 1 gpm and the source of the leakage is not identified, reduce unidentified primary coolant system leakage to less than I gpm within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or place the reactor in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." Technical.Specification 3.l.5d invokes a more restrictive leakage limit of 0.6 gpm during startups.
Technical Specification Table 4.2.2, Item 7, requires this leak rate determination on a daily basi~. Considering the conservatism established in the licensees' Technical Specifications for
4 monitoring leakage in the reactor coolant system, it is ~easonable to conclude that leakage, if occurring, will be detected.
The INEL staff believes that imposing the Code requirement on the licensee during pressure tests will result in a burden without a compensating increase in quality and safety. However, INEL staff believes that personnel hazards are minimized in cold shutdown during refueling operations.
As such, a VT-2 visual examination for evidence of leakage, could be performed, providing reasonable assurance of operational readiness, with minimal burden on the licensee.
As a result, it can be concluded that, the licensee's proposed alternative, to monitor for leakage in accordance with Technical Specifications, in conjunction with a VT-2 visual examination following a run cycle, will verify that leakage is not occurring, providing reasonable assurance of operational readiness.
Therefore, it is recommended that the licens~e's proposed alternative be authorized pursuant to 10 CFR 50.55a(3)(ii), provided that the licensee performs a VT-2 visual examination for evidence of leakage in the reactor pressure vessel cavity during cold shutdown following a run cycle.
B.
Request for Relief PR-04, Examination Category C-H. Item C7.30, C7.40. C7.70. and C7.80, Pressure Tests of Class 2 Piping and Valves at Containment Penetrations Where the Balance of the System is Ncnclass Code Requirement: Section XI, Table IWC-2500-1, Examination Category C-H, Items C7.30, C7.40, C7.70, and C7.80 require system pressure tests each period and system hydrostatic tests once each interval.
licensee's Code Relief Request: Relief is requested from the Code-required pressure tests for the following system penetration piping of the Containment Penetration Piping Containment Isolation System (CIS).
-1
l 5
PENETRATION DESCRIPTION NUMBER MZ-la Purge Air Exhaust HZ-lb Purge Air Exhaust Bypass MZ-lc Purge Air Exhaust Mz-10*
Service Air MZ-lOa Service Air MZ-11 Condensate to Shield Cooling Surge Tank MZ-17 Containment Pressure Instrument Line MZ-18 Transfer Winch Cable MZ-18a Fuel Transfer Tube MZ-19 Personnel Air Lock MZ-21 Hydrogen Monitoring Return Line MZ-21a Hydrogen Monitoring Supply Line MZ-25 Clean Waste Receiver Tank Vent MZ-26 Nitrogen to Quench Tank MZ-27 ILRT Fill Line MZ-28 Containment Air Sample Line MZ-33 Safety Injection Tank Drain MZ-37 Primary System Drain Tank Recirculation MZ-38 Containment Building Heating Steam Return MZ-39 Containment Building Heating Steam Supply
- MZ-40a Hydrogen Monitoring Return Line MZ-40b Hydrogen Monitoring Supply Line MZ-41 Degasifier Pump Discharge MZ-42 Demin Water to Quench Tank MZ-44 Primary Coolant Pump Controlled Bleed Off MZ-46 Containment Vent Header MZ-47 Primary System Drain Tank Pump Suction MZ-48 Containment Pressure Instrument Lines
6 PENETRATION DESCRIPTION NUMBER MZ-49 Clean Waste. Receiver Tank Circulation Une MZ-50 Emergency Access MZ-51 Equipment Hatch MZ-52a Containment Sump Level Instrument MZ-52b Containment Sump Level Instrument MZ-64 Reactor Cavity Fill and. Recycle MZ-65 Instrument Afr MZ-66 ILRT Instrument Line MZ-67 Clean Waste Receiver Tank Pump Discharge MZ-68 Air Supply to Air Room MZ-69 Clean Waste Receiver Tank Pump Suction MZ-72 Reactor Cavity Drain and Recycle Licensee's Basis for Requesting Relief (as stated}:
"Pursuant to 10 CFR 50.55a(a}(3}(i}, the proposed alternative provides an acceptable level of quality and safety.
The proposed alternative will reduce the level of redundant testing.
Imposition of the IWC-2500-1 test requirements would result in additional testing as follows:
- 1.
Water systems would be flooded and pressurized through the associated penetration test tap and a VT-2 visual examination performed during the pressurization.
- 2.
Air and gas systems would be tested in an identical fashion to 10CFR50, Appendix J test method and a VT-2 visual examination would be performed during the pressurization period.
°For the penetrations listed, IOCFR50, Appendix J testing is performed by draining the test volume, if required, venting downstream of the test volume and pressurizing the test volume to 55 psig. The rate of pressure decay is determined and compared to acceptance criteria based on allowable containment leak rate.
"Leakage from water systems would be indicated during the 10CFR50, Appendix J test more readily than the IWC-2500-1 test due to the lower density of the air test medium.
Based on this fact, the use
7 of the Appendix J test program is conservative when compared.to ASME Section XI program.
"Leakage from air and other gas systems would be indicated by 10CFR50, Appendix J testing in a similar manner to an IWC-2500-1 test. Based on this fact, the two programs are essentially equivalent.
"The specified frequency of testing for the 10CFR50, Appendix J test program is once per refueling cycle, which is approximately 18 months.
The IWC-2500-1 specified test frequency is once per inspection period which is approximately 3-1/3 years. A comparison of the specified test frequencies indicates the 10CFR50, Appendix J test program is conservative when compared to the ASME Section XI test program.
"Palisades 10CFR50, Appendix J testing is performed by plant auxiliary operators (AO's).
Palisades certifies the AO's as VT-2 Level II examiners in accordance with Section XI.
Therefore, Appendix J testing is performed by personnel trained to recognize unacceptable leakage from a pressure boundary."
Licensee's Proposed Alternative Examination (as stated):
"In lieu of the requirements specified in IWA-5211 and IWC-5210, Palisades will follow the guidance of Code Case N-522 for pressure testing of safety class 2 containment penetrations associated with non-safety class systems.
"Testing required by 10CFRS0.55, Appendix J may be used as.
an alternative to the rules in Table IWC-2500-1, Category C-H, for pressure testing piping that penetrates a containment vessel, when:
I.
The piping and isolation valves that are part of the containment system are Class 21 but the balance of the piping system is outside the scope of Section XI."
Evaluation:
The licensee proposes to implement the alternatives contained in Code Case N-522, Pressure Testing of Containment Penetration Piping, in lieu of the Code-required pressure tests for portions of the subject lines that are Class 2 at the containment penetration. These segments of lines are safety-related only because they function as part of the containment pressure boundary and are relied on for containment integrity. Therefore, it is logical to test the penetration piping portion of the associated
8 systems to the containment test criteria found in 10 CFR50.55a, Appendix J.
Appendix J pressure tests verify the leak-tight integrity of the primary reactor containment and of systems and components that penetrate containment by local leak rate and integrated leak rate tests. In addition, Appendix J test frequencies provide assurance that the containment pressure boundary is being maintained at an acceptable level while monitoring for deterioration of seals, valves, and piping.
The Class 2 containment isolation valves (CIVs) and connecting pipe segments must withstand the peak calculated containment internal pressure related to the maximum design containment pressure.
The containment penetration piping is classified as Class 2 because of its function as part of the containment pressure boundary, and because containment integrity is the only safety-related function performed by this penetration piping. Therefore, it is logical to test the penetration piping of the associated system to the Appendix J criteria. The INEL finds that the pressure-retaining integrity of the CIVs and connecting piping and their associated safety functions may be verified with an Appendix J, Type C test if conducted at the peak calculated containment pressure.
The seal between the connecting pipe segment and containment may be verified using an Appendix J, Type B test. Therefore, when the connecting pipe segment is subjected to either a Type B or C test, its safety function is verified by the Appendix J test.
Section XI, IWC-5210(b) requires that where air or gas is used as a testing medium, the test procedures shall include methods for detection and location of through-wall leakage in components of the system tested. If the licensee's Appendix J, Type C test uses air as a testing medium, the test procedure should meet the above requirement for the CIVs and pipe segments between the CIVs.
9 The INEL staff believes that an acceptable level of quality and safety will be provided by Appendix J tests, provided that the licensee performs the leak test at the peak calculated containment design pressure and that a test procedure is implemented that provides for detection and location of through-wall leakages in the pipe segments that are being tested. Therefore, it is recommended that the licensee's proposed alternative to the Code-required pressure tests be authorized pursuant to 10 CFR 50.55a{a){3){i),
provided that the licensee performs the leak test at the peak calculated containment pressure and uses a test procedure that provides for detection and location of through-wall leakages in the pipe segments being tested. The use of alternatives contained in Code Case N-522 should be authorized for the current interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this Code Case, the licensee is to follow all
. provisions in Code Case N-522 with limitations issued in Regulatory Guide 1.147, if any.
3.0 CONCLUSION
The INEL staff has evaluated Requests for Relief PR-02 and PR-04.
For Request for Relief PR-02, it is recommended that the licensee's proposed alternative to the Code requirements be authorized pursuant to 10 CFR 50.55a{a){3){ii), provided that the licensee performs a VT-2 visual examination for evidence of leakage in the reactor vessel cavity during cold shutdown following a run cycle.
For Request for Relief PR-04, it is recommended that the licensee's proposed alter~ative pressure test be authorized pursuant to 10 CFR 50.55a{a){3)(i), provided that the licensee performs the leak test at the peak calculated containment pressure and uses a test procedure that provides for detection and location of through-wall leakages in the pipe segments being tested.