ML20215N291
| ML20215N291 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/30/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215N226 | List: |
| References | |
| TAC-10710, TAC-64785, NUDOCS 8611050189 | |
| Download: ML20215N291 (8) | |
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'n NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555
't, sN SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING EXEMPTION FROM 10 CFR 50, APPENDIX J, AND AMENDMENT NO. 131 TO FACILITY OPERATING LICENSE NO. DPR-57 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-321
1.0 INTRODUCTION
By letter dated August 7, 1985, the NRC requested Georgia Power Company (GPC) to review its containment leakage testing program for Edwin I. Hatch Nuclear Plant, Unit 1 (Hatch 1), and the associated Technical Specifications for compliance with the requirements of Appendix J to 10 CFR Part 50.
Appendix J to 10 CFR Part 50 was published on February 14, 1973.
Since by this date there were already many operating nuclear plants and a number more in advanced stages of design or construction, the NRC decided to have these plants reevaluated against the requirements of this new regulation.
Therefore, beginning in August 1975, requests for review of the extent of compliance with the requirement of Appendix J were made of each licensee.
Following the initial responses to these requests, NRC staff positions were developed which would assure that the objectives of the testing requirements of the above cited regulation were satisfied.
Subsequently,Section III.D.2 of Appendix J was revised effective October 22, 1980, and conformance is considered in our evaluation. These staff positions have since been applied in our review of the submittals filed by the licensee for Hatch 1.
The results of our evaluation are provided below.
2.0 EVALUATION Our consultant, the Franklin Research Center (FRC), has reviewed the licensee's submittals dated August 28, 1975, November 16, 1977, and March 5, 1979, and prepared the attached evaluation of containment leakage tests for Hatch 1.
We have reviewed this evaluation and concur in its bases and findings, as modified below.
Several changes to the consultant's report should be noted.
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1 A.
The FRC identified six exemptions from the requ+rements of Appendix
'J; however, additional staff review has shown that four of the items are not bonafide exemptions, and,another ites, concerning air lock i
testing, is no 1cnger an exemption because Appendix J has been revised and the proposed testing is now in compliance with Appendix J.
i The following paragraphs discuss the six items which were identified
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as exemptions:
1 1.
Isolation Valves Tested with Water The licensee proposes to test certain isolation valves using water at a pressure of 1.10 Pa, in lieu of air, for systems which remain water-filled post LOCA.
The measured leak rates are not included in the local leak rate test program result.
Appendix J'to 10 CFR 50 requires that unless valves are sealed with 1
fluid from a seal system, they shall be pressurized with air or nitrogen for leak testing purposes (Paragraph III.C.2).
There are a number of valves, however, that are designed to remain covered with water after a LOCA and thus provide a water seal for the isolation valves or ensure that only liquid leakage from the containment will j
occur.
For such valves, the licensee purposes to perform hydrostatic t.esting to determine their leak tightness.
These valves fall into
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two categories, as discussed below.
A.
Sealed by Water from the Torus The following penetrations and systems are connected to the torus:
i 203 RCIC Pump Suction i
204 A, B, C, D RHR Pump Suction 207-HPCI Pump Suction 208 A, B Core Spray Pump Suction 3
210 A, B RHR/ Core Spray Test Line i
The piping for these systems penetrates the torus and terminates i
below the water line of the torus.
As a supply of water in the torus is assured during post-accident conditions, these valves will remain i
sealed with water. Therefore, in accordance with Sections III.C.2 j
' and III.C.3 of Appendix J, the valves need not be tested with air.
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. Although the licensee proposes to test them with water, this is not necessary, as the purpose of the water leak test is to assure a supply of sealing water for 30 days following onset of an accident.
- As the torus is postulated to always remain filled with water, no leak test is necessary to satisfy Appendix J requirements.
For the above reasons, the staff finds the proposed testing of the isolation valves in the above penetrations to be in compliance with the requirements of Appendix J.
B.
Closed Systems Inside Containment The following penetrations and systems are discussed in this section:
20 Service Water Supply 44 Service Water Return 23 Reactor Building Closed Cooling Water Supply 24 Reactor Building Closed Cooling Water Return The Service Water and Reactor Building Closed Cooling Water (RBCCW) systems are closed systems inside containment. These closed systems constitute one of the two containment isolation barriers for each of the penetrations listed above, and are subject to ASME Section XI in-service inspection requirements.
They are designed to remain intact and water filled post-LOCA.
In accordance with Sections III.C.2 and III.C.3 of Appendix J, the licensee proposes to leak test the isolation valves in these systems with water at a pressure of 1.10 Pa; the leakage acceptance criteria are based upon maintaining a 30-day inventory of water for sealing the valves.
Therefore, the staff finds the proposed testing of the isolation valves in the above penetrations to be in compliance with the requirements of Appendix J.
2.
Main Steam Isolation Valves Appendix J to 10 CFR 50 requires leak rate testing of BWR main steam isolation valves (MSIVs) (Paragraph II.H.4) at Pa, the peak calculated containment pressure related to the design-basis accident (Paragraph III.C.2).
Further, Appendix J requires that the measured leak rates be included in the summation for the local leak rate tests (Paragraph III.C.3).
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. The licensee proposes to leak test the MSIVs at a reduced pressure and exclude the measured leakage from the combined local leak rate test results.
The staff has determined that an exemption to Appendix J is required for this proposal.
The basis for this determination is discussed below.
Each main steam line is provided with two MSIVs that are oriented to seal in the direction of post-accident containment atmosphere out-leakage.
The design of the MSIVs is such that testing in the-reverse direction tends to unseat the valve.
Simultaneous testing of the two valves, at design pressure, by pressurizing between the valves, would lift the disc of the inboard valve and result in a meaningless test.
The proposed test calls for a test pressure of 28 psig (one-half of Pa) to avoid lifting the disc of the inboard valve.
The total observed leakage through both valves (inboard and outboard) is then conservatively assigned to the penetration.
The staff concludes that this procedure is acceptable.
Furthermore, excluding the leakage from the summation for the local leak rate tests is acceptable because a separate leakage rate acceptance criterion of 11.5 standard cubic feet per hour is used for the MSIVs.
This separate limit was found acceptable during the operating license review for Hatch 1, as discussed in Section 5.4.4 of the SER, dated May 11, 1973, and Supplement No. l'to the SER, dated December 10, 1973.
The radiological consequence of this separate leakage was considered generically as described by Regulatory Guide 1.96, " Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants," Rev. 1, dated June 1976, which concluded that the Hatch 1 plant did not need to add such a leakage control system in order to reduce the radiological consequence. The separate limit of 11.5 scfh was also included in the original facility Technical Specifications.
The staff concludes that leak testing the MSIVs in the way described above is an acceptable alternative to the requirements of Appendix J, and that an exemption to Appendix J is justified and acceptable.
3.
Air L'.ck 1
The licensee's proposal to test personnel air lock door seals by j
pressurizing the volume between the double door seals to a pressure of 10 psig required an exemption when originally proposed and reviewed by FRC.
Because of a subsequent revision to Appendix J (October 22, 1980), this testing no longer requires an exemption, but rather complies with the current requirements of Appendix J.
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Closed Systems Outside Containment The following penetrations and systems are under discussion in this section:
12 RHR Suction 13 A, B RHR Return to recirc.
16 A, B Core Spray 17 RPV Head Spray 39 A, B Containment Spray 1
211 A, B Torus Spray l
For each of these penetrations, the inner isolation barrier is an isolation valve that is Type C tested in accordance with Appendix J; the outer barrier is a closed system outside containment, having no containment isolation valve. Thus, the closed systems outside i
containment cannot be Type C tested. However, the systems are subject to the inservice inspection requirements of the AMSE Boiler I
and Pressure Vessel Code,Section XI, for Nuclear Class 2 piping, which requires that the entire closed system be pressurized and any i,
visible leakage be repaired. They are visually inspected up to the containment isolation valves when the various system pump functional j
tests are performed. The leakage is not added to the local leak rate test program result for pneumatic testing. Because the isolation valves in these systems are Type C tested, the staff finds the testing program to be in compliance with the requirements of Appendix J.
5.
Traversing Incore Probe System The traversing incore probe system is equipped with a ball valve.in each guide tube that provides shutoff capability following cable withdrawal. A shear valve is also provided for each guide tube to cut the cable and isolate the tube if the drive cable cannot be I
withdrawn.
The licensee vill perform a Type C test on the ball valve.
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the shear valve recuires testing to destruction, the licensee cannot perform periodic Tyre C tests on these valves. However, manufacturer. Failure of a single shear valve to meet the 10-g the statistically cFo:en uraples of the shear valves are tested b cc/sec j
leakage criterion results in rejection of the entire lot. -The licensee has comitted, by letter dated May 14, 1986, to test the explosive charges which operate the shear valves using procedures similar to those currently used for the standby liquid control system i
to comply with Hatch Unit 1 Technical Specification Sections 3.4/4.4.
j These Sections require that a portion of explosive charges installed in the valves be fired each refueling cycle to assure that the
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i installed charges are operable and require that all installed charges are tested during the course of two fuel cycles. They.also require that replacement charges be selected from batches that have w -.
, been tested. Continuity of the electrical system that fires the charges to operate the shear valves is continuously monitored via indicating lights. Should continuity be lost, indicating light illuminate, and an alarm is received in the control room.
Based on the above discussion, the staff concludes that the leak testing of the traversing incore probe system is acceptable, and no exemption is required, as the testable valves (ball valves) are Type C tested and the shear valves, which cannot be Type C tested, undergo alternative surveillance.
6.
Control Rod Drive The design of the Control Rod Drive (CRD) insert and withdraw lines does not facilitate Type C testing, as there are no containment isolation valves in these lines.
However, adequate leakage moni-toring of the CRD lines is provided by' normal plant operating proce-dures and the Type A leakage rate tests.
Since the insert and withdraw lines are pressurized to'at least reactor operating pressure (1000 psi) by the cooling water flow during normal plant operation, leakage from these lines would be immediately evident.
I The hydraulic control units are installed in a relatively high traffic area of the reactor building.
In addition, plant procedure requires that an operator make a visual inspection of the CRD hydraulic control units (operating pressure 1000 psi) for leakage at' least once per shift and that he record the inspection. Furthermore, because the reactor pressure vessel and nonseismic portion of the control rod drive system are vented during Type A tests, leakage monitoring of the control rod drive insert and withdraw lines is provided by Type A leakage rate tests.
The CRD system does not contain isolation valves that fall into the categories defined in Section II.H " Type C tests," of Appendix J.
Furthermore, because of the foregoing considerations, the CRD system does not constitute a potential containment atmosphere leak path.
Therefore, the CRD system does not require Type C testing. The staff concludes that leakage monitoring of the control rod drive system in the manner described above meets the requirements of Appendix J.
B.
Table 5 (page 8) of the report discusses a proposed change to Technical Specification 4.7.E to update the reference to Appendix J so as to include the latest revision of Appendix J.
Subsequent to the writing of the consultant's report (April 1980), the licensee, by letter dated February 7,1984, has determined that reference to a specific revision to Appendix J should be deleted from Technical Specification 4.7.
This eliminates the need for future revisions to the Technical Specifications whenever Appendix J is revised. The Technical Specification is thereby, always consistent'with the current Appendix J to 10 CFR 50. We conclude that.this change is-acceptable.
4 C.
The FRC accepted the licensee's proposal to delete Tables 3.7-2, 3.7-3, and 3.7-4 from the current Technical Specifications. These tables contain list of primary containment penetration's with double 0-ring seals, j
containment penetration's isolation valves, respectively. The licensee stated that, with respect to the updated program, these tables are inaccurate and incomplete.
Rather than to include all of the i
penetrations and valves in these tables, the licensee determined that it 4
would be more prudent to incorporate statements in the surveillance requirements outlining the programs for the Type A, B, and C tests to be in accordance with Appendix J.
The tables would then be maintained as part of the plant's Appendix J program procedure.
However, based on additional review of this request the staff has i
determined that these tables should not be deleted from the Technical Specifications at this time, as they provide guidance to the NRC's regional inspectors in measuring the compliance of the licensee with the requirements of Appendix J.
Deletion' would also be contrary to current standard Technical Specifications. Therefore, the staff concludes that deletion of the tables is not acceptable at this time. However, as part i
of the staff's ongoing effort to generically improve Standard Technical l
Specifications, such a deletion may be reconsidered.in the future.
In the meantime, the licensee.has informed the staff that updated tables will be submitted for inclusion in the Hatch 1 Technical Specifications.
Based on our review of the licensees request and on our review of the l
attached Technical Evaluation Report as prepared by the FRC, we have made the following conclusions regarding the Appendix J review for Hatch 1:
1.
The updated containment leak rate test program submitted by GPC in the March 5, 1979, letter is acceptable.
In addition, the associated proposed exemption from the requirements of Appendix J, concerning l
MSIV testing, is acceptable.
2.
The proposed piping modifications submitted by GPC with its March 5, 1979, letter are acceptable with regard to the proper testing of l
valves per the requirements of Appendix J.
l 3.
The proposed changes to the Technical Specifications requested by GPC l
in its March 5,1979 letter, as supplemented by a February 7,1984 j
letter, are acceptable, except as described in section C above.
3.0 ENVIRONMENTAL CONSIDERATION
An Environmental Assessment and Final Finding of No Significant Impact has been issued for the Exemption.
The amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 l
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. CFR Part 20.
We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occ~upational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, the amendment meets the eligibility-criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this action will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
J. Pulsipher Dated: October 30, 1986
Attachment:
Technical Evaluation Report
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