ML20212N001
| ML20212N001 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/04/1987 |
| From: | James O'Reilly GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20212N003 | List: |
| References | |
| SL-2074, TAC-64785, NUDOCS 8703120306 | |
| Download: ML20212N001 (13) | |
Text
c c-Georgia Pbwet Cornpany 333 Piedmont Avenue -
',s I-Atlanta. Georgia 30308
+~
i-Tele @one 404 526 7851 Mailing Address:
.s Post Off,ce Box 4545
[
[
' Atlanta. Georgia 30302 r V ce Pre dent e
Nuclear Operations b~
74 March 4,1987 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES Gentlemen:
In accordance with the provisions of 10 CFR 50.90, as required by 10 CFR 50.59(c)(1), Georgia Power Company (GPC) hereby proposes changes to the Pl ant Hatch Unit 1 Technical Specifications, Appendix A to Operating License DPR-57.
The following changes are proposed:
Proposed Changes 1 through 3 would delete Appendix J 1eak rate testing requirements for certain valves from TS Table 3.7-4.
Proposed Change 4 would provide an administrative correction - to the Main Steam Isolation Yalve (MSIV) test pressure requirement.
Proposed Change 5 would provide editorial corrections to certain information depicted in TS Table 3.7-4.
We respectfully request that the requested changes be granted prior to commencement of the upcoming Plant Hatch Unit 1 refueling outage, scheduled to commence on approximately April 22, 1987.
Justifications for this request are provided in Enclosure 1. provides detailed descriptions of the proposed changes and the bases for each change request. details the basis for our determination that the l
proposed changes do not involve a significant hazards consideration. provides page change instructions for incorporating the proposed changes into the Unit 1 Technical Specifications.
The proposed changed Technical Specifications pages follow.
A?(
h deck 0703120306 670304 V
ADOCK 050 1
O pDR ggg,b
..____~ _ __
4 L
' U. S. Nuclear Regulatory Commission i
ATTN: Document Control Desk March 4,1987 Page Two i
Payment of the filing fee in the amount of one hundred and fifty dollars is enclosed, f
Pursuant to the requirements of 10 CFR 50.91, a copy of this letter and all applicable enclosures will be sent to Mr. J. L. Ledbetter of the i
Environmental Protection Division of the Georgia Department of Natural Resources.
j Mr. James P. O'Reilly states that he is Senior Vice President of Georgia Power Company and is authorized to execute this oath on behalf of Georgia Power Company, and that to the best of his knowledge and belief, the facts set forth in this letter and enclosures are true.
)
GEORGIA POWER C0FFANY i
I i cumtt h. L9 Mb By:
James P. O'Reilly j
Sworn to and subscribed before me th th y of March 1987.
~
(7C hv A Clo,*
Ndtery Pubhc. ClaytoE County, Geeres Notary Public My Commrssion Evres Dec. 12.1909 l
REB /lc Enclosures i
c: Georf f a Power Company U.S. Nuclear Regulatory Commission Mr. s. T. Beckham, Jr.
Dr. J. N. Grace, Regional Administrator Mr. H. C. Nix, Jr.
Mr. P. Holmes-Ray, Senior Resident GO-NORMS Inspector - Hatch Hr. G. Rivenbark, Licensing Project l
1 Manager - Hatch i
State of Georgia l
Hr. J. L. Ledbetter i
I f
1164C L
1 ENCLOSURE 1 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT-ISOLATION VALVES BASES FOR CHANGE REQUEST PROPOSED CHANGE 1:
Delete certain valves from Technical Specifications (TS) Table 3.7-4 pursuant to NRC Appendix J Safety Evaluation.
By letter of October 30, 1986, the NRC issued to GPC a Safety Evaluation (SE) addressing GPC's program to implement the requirements of 10 CFR 50 Appendix J, which provides for leak testing of containment isolation valves.
This SE justified several categories of containment penetrations as not requiring leak rate testing of associated valves pursuant to Appendix J requirements.
Plant Hatch Unit 1 Technical Specifications (TS) Table 3.7-4 provides a listing of containment isolation valves subject to leak rate testing pursuant to the requirements of Appendix J.
Deletions to this Table are now appropriate based on the conclusions of the above NRC SE.
This request covers only a portion of the valves addressed by the SE as not requiring Appendix J 1eak rate testing requirements.
The valves hereby requested for deletion are those having the most significant outage impact in terms of time and resource requirements, and have been expeditiously identified with the intent of obtaining NRC approval of the requested TS changes prior to the upcoming Unit 1 refueling outage, currently scheduled to commence approximately April 22, 1987.
The remaining valves justified as acceptable for deletion will be the subject of a later submittal.
Proposed Change 1 would delete certain valves from TS Tabic 3.7-4 as listed below:
Penetration: 210A RilR/ Core Spray test line Valves:
E11 (RHR) System:
F007A, F0ll A, F024A, F025A, F026A, F029, FOSSA, F097, F103A, F3078A (thermal relief valve)
E21 (Core S aray) System:
- F015A, 7031A, F044A E51 (RCIC) System:
F019, F021 1164C El-1 03/04/87
Ls ENCLOSURE 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAIHMENT ISOLATION VALVES BASES FOR CHANGE REQUEST PROPOSED CHANGE 1 (Continued):
Penetration:
210B RHR/ Core Spray test line Valves:
E11 (RllR) System:
F0078, FullB, F024B, F0258, F026B, F055B, F103B, F3078B (thermal relief valve)
E21 (Core S] ray) System:
- F015B, 7031B, F044B E41 (HPCI) System:
F012, F046 Penetration:
203 RCIC pump suction Valves:
E51 (RCIC) System:
F003, F031 Penetration: 204A, B, C, D RHR pump suction Valves:
Ell (RHR) System:
F004A, B, C, D F030A, B, C, D Penetration:
207 HPCI pump suction Valves:
E41 (HPCI) System:
F042, F051 Penetration:
208A, B Core Spray pump suction Valves:
E21 (Core S] ray) System:
- F001A, 3
Basis for Proposed Change 1:
TS Table 3.7-4 specifies valves that are required to be tested pursuant to Appendix J requirements.
The basis for deletion of Appendix J test requirements for the above valves is contained in Item 2.0. A.1 or the NRC's SE dated October 30, 1986.
Item 2.0.A.1 states the following:
1164C El-2 03/04/87
ENCLOSURE 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES BASES FOR CHANGE REQUEST PROPOSED CHANGE 1 (Continued)
(Quoting from NRC SE dated October 30, 1986)
The following penetrations and systems are connected to the torus:
203 RCIC Pump Suction 204 A, B, C, D RHR Pump Suction 207 IIPCI Pump Suction 200 A, B Core Spray Pump Suction 210 A, B RHR/ Core Spray Test Line The piping for these systems penetrates the torus and terminates 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 sealed with water.
Therefore, in accordance with Sections III.C.2 and III.C.3 of Appendix J, the valves need not be tested with air....As the torus is postulated to always remain filled with water, no leak test is necessary to satisfy Appendix J requirements.
Although these valves may be exempt from Appendix J, Type C testing, leak rate testing per the requirements of ASME Code,Section XI, IWV-3420, will be performed.
PROPOSED CHANGE 2:
Delete from T5 Tahl. 3.7-4 the following components:
Penetration:
218A Valves: 1G51-F002, F0ll, F012 Blind Flange:
lG51-0001 Basis for Proposed Change 2:
These torus drainage system components remain submergea following a LOCA and, therefore, do not constitute a potential bypass leakage pathway.
The basis is the same as Proposed Change 1 above; however, this penetration was not explicitly addressed by the reference NRC SE.
These components were added following the original GPC submittal which provided the basis for the SE.
1164C El -3 03/04/87
ENCLOSURE 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES -
BASES FOR CHANGE REQUEST PROPOSED CHANGE 3: Delete from TS Table 3.7-4 the following valves:
Penetration:
211 A, B RHR Torus Spray Valves: 1 Ell-F027A, B Basis for Froposed Change 3:
The proper inboard isolation barriers for these torus spray lines are the 1 Ell-F028A, B valves, which are currently tested pursuant to Appendix J (in addition to the lEll-F027A, B valves).
However, the outer barrier is properly provided by a Quality Group B, Seismic Category 1, missile-protected, closed system (RHR system), as identified in Table 7.3-1 (note 30 for penetrations 211 A and B) of the Plant Hatch Unit 1 FSAR.
Thus, Appendix J testing of the lEll-F027A, B valves should not be required.
Closed systems located outside containment do not require Appendix J testing.
However, the systems are subject to the inservice inspection requirements for Class 2 piping stated in the ASME Code,Section XI, which requires that the entire closed system be pressurized and any visible leakage repaired.
When the various system pump functional tests are performed, the systems are visually inspected up to the containment isolation valves.
Since the system is a closed system, no potential bypass leakage pathway exists, and fission-product retention would be maintained following a design basis accident.
PROPOSED CHANGE 4: Provide corrections to TS Table 3.7-4 as follows:
Penetrations: 20, 25, 35E, 205 Valves: P41-F050, T48-Fil88, C51-Nitrogen Inerting Basis for Proposed Change 4:
These miscellaneous changes are mado to reflect correct penetration assignments and nomencl ature in TS lable 3.7-4 as follows:
a.
Valve P41-F050 is incorrectly assigned to penetration 20; the correct penetration is number 44, b.
Valve T48-Fil8B is incorrectly assigned to penetration 25; the correct penetration is number 205.
c.
Valve T48-F104 is associated with penetrations 25 and 205.
d.
The proper nomenclature for valve C51 is C51-Nitrogen Inerting Check Valve.
Il64C El-4 03/04/87
L ENCLOSURE 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES BASES FOR CHANGE REQUEST PROPOSED CHANGE 4 (Continued):
These changes ace purely editorial in nature and reflect no change in testing requirements.
PROPOSED CHANGE 5: Provide editorial correction to MSIV test pressure.
Basis for Proposed Change 5:
Plant Hatch Unit I was originally licensed with a requirement that the Main Steam Isolation Yalves (MSIVs) be tested at a test pressure of 28 psig.
This value, which is slightly less than 1/2 Pa (29.5 psig), was allowed because the inboard MSIVs are tested in the direction opposite normal flow and tend to lift when tested at higher pressures, resulting in a meaningless test.
This is particularly true of the Atwood Morril valves installed on Unit 1.
Experience has shown that the inboard valve will begin to lift at test pressures of approximately 28.5 psig, resulting in high leak rates which are not necessarily indicative of a degraded valve.
The reference NRC SE provided an exemption to Appendix J to allow a test pressure of 1/2 Pa.
When Amendment 131 to the TS was issued, the value "28 psig", previously stated in TS 4.7. A.2.h, was replaced with "l/2 Pa".
Another reference to the 28 psig test pressure, in Table 3.7.4, Note (2), was not changed.
We now request, as an administrative item, that the previous value of 28 psig be returned to TS 4.7. A.2.h.
If necessary, we request that the exemption to Appendix J be reevaluated and issued for the 23 psig test pressure rather than 1/2 Pa.
It is very unlikely that the MSIVs would pass leak rate testing at 1/2 Pa, when tested in the upcoming outage, due to the lifting of the inboard valve as explained above.
Thus, it is expected that this change will be necessary to allow unit startup from the upcoming outage.
Il64C El-5 03/04/87
ENCLOSURE 2 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH llVCLEAR PLANT UNIT 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINf1ENT ISOLATION VALVES 10 CFR 50.92 EVALUATION Pursuant to the requirements of 10 CFR 50.92, Georgia Power Company has evaluated the proposed amendment for the Plant Hatch Unit 1 Technical Specifications and has determined that its adoption would not involve a significant hazards consideration.
The bases for this determination are as follows:
PROPOSED CHANGE 1:
This proposed change would delete certain categories of valves from Table 3.7-4 of the Plant Hatch Unit 1 Technical Specifications (TS), pursuant to NRC's Safety Evaluation (SE) dated October 30, 1986.
Yalves deleted will no longer be subject to leak rate testing pursuant to 10 CFR 50 Appendix J.
This is an essentially administrative change as NRC has already provided the safety evaluation.
Basis for Proposed Change 1:
Penetrations and associated valves proposed for deletion are connected below the minimum water level of the torus and will remain water sealed during postulated post accident conditions.
The NRC has evaluated this design and determined that, since no potential containment leakage path exists due to the water seal, those valves should not be tested pursuant to the requirements of Appendix J.
This evaluation is documented in the reference SE.
These valves will continue to be tested pursuant to ASME code requirements.
Accordingly, the implementation of this change to the Technical Specifi-cations would not involve a significant hazards consideration, because:
1.
The probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety are not j
increased above those previously evaluated, because the valves have been determined by NRC not to represent a potential containment leakage path.
Thus, leak rate testing pursuant to l
Appendix J is not necessary to provide assurance of containment integrity.
ll64C E2-1 03/04/87 I
L ENCLOSURE 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES 10 CFR 50.92 EVALUATION 2.
The possibility of a new or different kind of accident from any previously evaluated would not result from this change, because the change does not represent a change to plant design or configuration.
This change only imposes proper testing requirements for certain valves based on an NRC evaluation..
3.
Margins of safety are not reduced because Appendix J testing is not required to maintain the containment integrity, due to design.
PROPOSED CHANGE 2:
This proposed change would delete valves 1G51-F002, F0ll, ana F012, in addition to blind flange IG51-D001 from Unit 1 Technical Specifications Table 3.7-4.
Basis for Proposed Change 2:
These torus drainage system components remain submerged following a LOCA and, therefore, do not constitute a potential bypass leakage pathway.
The basis is the same as Proposed Change 1 above; however, this penetration was not explicitly addressed by the reference NRC SE.
These components were added following the original GPC submittal which provided the basis for the SE.
Accordingly, the implementation of this change to the Technical Specifi-cations would not involve a significant hazards consideration, because:
1.
The probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety are not increased above those previously evaluated, because the valves have been determined not to represent a potential containment leakage path.
Thus, leak rate testing pursuant to Appendix J is not necessary to provide assurance of containment integrity.
2.
The possibility of a new or different kind of accident from any previously evaluated would not result from this change, because the change does not represent a change to plant design or configuration.
This change only imposes proper testing requirements for certain valves based on plant design.
3.
Margins of safety are not reduced because Appendix J testing is not required to demonstrate capability to maintain containment integrity.
Il64C E2-2 03/04/87
e ENCLOSURE 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES 10 CFR 50.92 EVALUATION PROPOSED CHANGE 3:
The proposed change would delete RHR torus spray valves 1 Ell-F02.7A, B from TS Table 3.7-4.
Basis for Proposed Change 3:
The proper inboard isolation barriers for these torus spray lines are the lEll-F028A, B valves, which are currently tested pursuant to Appendix J (in addition to the lEll-F027A, B valves).
However, the outer barrier is properly provided by the closed RHR system, as identified in Table 7.3-1 (note 30 for penetrations 211A and B) of the Plant Hatch Unit 1 FSAR.
Thus, Appendix J testing of the lEll-F027A, B valves shoul d not be required.
Closed systems located outside containment do not require Appendix J testing.
Since the system is a closed
- system, no potential bypass leakage pathway
- exists, and fission-product retention would be maintained following a design basis accident.
Accordingly, the implementation of this change to the Technical Specifi-cations would not involve a significant hazards consideration, because:
1.
The probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety are not increased above those previously evaluated, because containment isolation is properly provided by valves lEll-F028A and B and the closed RHR system.
2.
The possibility of a new or different kind of accident from any previously evaluated would not result from this change, because the change only serves to provide proper testing requirements for certain valves based on plant design considerations.
Plant operation is unaffected.
3.
Margins of safety are not reduced because containment integrity will be prbperly preserved.
PROPOSED CHANGE 4:
This proposed change would incorporate revisions which reflect accurate valve / penetration relationships and specify correct nomenclature.
Basis for Proposed Change 4:
These changes are administrative in nature and have no impact upon the plant safety analyses.
ll64C E2-3 03/04/87
~.
R.a ENCLOSURE 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES 10 CFR 50.92 EVALUATION PROPOSED CHANGE 4 (Continued):
Accordingly, the implementation of this change to the Technical Specifications would not involve a significant hazards consideration, because:
1.
The probability of the occurrence or the consequences of an accident or malfunction or equipment important to safety are not increased because the change is purely administrative in nature and has no impact on plant operation or the accident analysis contained in the FSAR.
2.
The possibility of a new or different kind of accident from any previously evaluated would not result from this administrative change.
3.
Margins of safety are not reduced because plant operation is not affected.
PROPOSED CHANGE 5:
This proposed change woul d provide consistency within the Technical Specifications relative to the pressure at which the MSIVs are leak tested.
The test pressure stated in Section 4.7. A.2.h would be changed from 1/2 Pa to the original value of 28
- psig, thus providing consistency with Note (2) of Table 3.7-4.
Basis for Proposed Change 5:
Each main steam line is provided with two MSIVs which are oriented to seal in the direction of post-accident containment atmosphere out-l eakage. The ddsign of the HSIVs is such that testing in the reverse direction tends to unseat the valve.
Simultaneous testing of the two valves, at one half design pressure, by pressurizing between the valves, tends to lift the disc of the inboard valve and result in a meaningless test.
To avoid lifting the disc of the inboard val ve, the proposed change calls for a return to the previously approved test pressure of 28 psig.
The total observed leakage through both valves (inboard and outboard) is conservatively assigned to the penetration.
This proposed change revises the test pressure setting to that previously ll64C E2-4 03/04/87
L EHCLOSURE 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES 10 CFR 50.92 EVALUATION PROPOSED CHANGE 5 (Continued):
contained in the plant Technical Specifications; therefore, this is an administrative change and the previously performed accident analyses are unaffected.
Accordingly, the implementation of this change to the Technical Specifications would not involve a significant hazards consideration, because:
1.
The probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety are not increased above those previously evaluated, because this change reflects the correct value as previously evaluated.
2.
The possibility of a new or different kind of accident from any previously evaluated would not result from this change, because the change simply specifies an appropriate pressure for the testing of the valves.
3.
Margins of safety are not reduced because the Technical Specifications allowable leakage rate is maintained.
Il64C E2-5 03/04/87
L ENCLOSURE 3 i
NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:
PRIMARY CONTAINMENT ISOLATION VALVES PAGE CHANGE INSTRUCTIONS The proposed changes to the Unit 1 Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as follows:
Remove Page Insert Page 3.7-6a 3.7-6a 3.7-24 3.7-24 3.7-25 3.7-25 3.7-26 3.7-26 3.7-26a 3.7-26a 3.7-26b 3.7-26b N
S
- On e
r J
4 j'
i
=^
-m n
J l
1164C 03/04/87
_ - - - - - - - -. _ _. - - _ _... _ _ _...., _ _. _ _ _ _ _ _ _. _ _