ML20207S783

Requests That NRC Reconsider Position Re Pressure Testing of RCPB Based Upon Listed Considerations.Urges Meeting to Resolve Concern Prior to Completion Refueling Outage.Asme Code Basis,Technical Considerations & Regulatory Basis Encl

ML20207S783
Person / Time
Site:	Hatch
Issue date:	03/13/1987
From:	Gucwa L GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
SL-2145, NUDOCS 8703200267
Download: ML20207S783 (12)

Text

Georgia Power Company 333 Piedmont Avenue Att:nti Georgia 30308 Tc cy.cr.s 404 52G4526 Maihng Address:

Post Office Box 4545 Atlanta, Georgia 30302 L Georgia Power L. T. Gucwa tN southern eitxtnc sgtem Manager Nuclear Safety and Licensing SL-2145 1199C March 13, 1987 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Hashington, D.C. 20555 NRC DOCKETS 500321, 50-366 OPERATING LICENSES DPR-57, NPF-5 EDHIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 HYDROSTATIC LEAK TESTS RE0 VEST FOR APPEAL MEETING Gentlemen:

In our February 24, 1987, letter to the NRC, we stated our intention to appeal the NRC staff's decision relative to the pressure testing of the reactor coolant pressure boundary (RCPB) at Plant Hatch. Georgia Power Company's (GPC) decision to request the NRC to reconsider its position is based upon several considerations, including the following:

1. He believe that the testing of the RCPB at pressure and elevated temperature is a more valid test than a water-solid test, since the pressurizing medium (steam, water, and noncondensibles) approximates that which would occur during plant operation.

2. Testing at an elevated temperature obviates any concerns regarding brittle fracture, since there is a significant margin (hundreds of degrees) to the adjusted reference temperature.

3. Based upon past history, we have shown that performing the pressure tests with the reactor critical can be done safely, efficiently, and adequately.

4. We believe that the responses provided by the American Society of Mechanical Engineers (ASME) to GPC's inquiries (as revised) concerning pressure testing support our position.

5. Since we have been performing the pressure tests with the reactor critical for a significant period of time, imposition of the NRC's position prohibiting criticality is clearly a backfit 8703200267 870313 PDR ADOCK 05000321 hgN P ppg ,I

Georgia Poker A U. S. Nuclear Regulatory Commission.

ATTN: Document Control Desk March.13, 1987 Page Two issue. However, in lieu of filing a formal backfit claim with the NRC, GPC desires to resolve this issue on a technical basis as generally suggested by the Deputy Executive Director of Operations in discussions with, and presentations to licensee representatives.

Enclosed are the ASME Code bases, technical / safety considerations, and . regulatory bases which support GPC's position. In order to' resolve this concern prior to the completion of the upcoming Hatch Unit I refueling outage, we would appreciate meeting with you to discuss the issues at your earliest convenience.

Should you have any questions or concerns, please contact- this office at any time.

Sincerely, d' &

L. T. Gucwa GKM/lc Enclosures c: Georaia Power Comoany U.S. Nuclear Reaulatory Commission Mr. J. P. O'Reilly Mr. J. H. Sniezek, Director, Mr. J. T. Beckham, Jr. Regional Operations and Generic Mr. H. C. Nix, Jr. Requirements Staff GO-NORMS Mr. R. M. Bernero, Director, Division of BWR Licensing Dr. J. N. Grace, Regional Administrator Mr. P. Holmes-Ray, Senior Resident Inspector - Hatch Mr. G. Rivenbark, Licensing Project Manager - Hatch 1199C 700775

GeorgiaPower1 ENCLOSURE 1 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 EDHIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 ASME BOILER AND PRESSURE VESSEL CODE. SECTION XI HYDR 0 STATIC LEAK TESTS A. ASME BOILER AND PRESSURE VESSEL CODE, SECTION XI CONSIDERATIONS

1. Core Criticality Durina Pressure Tests The central issue on which Georgia Power and the NRC disagree is whether the reactor can be critical when performing required hydrostatic and leakage pressure tests. In an NRC internal memorandum (Reference 1 of this enclosure), the NRC issued the following statement:

"The ASME Code intended both System Leakage and Hydrostatic Tests to be performed prior to reactor criticality from a refueling."

In verbal conversations with the NRC, followed up by our letter dated April 30, 1986, (Reference 2), we requested that the NRC grant temporary relief from the above-stated requirement relative to Plant Hatch Unit 1. He also committed to seek determination from the ASME Code Committee of whether Section XI of the ASME Boiler and Pressure Vessel Code prohibited the reactor from being critical when performing pr_ essure tests. The following revised inquiry was submitted to the ASME:

"DoesSection XI, Division 1, IHA-5211 require that a reactor be in a non-critical state when pressure tests (hydrostatic and leakage tests) are performed?"

In response to the subject inquiry, the ASME transmitted the following response to GPC:

" Core criticality during pressure testing is nat addressed by Section XI, Division 1." (emphasis added)

He are of the opinion that the ASME's response is not supportive of the NRC's position stated in Reference 1, as quoted above.

It should be noted that 10 CFR 50, Appendix G, specifically contemplates that hydrostatic testing may be performed when the reactor is critical, as indicated by Section IV.A.5 which states:

1199C El-1 03/13/87

Georgia Powerkh ENCLOSURE 1 ASME BOILER AND PRESSURE VESSEL CODE. SECTION XI HYDROSTATIC LEAK TESTS "If there is fuel in the reactor during system hydrostatic pressure tests or leak tests, the requirements of paragraphs 2 or 3 of this section apply, dep_eDJina on whether the core is critical durina the tts_t." (emphasis added).

He are theref3re of the opinion that the NRC's regulations corsidered that performing the tests while the reactor 1. critical is an acceptable method of achieving the raquired temperatures.

2. Testina Medium In Reference 1, the NRC stated:

" Prudence dictates that both of these tests be performed at the lowest temperatures that are consistent with the fracture prevention criteria for the reacter vessel or other component so that stored energy can be minimized during testing conditions by having the system water solid."

Based upon our verbal commitment regarding the granting of temporary relief -(Reference 2), GPC sought an ASME interpretation regarding whether the reactor must be water-solid when performing pressure tests. The following revised inquiry was submitted to the ASME:

"Section XI, Division 1, IHB-5210(b) states ' Reactor coolant shall be used as the pressurizing medium, when performing system pressure tests. Can a mixture of steam, water, and non-condensible gases in a proportion no greater than that present during normal startup, be used as the pressurizing medium to meet the requirements of IHB-5210(b) when performing system pressure tests of the primary reactor coolant systems, for Boiling Hater Reactors?'"

The ASME responded in the affirmative.

3. Test Temoerature/ Pressure As previously stated in 2. above, the NRC in Reference 1 stated:

1199C El-2 03/13/87 700775

Georgia Power d ENCLOSURE 1 ASME BOILER AND PRESSURE VESSEL CODE. SECTION XI HYDR 0 STATIC LEAK TESTS

" Prudence dictates that both of these tests be performed at the lowest temperatures that are consistent with the fracture prevention criteria for the reactor. vessel or other component..."

In regard to the NRC's contention that hydrostatic and leakage tests be performed at the " lowest possible temperatures that are consistent with the fracture prevention criteria...", we present the following information for consideration. ASME Code,Section XI, Article IHB-5222, allows performing the tests at . elevated temperatures (and reduced pressures) to meet fracture prevention criteria. The fracture prevention requirements stated in 10 CFR 50, Appendix G, preclude pressurizing the vessel to certain levels unless the vessel is at temperatures in excess of those calculated in accordance with the Appendix G requirements.

Contrary to the NRC position, from a fracture prevention standpoint, it is prudent to test at temperatures far in excess of those required per 10 CFR 50, Appendix G.

In Reference 1, the NRC stated:

"The later Code position (footnote 7 to Table IHB-2500-1, Category B-P) permitting the System Hydrostatic Test to be used in lieu of the Leakage Test is a clear indication that the code intended the system hydrostatic [ sic] also be performed at low temperatures consistent with fracture prevention considerations prior to reactor start-up."

Footnote 7 states "A system hydrostatic test (IHB-5222) and the accompanying VT-2 examination are acceptable in lieu of the system leakage test (IHB-5221) and VT-2 examination." It is our opinion that the footnote was written to preclude performing both tests, when a hydrostatic test is required, since any leakage would be detected by that test.

4. Test Puroose In Reference 1, the NRC staff also stated:

"The Hydrostatic Test is a proof telt of repairs on the reactor coolant pressure boundary or other component." (emphasis added).

Il99C El-3 03/13/87 700775

a Georgia Powerkh ENCLOSURE 1 ASME BOILER AND PRESSURE VESSEL CODE. SECTION XI HYDR 0 STATIC LEAK TESTS However, the NRC's January 5,1987, letter to the chairman of the BHR Owners Group (Reference 4), transmitted the following information which appears to be different than the Reference 1 statement.

"Despite the fact that the Code-required hydrostatic test is intended to reveal gross negligence in a repair operation, the principal use of the hydrotest (1100 psig) as well as the leak test is to perform a visual leak test at a pressure equal to or slightly higher than operating pressure. Given the history of pipe cracks in BHRs, the staff feels that the visual inspection for leaks during pressure tests is very important. Although the hydrotest may provide some information on reactor vessel integrity, it is doubtful that hydrotests serve at a proof leit, and they surely do not serve as a substitute for adequate inspection when assessing future reactor pressure vessel integrity." (emphasis added).

B. TECHNICAL / SAFETY CONCERNS

1. Personnel Safety and Leak Detection In Reference 1, the NRC stated:

"He believe the Code section allowing reduction of temperatures below 200*F at corresponding pressures is prudent for the visual examination in that risk to plant personnel is reduced and any leakage would be liquid and, therefore, more readily detectable."

Georgia Power Company considers personnel safety to be a prime consideration in all activities conducted at Plant Hatch. Our dedication to assuring that all work is performed as safely as possible is evidenced by our safety record. That is, at Plant Hatch approximately nine million manhours have been worked in which a lost-time accident has not occurred. This record now stands as the "best" in the nuclear power industry.

Historically, the ASME-required VT-2 examinations have identified steam leaks around flanges and valve stem packing during

" critical" pressure testing. Steam leaks are very identifiable due to the steam " cloud" in the immediate vicinity of the leak.

Il99C El-4 03/13/87 700775

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n i Georgia Pbwer d

> ENCLOSURE 1 ASME BOILER AND PRESSURE VESSEL CODE. SECTION XI HYDROSTATIC LEAK TESTS 4

The minimal unidentified drywell leakage encountered once the units are placed back- into service. is- a testimony to the quality of ' the inspections. As previously stated, the purpose of the pressure tests is to perform a leak test at, or s_ lightly higher

.than~ operating pressure. The test should also be performed at an T'

elevated temperature,. thereby more closely duplicating the

' conditions during the operating cycle.

Since some areas of the drywell, such as the drywell head area, are not accessible when L the reactor is critical, we asked General Electric to address the

adequacy of the leak detection instrumentation in the drywell. A i copy of General Electric's letter- (G-GPC-6-156, dated April 22, '.

, 1986). is enclosed. General Electric concluded that instrumentation exists to identify leakage within the drywell area. He emphasize that inspections for leaks in other areas of the plant are-conducted when the plant is in operation, and the temperatures and pressures in lines, such as the main steam c

~~

lines, are comparable to those encountered during an ASME pressure test when the reactor is critical.

2. Effect of Elevated Temoerature on Vessel Integrity of the vessel is an overriding concern during plant testing .or operation. Performing the pressure tests at the elevated temperature obtained during critical- heatup obviates any concern regarding ' brittle fracture. As far as other reactor coolant- pressure boundary components (pipes, pumps, valves) are

. concerned, testing at an approximate operating temperature and pressure is a more valid test.

3. Criticality Performing pressure tests with the core critical is not a safety

-concern, because:

a. Since -a significant amount of time transpires between the

. time the reactor is shutdown and the performance of the pressure tests, fission products will have substantially decayed.

b. The reactor power is low.

c. The probability of an accident occurring during the limited time in which the inspection is taking place is extremely low.

4 1199C El-5 03/13/87 700775

GeorgiaPowerA ENCLOSURE 1 ASME BOILER AND PRESSURE VESSEL CODE. SECTION XI HYDROSTATIC LEAK TESTS

-The enclosed General Electric letter dated March 6, 1987, from Mr. R. F. Daly to Mr. L. T. Gucwa (GPC), addresses the safety concerns associated with criticality. The letter concludes that testing with the reactor critical results in no added safety considerations..

C. REGULATORY CONSIDERATIONS As previously stated, it is our view that 10 CFR 50, Appendix G, anticipated that system pressure tests might be performed while the reactor is critical. Additionally, we wish to note that GPC historically has performed pressure tests with the reactor critical.

In a letter dated May 5, 1986, (Reference 3), the NRC acknowledged GPC's past history. He believe that the backfitting requirements of 10 CFR 50.109 are applicable to this issue.

REFERENCES:

1. Internal memorandum from Robert M. Bernero (NRC) to Albert f. Gibson (GPC), dated April 29, 1986.

2. Letter from L. T. Gucwa (GPC) to Daniel R. Muller (NRC), dated April 30, 1986.

3. Letter from Daniel R. liuller (NRC) to J. T. Beckham, Jr. (GPC), dated May 5, 1986.

4. Letter from G. A. Arlotta (NRC) to T. A. Pickens (BWROG), dated January 5, 1987.

1199C El-6 03/13/87 1 #,n5

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2 GENERAL $ ELECTRIC NUCLEAR ENERGY BU$1 Ness OPERATIONS GENERA'. ELECTRIC CCWANY. T9 TECHNOLOGY WK NORCEO55. GEOGA 3D2 PO. BOX l'80*A

• ATLANTA, GCC,RGIA 333A9 G-GPC-6-156 cc: Georgia Power Company April 22, 1986 J.T. tieckham, Jr.

P.R. Bemis B.K. McLeod H.C. Nix, Jr. (2)

Mr. L.T. Gucwa J.P. O'Reilly Manager, Nuclear Safety Southern Company Services and Licensing Department W.F. Garner Georgia Power Company L.B. Long P.O. Box 4545 General Electric Company Atlanta, GA 30302 E.O. Swain H.A. Upton R.R. Willems

SUBJECT:

JUSTIFICATION OF USING NUCLEAR STEAM AT OPERATING PRESSURE AND TEMPERATURE FOR HYDR 0 STATIC LEAK TEST - HATCH 1 & 2

Dear Mr. Gucwa:

Concern It is Georgia Power Company practice to perform the system leakage test utilizing nuclear steam at operating pressure and temperature after a reactor outace. ASME Section XI,1980 Edition Article IW3-5000 (System Pressure Tests) Table IWB-2500-1 examination category B-P (for all p(ressure IWB-5221)retaining components) shall be conducted prior toNote plant5startup says following that "theeach system leakage tes reactor outage". ASME Section XI, Article IWA-5240, also requires that a visual examination (VT-2) be performed during the system leakage test.

A visual examination of the reactor vessel was not performed during the system leakage hot testing due to the hot environment (radiation and temperature)atthoselocationsinthedrywell.

l

Response

. It is our opinion that the nuclear steam system leakage test performed l by GPC as described above is acceptable. It is also our opinion that i the nuclear steam system leakage test performed (hot) is less

! detrimental to reactor pressure vessel and components than performing the leakage test prior to plant startup. Several steam leakage l detection devices are available to detect any level of steam leakage and

location of leakage source. Therefore, visual examination during the

! system leakage test can be waivei because of the following built-in I feature in the design.

1. The availability of the reactor pressure vessel main closure seal, which is double seal designed to have no leakage through the inner or outer seal at all normal and upset operating conditions as well as testing conditions. The seal cavities are located in the vessel head flange face. Each seal cavity contains a metal 0-ring.

l- ..

-*+

t G-GPC-6-156 April 22, 1986 Page 2 l

A vent is installed in the vessel flange between the 0-rings t

seals to indicate any leakage of water or steam through the inner seal. The vent is communicated with the outside of the inner seal grcove and the inside of the outer seal groove through the seal leak detection nozzle and extension with the leak detection system which trips an alarm in case of a leakage. The seal vent is designed for the full design pressure of the vessel.

Hatch-2 Document No. 922D167 Rev 9, Seal Detection Nozzle No. N-13 VPF #3062-213-3 Leakage from the Reactor Vessel Head Seal Leak Off is routed to instruments (821-N001 level switch, B21-N002 pressure switch, and B21-R001 pressure indicator). Here the flow is gathered and when flow is excessive, a high ficw alarm is indicated in the control room. Following the instruments, the drain line is double valved with both valves normally closed. So leakage would have to cross both valves to drain to the drywell equipment sump.

This sump flow is also measured and alarmed in the control room.

The drywell equipment sump is considered to be part of the drywell identified leakage. The total drywell identified leakage limit of 25 gpm in tech specs will also be another check on excessive flow.

References:

761E250BA B21-1010 P&ID Hatch 2 729E616BA B21-1010 P&l0 Hatch 1 22A1441AC A61-4040 Nuclear Boiler Hatch 1 22A1441AU A61-4040 Leak Detection Sys Hatch 2

2. The availability of several thermocouple pads which n'onitor the changes in temperature. These pads are located at vessel flange, vessel wall adjacent to flange. For more detail See Drawing No. 9220167 Rev. 9, sheet 5. ,

3. Drywell ambient temperature is another indicator of reactor coolant or steam leakage. Terrperature elements are located on the drywell wall and send signals to recording and alarm circuit on the back panel in main control room.

Very t ly yours, R.P. Daly Services Project Manager RPD/ rah b- .-

l

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GENERAL $ ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS GENERAL ELECTRIC COMPANY + 99 TECHNOLOGY PARK e NORCROSS, GEORGIA 30092 RO. BOX 105064

• ATLANTA, GEORGIA 30348 C-CPC-7-075 March 12, 1987 cc: Georgia Power Company J. T. Beckham, Jr.

P. R. Bemis J. D. Heidt H. C. Nix, Jr.

J. P. O'Reilly General Electric Cpmpany G. L. Sozzi R. R. Willems Mr. L. T. Cucwa Manager, Nuclear Safety & Licensing Georgia Power Company Pont Office Box 4545 Atl.anta, Ga 30302

Subject:

Plant Hatch Leak Rate Testing

Dear Mr. Gucwa,

Enclosed is a copy of the safety evaluation which justifies the reactor coolant pressure boundary leakage testing following a major outage while the reactor is critical. Please contact me if you have any questions regarding this transmittal.

Very er ly yours, lt.

9' l R. P. Daly Services Project Manager RPD/it Enclosure 9 0 bM 9NI

l LEAKAGE TESTING WITH CORE CRITICAL -l March 6, 1987 Hydrostatic or leaka5e Pressure tests of the reactor coolant pressure boundary are required. when returning to service following a refuelin5 or other major outage. To achieve high reactor system temperature either pump heat or auxili-ary steam, or reactor nuclear generated steam can be utilized. General Electric Company has evaluated the reactor safety concerns of performing those tests for the Plant Hatch units while the reactor core is critical and has concluded that there are no added safety implications for the following reasons:

1. The tests are only performed upon returning to service following major outages (approximately once per year), and last only a short duration (few hours).

2. The tests are performed while reactor power is low (less than 104).

Therefore, the implications or consequences of hypothetical loss-of-coolant accident and other post.ulated transient events is si 5nifi-cantly less severe than any previously analyzed events or accidents.

Further, the probability of such an event occyrring during the short test duration is jud5ed. to be less than 10 and thus is considered an incredible event.

3. Following an outa5e, and at low core power, there are little fission products present, and hence decay heat. Any hypothetical accident or transient initiated from these conditions will not lead to fuel damage or fission product releases.

Based on the above, it is concluded that if performing the hydrostatic or leakage test when the reactor is critical (relative to usin5 non-nuclear heat):

1. There is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety,

2. There is no increase in the possibility of an accident or malfunction of a different type than analyzed, and

3. There is no teduction in the mar 81 n of safety as defined in the basis for the Technical Specifications.

i Approved by: dO v1/7 __

[IL. So/zi,pa[r,er Application En5 1neering Services GENERAL $ ELECTRIC

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