ML20212M977

Application for Amend to License DPR-57,allowing Performance of Hydrostatic or Leak Testing W/Noncritical Reactor Core. Fee Paid

ML20212M977
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
ML20212M979	List: ML20212M977 ML20212M987
References
SL-2084, NUDOCS 8703120298
Download: ML20212M977 (12)

Text

N:

~ 1 Georgia Power Company

' v: _ ,* *- ' 333 Piedmont Avenue .

".e  ; Atlanta, Georgia 30308' Telephone 404 526-7851

'g Mailing Address; ~

' Tbst Office Box 4545 (\1i Atlanta, Georgia 30302 -

- r;. <

' Ja es P O'Reill the Southern MrfnC Sy$ fetu

" Nuclear Operations .

SL-2084 1173C i

' March 4, 1987

'U. S. Nuclear Regulatory Consission 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 l I

TO ALLOW PERFORMANCE OF HYDROSTATIC AND LEAK TESTING USING NON-NUCLEAR HEAT l

Gentlemen:

In accordance with' the provisions of 10 CFR 50.90, as required by 10 CFR 50.59(c)(1), Georgia Power Company hereby proposes changes to the Technical Specifications, Appendix A to Operating License DPR-57 -for Plant Hatch Unit 1 The proposed . changes are necessary to allow . performance of hydrostatic or leak testing with a non-critical reactor' core. We request approval of these changes as soon as possible, to support testing i following the upcoming Unit 1 refueling outage, currently scheduled to

~

comence on approximately April 22, 1987.

Enclosure 1 provides detailed descriptions of the proposed changes and the bases for the' change request.

Enclosure 2 details the basis for our determination that the proposed changes do not-involve significant hazards considerations.

Enclosure 3 provides page change instructions for incorporating the

. proposed changes into the Plant Hatch Unit 1 Technical Specifications.

The proposed changed Technical Specifications pages follow Enclosure 3.

I 19?

yI ifC

'8703120298 870304 PDR

'P' ADOCK 05000321 PDR f(505

U. .S. Nuclear Regulatory Commission.

March 4,1987

, Page Two' Payment. of the filing fee in the amount of one hundred and fifty.

dollars' is enclosed.

Pursuant .to the requirements . of 10 CFR 50.91, a copy of this letter and all ca pplicable enclosures will be sent to Mr. J. L Ledbetter of the

- Environmental Protection Division of the . Georgia Departn:ent of Natural Resources.

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 the letter and enclosures are true.

GEORGIA POWER C0fPANY 2

1 By: catik. 6 b James P. O'Reilly f March 1987.

t Sworn to and subscribed before me this 4

= w Y Yd>tb

, l %tY Public Cb

""'Y' '*T' M1Commnsston E*** U* - 12 1989 REB /lc Enclosures c: Georcia Power Company U.S. Nuclear Regulatory Connission Mr. s . T. Beckham Dr. J. N. Grace, Regional Administrator Mr. H. C. Nix, Jr. Mr. P. Holmes-Ray, Senior Resident

' GO-NORMS Inspector-Hatch Mr.' G. Rivenbark, Licensing Project Manager - Hatch

- State of Georgia q

Mr. J. L. Ledbetter l

i 3

1- I l 1173C

,A_j. _ .a A-; 44-- ---+b. A s. 4 A e b y e 4 -.4--m e,- ------r-.

., . _ . e a4.-

- ; ? ,- . . m ENCLOSURE ~ 1 1.

1-

.NRC DOCKET 50-321

! - OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 E REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW PERFORMANCE OF HYDROSTATIC

- AND LEAK TESTING USING NON-NUCLEAR HEAT I

BASIS FOR CHANGE REQUEST Georgia Power Company (GPC) has historically used nuclear heat for 1 l performance of inservice- hydrostatic and leak testing for the Plant Hatch

. Units. This testing 'is performed with .a critical core. The ASME code allows for the use of a reduced test pressure wh en th e test is performed at

] elevated. temperatures, as is the case for a hydrostatic test using nuclear heat. Because of the elevated temperatures and reduced pressures used for such testing, and the fact that nuclear steam is being produced, these i tests may be -performed, for Plant Hatch Unit 1, under the requirements of i ,

the current Plant Hatch Unit 1 Technical Specifications (TS).

GPC's use of nuclear heat for the performance of such testing is now being questioned by NRC. Should the use of nuclear heat be -disallowed for future hydrostatic and leak testing, recirculation pump operation - and a water-solid reactor pressure vessel (except for an air bubble for pressure control) must be used to achieve necessary temperatures and pressures. The tests would be performed with all~ control rods in the core and the mode i switch .in the shutdown or refuel position. The use of non-nuclear heat for the performance of these tests results in the need for TS revisions.

Two areas of change are necessary. The first involves addition of footnotes to those systems, whose operability is technically required by TS i during the pressure / temperature regimes encountered in the non-nuclear

! hydrostatic or leak test, 'but which, because of test conditions, cannot be operable. The second involves a refinement of TS Figure 3.6-1 which will

provide multiple vessel ' neutron exposure curves for determining NDT shift l and resul tant minimum temperature versus pressure requirements for performance of tests. This will allow tests to be conducted with minimum '

j. heatup.

f PROPOSED CHANGE 1:

- Hydrostatic or leak testing using non-nuclear heat is performed near the lowest temperature allowed by the reactor pressure vessel

, pressure / temperature limit curves (TS Figure 3.6.1), and at a pressure

equal to (leak testing) or greater than (hydrostatic testing) normal operating pressure. The following table provides approximate pressures and temperatures at which future non-nuclear hydrostatic or leak testing would be expected to be perfonned ;if required), based on vessel neutron fluence (see Proposed Change 2), in Effective Full Power Years (EFPY). The listed l temperatures are minimum RPV metal temperatures and actual coolant temperatures may be somewhat higher.

I 1173C El-1 03/04/87 i

I

~

ENCLOSURE 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW PERFORMANCE OF HYDROSTATIC AND LEAK TESTING USING NON-NUCLEAR HEAT BASIS FOR CHANGE REQUEST PROPOSED CHANGE 1 (Continued):

Leak Test Hyrdostatic Test EFPY Temperature Pressure Temperature Pressure 8 1920F 1005 psig 1990F 1106 psig 10 2020F 1005 psig 2100F 1086 psig 12 2110F 1005 psig 2190F 1086 psig 14 2200F 1005 psig 2280F 1086 psig 16 2280F 1005 psig 2360F 1086 psig The High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems are required to be operable by TS 3.5.D.1 and 3.5.E.1, respectively, whenever irradiated fuel is in the reactor vessel and the reactor pressure is greater than 1050 psig. These two criteria are met for performance of the hydrostatic or leak test. However, the primary system will be water-solid and no steam will be available to drive the HPCI and RCIC turbines. Thus, these systems cannot physically be operable for performance of the tests. The following note is thus proposed for addition to the TS for these two systems:

HPCI (RCIC) is not required to be operable for performance of inservice hydrostatic or leak testing, with reactor pressure greater than 150 psig and c.11 control rods inserted.

As can be seen from the above table, performance of non-nuclear hydrostatic testing will require test pressures greater than the lift pressures for the safety / relief valves (S/RVs). In order to perform the test, the S/RVs must be gagged to prevent their opening. Because of this, the TS requirements for the S/RVs and the automatic depressurization systems (ADS) must be addressed. Currently, the S/RVs l and ADS are required by TS 3.6.H.1 and 3.5.F, respectively, to be operable when ever there is irradiated fuel in the reactor vessel and the reactor is above 113 psig. Again, these two criteria are met for performance of the hydrostatic test. However, these systems cannot be operable since the S/RVs will be gagged. The following note is thus preposed for addition to the TS for these two systems:

l

The ADS (Relief / Safety) valves are not required to be operable for performance of inservice hydrostatic or leak testing, with reactor pressure greater than 113 psig and all control rods inserted.

1173C El-2 03/04/87

T ENCLOSURE 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS

-TO ALLOW PERFORMANCE OF HYDROSTATIC AND LEAK TESTING USING NON-NUCLEAR IIEAT BASIS FOR CHANGE REQUEST PROPOSED CHANGE 1: (Continued):

Since the hydrostatic or leak test is perfomed water-solid, with all rods inserted into the reactor core, at low decay heat values, and at or near cold shutdown conditions, the stored energy in the reactor core will be very small. The above high pressure Emergency Core Cooling and overpressurization protection functions are incapable of functioning because of test conditions which are NR0 mandated. Thus, the above-requested changes are necessary and appropriate. It should be noted that Standard Technical Specifications, as currently written, do not require relief in the above areas for performance of the hydrostatic or pressure tests (because of Operational Condition applicabilities) as long as the reactor coolant temperature is below 2120F.

PROPOSED CHANGE 2:

TS Figure 3.6-1 provides pressure / temperature limits for performance of hydrostatic and pressure testing based on vessel neutron fluence. This figure currently contains a single curve based on 16 EFPY of operation.

The actual EFPY value for the Spring 1987 Hatch Unit 1 outage will be approximately 7.12 EFPY. It is proposed to refine Figure 3. 6-1 to replace the single 16 EFPY curve with a set of curves in two EFPY increments from 8 EFPY to 16 EFPY, inclusive. This will eliminate excessive conservatism and allow tests to be performed at lower temperatures, which require less recirculation pump heatup time (potential critical path outage time). The spring 1987 test will be performed using the 8 EFPY curve. Figure 3. 6-1 is applicable for quasi-steady state temperature. For vessel heatup and cooldown involving larger temperature versus time gradients, Figure 3.6-2 is applicable.

This figure is likewise based on 16 EFPY of operation. However, no change is propose'd to this figure since the heatup and cooldoun operations will not approach the limiting temperature / pressure curve.

Thus, the extra conservatism is not a factor.

This change is essentially administrative in nature and will allow testing to be performed with the minimum amount of stored energy in the primary system necessary to meet vessel NDT requirements.

1173C El-3 03/04/87

ENCLOSURE 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW PERFORMANCE OF HYDR 0 STATIC AND LEAK TESTING USING HON-NUCLEAR HEAT BASIS FOR CHANGE REQUEST PROPOSED CHANGE 3:

For the purpose of future hydrostatic or leak tests which will require reactor coolant temperatures greater than 2120F, it is requested that an exemption be granted from the current requirement of TS 3.7. A.2.c, which specifies that Primary Containment Integrity be maintained at all times whenever the reactor water temperature is above 2120F and fuel is in the reactor vessel.

The performance of hydrostatic or leak testing is for the purpose of inspecting the primary system for leakage. This inspection requires frequent and unimpeded access to potential leakage points inside containment. Many utilities perform the hydrostatic or leak test with the containment head removed for this purpose. Now that NDT limit shifts will push the test temperature near 2120F, TS relief is necessary to allow continued unimpeded access to containment. Also, critical path may be affected if all containment testing and repair operations necessary to assure primary containment integrity are required to be complete before performance of the hydrostatic or leak test.

TS 3.7. A.2.c requires establishment of containment integrity prior to pulling rods, or whenever the reactor is critical, as well as whenever the reactor coolant temperature is above 2120F. Primary containment is a pressure vessel designed to contain the amount of energy that would be released from the design basis accident. TS 3.7. A.2.c is intended to cover those situations where large amounts of stored energy are contained in the primary system. However, for the purpose of hydrostatic or leak testing, the amount of contained energy is very small. This is due to a non critical core (all control rods inserted), water solid conditions, low temperatures (maximum hydrostatic test temperature of 2360F at 16 EFPY), and low fuel decay heat values (no critical operation for the length of the outage).

The consequences of testing above 2120F are potential steam, rather than water, leaks. There is no mechanism to impart fission products into the reactor coolant. Therefore, the amount of radioactivity contained in postulated steam leaks will be very small and well within the capabilities of secondary containment and the standby gas treatment system. The secondary containment is designed to handle the consequences of airborne radiation and steam leaks and will be operable, pursuant to TS 3.7.C.2, for the performance of hydrostatic or leak testing. Plant Hatch Unit 1 FSAR Section 14.4.5.1.3 provides a description of the Main 1173C El-4 03/04/87

L ENCLOSURE 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW PERFORMANCE OF HYDROSTATIC AND LEAK TESTING USING NON-NUCLEAR HEAT BASIS FOR CHANGE REQUEST PROPOSED CHANGE 3 (Continued):

Steam Line Break Outside Containment. A total of 20,000 pounds of high energy steam and 120,000 pounds of water are released to secondary containment in this postulated event. We believe the capability of secondary containment to contain any leakage or radiation which could escape in the hydrostatic or leak test is bounded by this evaluation.

The note added to TS 3.7.A.2 will specify that operation without primary containment integrity and temperatures above 2120F is permissible only for the performance of hydrostatic or leak testing with all control rods inserted.

1173C El-5 03/04/87

L ENCLOSURE 2 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW PERFORMANCE OF HYDR 0 STATIC AND LEAK TESTING USING HON-HUCLEAR HEAT 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 (TS) 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 provide footnotes excepting operability requirements for RCIC, HPCI, ADS, and S/RVs for the performance of a non-nuclear hydrostatic or pressure test.

Basis for Proposed Change 1:

These systems cannot be operable for performance of a hydrostatic or leak test because of test conditions. Since the hydrostatic or leak test is perfonned water-solid, with all rods inserted into the reactor core, at low decay heat values, and at or near cold shutdown conditions, the stored energy in the reactor core will be very small. The above high pressure Emergency Core Cooling and overpressurization protection functions are incapable of functioning because of test conditions which are NRC mandated. Thus, the above-requested changes are necessary and appropriate. It should be noted that Standard Technical Specifications, as currently written, do not require relief in the above areas for performance of the hydrostatic or pressure tests (because of Operational Condition applicabilities) as long as the reactor coolant temperature is below 2120F.

Accordingly, the implementation of this change to the Technical Specifi-cations would not involve a significant hazards consideration, because:

i. 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 test occurs with minimal primary system energy due to all control rods being inserted, low temperature, and low fuel decay heat values. Such testing is allowed under the provisicns of DWR Standard Technical Specifications without the need for Technical Specification changes.

1173C E2-1 03/04/87

L ENCLOSURE 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

PRIMARY CONTAINMENT ISOLATION VALVES 10 CFR 50.92 EVALUATION PROPOSED CHANGE 1 (Continued):

, 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 provides for performance of ASME Codo required hydrostatic and pressure testing with all control reds inserted.

3. Margins of safety are not reduced because plant operation is not affecteo and analyzed margins of safety are unchanged.

PROPOSED CllANGE 2:

This proposed change would provide a refinement to TS Figure 3.6-1 by providing separate pressure / temperature limit curves for different reactor pressure vessel neutron fluence values.

Basis for Proposed Change 2:

This essentially administrative change will allow performance of hydrostatic and pressure testing at lower temperatures, while still ensuring that minimum vessel temperature requirements are met.

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 change only provides a refinement for the pressure / temperature limit curves used for performance of hydrostatic and leak testing.

Plant operation is unaffected.

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 provides for flexibility in the performance of ASME Code required hydrostatic and pressure testing with all control rods inserted.

3. Margins of safety are not reduced because plant operation is not affected and analyzed margins of safety are unchanged.

I173C E2-2 03/04/87 t----_-_--__ - - - _ - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - _ - - - - - - - - - _ - - -_---- _ --- - ---- _ --_ _ __---_--_-- _ -- _ --- _--- _ -_----

L EHCLOSURE 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

PRIMARY CONTAINMENT ISOLATION VALVES 10 CFR 50.92 EVALUATION PROPOSED CHANGE 3:

This proposed change would allow hydrostatic or pressure testing with all control rods inserted, at reactor coolant temperatures greater than 2120F, without requiring primary containment integrity.

Basis for Proposed Change 3:

This change will allow for flexibility and expediency in the performance of hydrostatic and pressure testing. TS 3.7. A.2.c requires establishment of containment integrity prior to pulling rods, or whenever the reactor is critical, as well as whenever the reactor coolant temperature is above 2120F. Primary containment is a pressure vessel designed to contain the amount of energy that would be released from the design basis accident. TS 3.7.A.2.c is intended to cover those situations where large amounts of stored energy are contained in the primary system. However, for the purpose of hydrostatic or leak testing, the amount of contained energy is very small. This is due to a non critical core (all control rods inserted), water solid conditions, low temperatures (maximum hydrostatic test temperature of 2360F at 16 EFPY), and low fuel decay heat values (no critical operation for the length of the outage).

The consequences of testing above 2120F are potential steam, rather than water, leaks. There is no mechanism to impart fission products into the reactor coolant. Therefore, the amount of radioactivity contained in postulated steam leaks will be very small and well within the capabilities of secondary containment and the standby gas treatment system. The secondary containment is designed to handle the consequences of airborne radiation and steam leaks and will be operable, pursuant to TS 3.7.C.2, for the performance of hydrostatic or leak testing. Plant Hatch Unit 1 FSAR Section 14.4.5.1.3 provides a description of the Main Steam Line Break Outside Containment. A total of 20,000 pounds of high energy steam and 120,000 pounds of water are released to secondary containment in this postulated event. We believe the capability of secondary containment to contain any leakage or radiation which could escape in the hydrostatic or leak test is easily bounded by this evaluation.

Accordingly, the impicmentation of this change to the Technical Specifi-cations would not involve a significant hazards consideration, because:

1173C E2-3 03/04/87

L ENCLOSURE 2 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

PRIMARY CONTAINMENT ISOLATION VALVES 10 CFR 50.92 EVALUATION PROPOSED CHANGE 3 (Continued):

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 test occurs with minimal primary system energy due to all control rods being inserted, low temperature, and low fuel decay heat values.

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 operation. This change only provides for performance of ASME Code required hydrostatic and pressure testing with all control rods inserted.

3. Margins of safety are not reduced because plant operation is not affected and analyzed margins of safety are unchanged.

Il73C E2-4 03/04/87

ENCLOSURE 3 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW PERFORMANCE OF HYDROSTATIC AND LEAK TESTING USING NON-NUCLEAR HEAT PAGE CHANGE INSTRUCTIONS The proposed changes to the Unit i Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as follows:

Remove Page Insert Page 3.5-6 3.5-6 3.5-8 3.5-8 3.5-9 3.5-9 3.6-9 3.6-9 Figure 3.6-1 Figure 3.6-1 3.7-2 3.7-2 1173C 03/04/87 4