ML17054A187

Clarifies & Suppls 780321 & 830502 Applications for Amend to License DPR-63 Changing Tech Specs Re Reactor Vessel Matl. Max Copper & Phosphorus Content of Vessel Plate Fabricator 0.27% & 0.021%,respectively

ML17054A187
Person / Time
Site:	Nine Mile Point
Issue date:	10/26/1983
From:	Mangan C NIAGARA MOHAWK POWER CORP.
To:	Vassallo D Office of Nuclear Reactor Regulation
Shared Package
ML17054A188	List: ML17054A187 ML20081D442
References
NUDOCS 8311010184
Download: ML17054A187 (42)

Text

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NO DOCKET FACIL;50 220 Nine Mile Point Nuclear -Station<

Uni~t it Niagara Powe 05000220 AUTH,NAME AUTHOR AFFILIATION MANGAN<C,V>

Niagara Mohawk"'power Corp, RECIP ~ NAME RECIPIENT AFFILIATION VASSALLOiDBB Operating Reactors Branch 2

SUBJECTS Clarifies L suppls 780321

'8 830502 'applications for.

amend -to License DPR-63 changing Tech" Specs re reactor vessel',matls Max copper 8, phosphorus content of vessel plate fabricator 0>27K 8 0+-021'Airespectively, DISTRIBUTION CODE:

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NIAGARAMOHAWKPOWER CORPORATION/300 ERIE BOULEVARDWEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474-1511 October 26, 1983 Director of Nuclear Regulation Attention:

Mr. Domenic B. Vassallo, Chief Operating Reactors Branch No.

2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Re:

Nine Mile Point Unit 1

Docket No. 50-220 DPR-63

Dear Mr. Vassallo:

Our January 31, 1978 letter provided information regarding the Nine Mile Point Unit 1 reactor vessel material.

That submittal indicated that the plate material contained no copper.

However, the weld material had a maximum level of 0.05 percent copper associated with it.

Based on that information and Regulatory Guide 1.99, Revision 1, an application for amendment to our Operating License DPR-63 was submitted on March 21, 1978.

That application constituted a change to Figure 3.2.2.c of the Nine Mjle Point Unit 1 Technical Specifications.

In your March 28, 1983 letter, you requested additional information regarding the basis for that change.

Our submittal of April 20, 1983 reiterated that the most limiting material was the weld wire with 0.05 percent copper.

That information was provided by the vessel fabricator.

That data indicated an absence of copper in the plate material.

Upon further investigations with the vessel plate fabricator, the maximum copper and phosphorus content of the plate material was found to be 0.27 percent and 0.021 percent, respectively.

This data is shown in the attached "Chemical Laboratory Heat Analysis" records from the vessel plate fabricator.-

Therefore, the most limiting material is the plate material and not the weld wire.

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Based on the above, the minimum reactor vessel temperature for pressurization curves have been revised.

These revised curves incorporate the recent changes to 10CFR50, Appendix G (Federal

Register, Vol. 48, No. 104, May 27, 1983).

We are providing additional information in this submittal to clarify and supplement our application for amendment to Operating License DPR-63 submitted March 21, 1978.

In addition, although not directly related to these curves, we are also providing a change to our May 2, 1983 application for amendment to Operating License DPR-63 which addressed reactor vessel material samples.

This is due to an editorial change.

The pages have been retyped in their entirety.

Revisions are indicated with marginal revision bars.

The following revisions should be incorporated.

1.

Replace page 77 in our current Technical Specifications with the attached revised page 77.

Specification 4.2.2.b has been deleted because the requirement has been fulfilled.

Therefore, Specification 4.2.2.c becomes 4.2.2.b.

2.

Replace page 78 of our May 2, 1983 amendment letter with the attached revised page 78.

Specification 4.2.2.c was moved to page 77 and changed to Specification 4.2.2.b.

Our May 2, 1983 application for amendment provides supporting information on the changes to Specification 4.2.2.b (formerly Specification 4.2.2.c).

3.

Replace pages 79, 80 and 81 in our current Technical Specifications with the attached revised pages 79,

79a, 80,

80a, 81,

81a, 82 and 82a.

The revised pages include revised curves and tables for minimim reactor vessel temperature for pressurization.

4.

Delete page 81 in our March 21, 1978 amendment letter and our current Technical Specifications altogether.

The information contained on this page has been incorporated into the revised curves.

5.

Replace page 82, Bases for 3.2.2 and 4.2.2.

MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION in our March 21, 1978 amendment letter and our current Technical Specifications with the attached revised page 82b.

The bases has been revised to compliment the new curves.

I The reactor vessel pressure temperature operating limits provided in the attached curves (Figures 3.2.2.a, 3.2.2.b, 3.2.2.c and 3.2.2.d) are for protection against non-ductile failure during:

(1) hydrostatic testing, (2) heat-up/cooldown with the core not critical and (3) heat-up/cooldown with the core critical.

The curves are applicable for up to ten effective full power years of operation.

Currently, Nine Mile Point Unit 1

has completed approximately 8.2 effective full power years of operation.

The curves represent the operating limits on the reactor vessel core beltline region and the vessel flange.

For normal operation, the core beltline region of the vessel is more restrictive from a pressure/temperature standpoint than the vessel flange at;

1) very low temperatures (i.e. below about 140'F when the core is critical and below about 100'F when it is not) and at 2) elevated temperatures (above 200'F when the core is critical and above 160'F when it is not).

The vessel flange region is controlling at intermediate temperatures.

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The fluence for ten effective full power years was calculated to be 7.73 x 1017 nvt

(

1 MeV) on the vessel inside surface.

This was based on Nine Mile Point Unit 1 flux wire data.

This includes

'a 30 percent uncertainty specified by General Electric in the report of, the flux wire measurements.

Fluence at the quarter wall thickness (1/4-T) location was estimated to be 75 percent of the fluence on the inside surface, as shown in General Electric Report NED0-21780, "Licensing Topical Report - Radiation Effects in Boiling Water Reactor Pressure Vessel Steels,"

dated October 1977.

In accordance with Nuclear Regulatory Commission Standard Review Plan, NUREG-75/087 dated November 24, 1975, the 1/4-T fluence was used for determining operating limits.

The analytical methods used for developing the revised curves were:

(1)

ASME Boiler and Pressure Vessel

Code,Section III, Appendix G,

1980 Edition with Winter 1982 Addenda, (2) Neutron Fluence extrapolated to the 1/4-T location from the inner wall by the method recommended by General Electric in General Electric Report NED0-21708, "Licensing Topical Report - Radiation Effects in Boiling Water Reactor Pressure Vessel Steels,"

October 1977 and (3)

Irradiation damage curves for 0.30 percent copper and worst case phosphorus content as outlined in NRC Regulatory Guide 1.99, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," April 1977.

In order to conservatively estimate vessel metal temperature at the 1/4-T location, reactor vessel outside diameter metal temperature is monitored during heat-up, while reactor coolant temperature is measured for cooldown.

However, in neither case will the rate of change of the coolant temperature exceed 100F/hour.

Measurement of coolant temperature and temperature rate in this manner will ensure compliance with the operating limits.

Niagara Mohawk plans to perform specimen testing (Charpy, tensile and flux wire scanning) on the vessel coupons and flux wires that were withdrawn during the recent reactor recirculation piping replacement outage.

After evaluating the results of the tests, a new set of pressure versus temperature operating limit curves extending beyond ten effective full power years will be prepared.

The operating limit curves will be based on NRC Regulatory Guide 1.99, Revision 1, dated April 1977 and the latest revisions of 10CFR50 Appendices G and H.

Based on these new curves, proposed Technical Specifications will be submitted.

The information contained in this submittal has been reviewed and approved by the Site Operations Review Committee and the Safety Review and Audit Board.

Sincerely,

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C.

V. Manghh Vice President Nuclear Engineering and Licensing CVM/MTG:djm Attachments

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION

~AH bi1i Applies to the minimum vessel temperature required for vessel pressurization.

Objective:

Applies to the required vessel temperature for pressurization.

Objective:

To assure that no substantial pressure is imposed on the reactor vessel unless its temperature is considerably above its Nil Ductility Transiti e Temperature (NDTT).

~Sinai a.

During reactor vessel heat-up and cooldown when the reactor is not critical the reactor vessel temperature and pressure shall satisfy the requirements of Figure 3.2.2. a.

b.

During reactor vessel heat-up and cooldown when the reactor is critical the reactor vessel temperature and pressure shall satisfy the requirements of Figure 3.2.2.b, except when performing low power physics testing with the vessel head removed at power 1 evel s not to exceed 5 mw(t).

To assure that the vessel is not subjected to any substantial pressure unless its temperature is greater than its NDTT.

Specificati on:

a.

Reactor vessel temperature and pressure shall be monitored and controlled to assure that the pressure and temperature limits are met.

b.

Vessel material surveill ance samples located within the core region to permit periodic mcnitoring of exposure and material properties shall be inspected on the follming schedule:

First capsule -

cne fcurth service life Second capsule - three fourth service life In the event the surveillance specimens at one quarter of the vessels service life indicate a shift of reference temperature greater than predicted the schedule shall be revised as follows:

Second capsule - one half service life

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT c.

During hydrostatic testing the reactor vessel temperature and pressure shall satisfy the require-ments of Figure 3.2.2.c if the core is not critical and Figure 3.2.2.d if the core is criti ca 1.

d.

The reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head are are equal to or greater than 100F.

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(HEATING"OR COOLING RATE < IOOF/HR)

FOR UP TO TEN EFFECTIVE FULL POWER YEARS OF CORE OPERATION

LIMIT FOR NON-CRITICAL OPERATION INCLUDING HEAT-UP/COOLDOWN AT UP TO 100F/HR PRESSURE-( si

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387 387 712 762 812 862 912 962 1012 1062 1112 1162 1212 1312 1412 TEMPERATURE.

F 100 100-160 160 166 172 177 182 187 192 196 199 203 207 213 219 TABLE 3.2.2.a MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEAT-UP OR COOLDOWN (REACTOR NOT CRITICAL)

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MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTING (REACTOR NOT'RITICAL)

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LIMIT FOR IN-SERVICE TEST (CORE NOT CRITICAL, FUEL IN VESSEL)

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F IGURE 3.2.2.d 0'/lINIIVIUM TEMPE RATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTI NG. (REACTOR CRITICAL' FOR UP TO TEN 'FFECTIVE FU-LL POWER YEARS OF CORE OPERATION

LIMIT FOR IN-SERVICE TESTS (CORE CRITICAL)

PRESSURE si 387 1278 1312 1412 TEMPERATURE F

156 170 173 181 TABLE 3.2.2.d MINIMUM TEMPERATURE FOR PRESSURIZATION OURING HYDROSTATIC TESTING (REACTOR CRITICAL)

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BASES FOR 3.2.2 AND 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION Figures 3.2.2.a and 3.2.2.b are plots of pressure versus temperature for a heat-up and cool down rate of 100F/hr.

maximum.

(Specificatim 3.2.1).

Figures 3.2.2.c and 3.2.2.d are plots of pressure versus temperature fee hydrostatic testing.

These curves are based on calculations of stress intensity factors according to Appendix G of Section III of the ASME Bo 11 er and Press ure Vessel Code 1980 Editi n with Hinter 1982 Addenda.

In additi e, temperature shifts due to integrated neutron flux at ten effective full power years of operation were incorporated into the figures.

These shifts were calculated from the fo.mula presented in Regulatory Guide 1.99, Revisicn 1 and the copper/phosphorus content of the reactor vessel.

These curves are applicable to the beltline region at low and elevated temperatures and the vessel flange at

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Reactor vessel flange/reactor head flange boltup is governed by other criteria as stated in cification 3.2.2.d.

The pressure readings on the figures have been adjusted to reflect the calculated elevation head ference between the pressure sensing instrument locaticns and the pressure sensitive area of the core beltline region.

The reactor vessel head flange and vessel flange in combination with the double "0" ring type seal are designed to provide a leak-tight seal when bolted together.

When the vessel head is placed on the reactor vessel, only that porticn of the head flange near the inside of the vessel rests on the vessel flange.

As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the "0" rings of the head and vessel flange.

Both the head and vessel and flange have a NDT temperature of 40F and they are not subject to any appreciable neutron radiatim exposure.

Therefcre, the minimum vessel head and head flange temperature for bolting the head flange and vessel flange is established as 40 + 60F or 100F.

Figures 3.2.2.a, 3.2.2.b, 3.2.2.c and 3.2.2.d have incorporated a temperature shift due to the calculated integrated neutron flux.

The integrated neutron flux at the vessel wall is calculated frcm core physics data and has been measured using flux monitors installed inside the vessel.

The curves are applicable fez up to ten effective full power years of operation.

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sel material surveillance samples are located within the core regicn to permit periodic monitoring of exposure and erial properties relative to control samples.

The material sample program conforms with ASTM E 185-66 except for the mater ial withdrawal scheduled which is specified in Specificatin 4.2.2.b.

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EEEAQHVM~PXRFMX Lukens Steel Compan Coatesville, PA 19320 June 30, 1983 Mr. M. Mosier Niagara Mohawk Power Corporation 300 Erie Boulevard, West

Syracuse, NY 13202

Dear Sir:

Attached are copies of the Lukens Steel Company "Chemical Laboratory Heat Analysis" records.

All,non-applicable information has been deleted from these copies.

We hope this will satisfy your request.

Sincerely, LUKENS STEEL COMPANY JPS:wlk Attachments John P.

Sunup ian Supt.,

Q.A. Inspection

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