ML20114A922

License Amend Request 92-06 to License NPF-86 Implementing Guidance of GL 88-11 to Enhance Safe Operation of Plant by Reducing Time in Which Existing Heatup & Cooldown Curves Applicable

ML20114A922
Person / Time
Site:	Seabrook
Issue date:	08/18/1992
From:	Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
Shared Package
ML20114A915	List: ML20114A913 ML20114A922 ML20114A924
References
GL-88-11, NUDOCS 9208240108
Download: ML20114A922 (5)

Text

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M North suonoox granon umr 1 acmam At/ antic Enorgy Service Corporation i

i Facility Operatin License NPF 86 Docket

o. 50 443 License Amendment Request No. 92 06 Revised RCS Pressure / Temperature Limits This License Amendment Request is submitted by North Atlantic Energy Service Corporation pursuant to 10CFR50.90. The following information is enclosed in support of this License Amendment Request:

Introduction and Description of Proposed Changes

. Section 1

. Section ll Markup of Proposed Changes Section ill Retype of i'roposed ChangesSection IV Safety Evaluation of Proposed Changes Determination of Significant Hazards for Proposed Chan0esSection V Section VI Proposed Schedule for License Amendment Issuance ar'd Effectiveness

. Section Vil Environmental Impact Assessment l

Sworn and Subscribed to before ma this

[

I?O' day Of beu.[

,1992

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}f O 4W dams % (S $dldomu Tod C. F%enbaum Nobry Public A

Senior Vice President pnd Chief Nuclear Officer 9208240109 920817 l'DR ADOCK 0500 3

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1.

Introduction and Description of Proposed Chnnnes A,

Introduction j

The purpose of the proposed Technical Specification changes is to revise f

Technical Specification 3/4.4.9, ' Pressure / Temperature Limits,' and its associated Bases to address the recommendations of Generic Letter (GL) 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its-Impact on Plant' Operations.'

i 10CFR50, Appendix A.

General Design Criterion (GDC) 31.

" Fracture Prevention of Reactor Coolant Pressure Boundary,* requires in part, that the reactor coolant pressure boundary be designed with sufficient margin to ensure that. when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non-brittle manner and (2) the probability of a rapidly propagating fracture is minimized. GDC 31 also requires that the design reflect the uncertainties in determining the effect of irradiation on material properties. 10CFR$0 Appendix G.

" Fracture Toughness Requirements

• and 10CFR$0 Appendix H,

' Reactor Vessel Material Surveillance Program Requirements,' which implement, in part, Criterion 31, necessitate the calculation of changes in fracture toughness of reactor vessel materials caused - by neutron

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radiation throughout th6 service life.

Regulatory Guide 1.99, Revision 2 'Kadiation Embrittlement of Reactor Vessel Materisis,' describes general o.

procedures acceptable to the NRC staff for calculating the ef fects of neutron radiation embrittlement of low alloy steels currently used for light water. cooled reactor vessels.

On July'12, 1988'the NRC issued ' GL 88-11 to notify operating reactor license holders and construction permit holders of the issuance of Revision 2 to Regulatory Guide 1.99, ' Radiation Embrittlement of Reactor Vessel Materials".

The NRC recommended in GL 88-11 that licensees use Revision 2 to Regulatory Guide 1.99 to predict the effect-of neutron radiation on reactor vessel materials as required by Paragraph V of 10CFR50 Appendix G.

North Atlantic responded to GL 88-11 on November 30, 1988 (Ref.

NYN-88155 enclosed in Section VIII).

The North Atlantic' response to GL 88-11 stated that a revision to the Heatup and Cooldown Curves of Technical Specification 3/4.4.9, _' Pressure / Temperature Limits,' would be. required prior-to exceeding eleven Effective Full Power Years (EFPY) of operation.

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Additionally, the North Atlantic response committed to submit the Technical

, Specification' changes required by GL 88-11 prior to the second refueling outage.

Based on calculations performed for North Atlantic by the Yankee Atomic Electric Company Nuclear Services Division (YNSD) utilizing the methodology described-in Revision 2 to Regulatory Guide 1;99 it was concluded that the current-heatup and cooldown curves do not require revision.if the service period is revised to 11.1 EFPY versus the current 16 EFPY.

The YNSD

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calcul'ations are available-for review at Seabrook Station.

In addition, the Reference Temperature for Nil Ductility Transition (RTm) for the 1/4T l'

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and 3/4T reactor vessel locations (T is the thickness of the reactor vessel wall) have been revised slightly downward to 103*F and 86*F. respectively.

Over time, the affects of fast neutron radiation on the reactor vessel walls results in their gradual embrittlement.

This embrit' clement is y

directly proportional to neutron fluence, and inversely proportional to o

the distance, T, through the reactor vessel wall. The result of the fast neutron radiation is an increase in RTor over time.

The value of RTg7 used to determine the initial heatup and cooldown limit curves specified in Technical Specification 3/4.4.9, ' Pressure / Temperature Limits,' includes a correction factor, which accnunts for neutron radiation embrittlement.

As discussed in the proposed revisions to the Bases for Technical Specification 3/4.4.9 the reactor materials have been tested to determine

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their initial RTy37: the results of these tests are shown in Table L 3/4.4-1.

Reactor nperation and resultant fast neutron (E greater than 1 MeV) irradiation.:an cause an increase in the RTgp7 Therefore, an adjusted reference temperature, based upon the fi.uenc e, copper content, and nickel content of'the material in question, can be predicted using Figure B 3/4.4 1 and the value of ARTm7 computed utiliging Regulatory Guide - 1.99, Revision 2,

' Radiation Embrittlement of Reactor Vessel Materials.'

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in rte 7 at the end of 11.1 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments.

Values of AR'Inr determined in this manner may be used until the results from the material surveillance program, evaluated according to ASTM E185, are available.

Capsules will be removed in accordance with the

-requirements of ASTM E185-73 and 10CFR50, Appendix H.

The lead factor represents the relationship between the fast neutron flux density at the

'I location of the capsule and the inner wall of the reactor vessel.

Therefore, the results obtained f rom the surveillance specimens can be used to predict future radiation damage te the reactor vessel material by using the lead factor and the withdrawal time of the capsule.-

Evaluation of-surveillance capsule data vill be conducted in accordance with NRC Regulatory Guide 1.99 Revision 2.

B..

Description of Proposed Changes The proposed Technical-Specification changes have been developed based on and are consistent with - the methodology described in Revision 2 to

- Regulatory Guide 1.99.

The proposed changes to Seabrook Station Technical Specification 3/4.4.9, " Pressure / Temperature Limits," and its associated Bases are described below:

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Technical Specification 3/4.4.9 and Associated Bases ' Pressure /Temperatura Limits' i

1.

Figures 3.4-2, Reactor Coolant System lleatup Limitatiors and 3.4-3 Reactor Coolant Cooldown System Limitations are being revised to state that the curve is applicable for the service period up to 11.1 EFPY. Calcuir*. ions performed by YNSD verify that the existing heatup I

and cooldown curves do not require reviolon if the applicable service period is revised to 11.1 EFPY.

e 2.

Figures 3.4-2, Reactor Coolant System Heatun Limitations and 3.4-3 Reactor Coolant System Cooldown Limitations are being clarified by specifying that the Copper content of the controlling material is 0.06WTI.

This is the Copper content of the governing sample used to calculate the Reference Temperature for Nil Ductility Transition.

3.

Figures 3.4-2, Reactor Coolant System Heatup Limitations and 3.4-3, Reactor Coolant System Cooldown Limitations.

The Reference Temperature for Nil Ductility Transition af ter 11.1 EFPY, is revised

-to 108'F and 86*F for the 1/47 and 3/4T reactor vessel locations respectively.

The Reference Temperatures for Nil Ductility Transition were calculated based on the methodology described in Revision 2 to Regulatory Guide 1.99.

4.

Bases for Technical Specification 3/4.4.9. Revised to indicate that the heatup and cooldown curves are applicable for 11.1 EFPY and that the methodology in Revision 2 to Regulatory Guide 1.99 was used to calculate the Reference Temperature for Nil Ductility Transition.

5.

Bases Figure B 3/4.4-2, Effect of Fluence and Copper Content on Shift of RTy37-for Reactor Vessels Exposed to 550'F is deleted. The figure being deleted was included in the original Seabrook Station Technical Specifications as an alternative method to Regulatory Guide 1.99 for predicting the effects of radiation embrittlement of reactor vessel amterial and the associated shif t in RTgq. This alternative method is not used and its inclusion in Technical Specifications

-serves no purpose, i

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Markun of Proposed Channes See attached markup of proposed changes to Technical Specifications.

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