ML20206P133

Proposed Tech Specs,Revising pressure-temp Limit Curves Contained in Figures 3.6.1,3.6.2,3.6.3 & 3.6.4,deleting Completed RPV Sample SRs & Requirement to Withdraw Specimen at Next Refueling Outage & Removing Redundant SR for SLCS

ML20206P133
Person / Time
Site:	Monticello
Issue date:	12/31/1998
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20206P125	List: ML20206P122 ML20206P127 ML20206P133
References
NUDOCS 9901080038
Download: ML20206P133 (18)

Text

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4 Exhibit B MONTICELLO NUCLEAR GENERATING PLANT License Amendment Request Dated December 31,1998 Proposed Changes Marked Up on Existing Technical Specification Pages

3. ===============================================_=__=_______________

Page v

94 122 133 134 135

~136 146 L

~3 B-1 990100003e 981231 PDR ADOCK 05000263 ,

p- PDR a

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i LIST OF FIGURES

( Fiour's No.- Page 3.4.1 Sodium Pentaborate Solution Volume-Concentration Requirements 97 1

3.4.2 Sodium Pentaborate Solution Temperature Requirements 98 I

3.6.1 Ccre Beltline Operating Limits Curve Adjustment l vs. Fluence 133 l 3.6.2 Minimum Temperature vs. Pressure for Pressure Tests 134 3.6.3 Minimum Temperature vs. Pressure for .

p or l Cooldown Without the Core Critical ri ried 135 l 3.6.4 Minimum Temperature vs. Pressure for CoreT)psikt on 136 4.6.2 Chloride Stress Corrosion Test Results @ 500*F 137 3.7.1 Differential Pressure Decay Between the Drywell and Wetwell 191

! 3.8.1 Monticello Nuclear Generating Plant Site Boundary for l

Uquid Effluents 198g 3.8.2 Monticello Nuclear Generating Plant Site Boundary for Gaseous l Effluents 198h l.

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v 12/24/98 Amendment No. 9,35,47,74,7-7,78,104

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

b. Explode one of two primer assemblies manufactured in the same batch to verify proper function. Then install, as a replacement, the second primer assembly in the explosion valve of the system tested for operation.

c. se in pressure r etwee .e-G50 ps{

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i 3.4/4.4 94 l:gl9; Amendment No. 56, W-

- . _ _ _ - - - - - - - _ - - - - ._---.--___--------_----__------------__-____---__---_-_______r_ _ _ _ _ _

, ~L 4.0 SURVEILLANCE REQUIREMENTS 3.0 LIMITING CONDITIONS FOR OPERATION i B. Reactor Vessel Temperature and Pressure B. Reactor Vessel Temperature and Pressure

1. During in-service hydrostatic or leak testing when  ;

1. During in-service hydrostatic or leak testing, the reactor vessel shell temperatures specified in the vessel pressure is above 312 psig, the following 4.6.B.1, except for the reactor vessel bottom head. temperatures shall be recorded at least every 15 ,

5 shall be at or above the temperatures shown on the minutes. i two curves of Figure 3.6.2, where the dashed curve, a. Reactor vessel shell adjacent to shell flange.

"RPV Core Beltline," is increased by the core '

b. Reactor vessel bottom head.

beltline temperature adjustment from Figure 3.6.1. Reactor vessel shell or coolant temperature c.

I The reactor vessel bottom head temperature shall representative of the minimum temperature of be at or above the temperatures shown on the solid the beltline region.

curve of Figure 3.6.2,"RPV Remote from Core Beltline," with no adjustment from Figure 3.6.1. 2. Test specimens representing the reactor vessel, ,

base weld, and weld heat affected zone metal shall  ;

2. During heatup by non-nuclear means (except with be installed in the reactor vessel adjacent to the the reactor vessel vented), cooldown following vessel wall at the core midplane level. The material ,

nuclear shutdown, or low level physics tests the sample program shall conform to ASTM E 185-66. m,_

., 4 ., n u , a m a ,. - m ,,, -, ,_. m ,,,a reactor vessel shell and fluid temperatures specified e ., -

in 4.6.A shall be at or above the higher of the {,Q.@d M;{ (( Qy yQ temperatures of Figure 3.6.3 where the dashed clude a quant ~ ative determi ation of the ateria!  !

I curve, "RPV Core Beltline," is increased by the expected shift in RTNornom Hgure 3.6.1.

ch mistries. (No : Analysis of e first sam e has been ompleted. e Figure 3.6. core beltlin temper ture adjustm nt curve refle ts the chemi try i

3. During all operation with a critical reactor, other than data obtu ed).

for low level physics tests or at times when the reactor vessel is vented, the reactor vessel shell 3. Neutron flux 'res shall be stalled in th reactor and fluid temperatures specified in 4.6.A shall be at ssel adjacent the reacto vessel wall the or above the higher of the temperatures of Figure cor mid-plane le 1. The wire shall be re ved  !

3.6.4 where the dashed curve,"RPV Core Beltline," andt ted during th first refueli outage to is increased by the expected shift in RTraor from experim tally verify t calculate alue of neu on Figure 3.6.1. fluence at e fourth of e beltline s 11 thicknes ,

t t is used t determine t NDTT shi from Figur f

3.6.

3.6/4.6 122 11/TSC Amendment No. 3,72

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Bases 3.6/4.6 (Continued):

The fracture toughness of all ferritic steels gradually and uniformly decreases with exposure to fast neutrons above a threshold value, and it is prudent and conservative to account for this in the operation of the reactor pressure vessel. Two types of information are needed in this analysis: 1) A relationship between the changes in fracture toughness of the reactor pressure vessel steel and the l neutron fluence (integrated neutron flux), and 2) A measure of the neutron fluence at the point of interest in the reactor pressure  :

vessel wall.

The relationship of predicted adjustment of reference temperature versus fluence and the copper and nickel content of the core beltline materials given in Regulatory Guide 1.99, Revision 2, wa sed to define the core beltline temperature adjustment versus fluence shown on Figure 3.6.1- originally A relationship between fu 1 power years of operation and neutron fluence has been experimentally determined for the reactor vessel.!

The vessel pressurization temperatures at any time period can be determined from the thermal energy output of the plant and Figure l with a. 3.6.1 used in conjunction with Figure 3.6.2 (pressure tests), Figure 3.6.3 (mechanical heatup or cooldown fc"odng nuiar j

}"g"e)ntical > shaMeea), or Figure 3E Jdosimeter

~?utron (operation wires with which a critical were installed core). adjacent During to thethe first vessel wallfuel werecycle, only removed calculated neutr:

to experimentally used. At the first refu' determine the neutron , nse versus full power years of operation. This experimental result was updated by testing additional dosimetry removed wn.. the first surveillance capsule.

Reactor vessel material samples are provided, however, to verify the relationship expressed by Figure 3.6.1. Three sets of l mechanical test specimens representing the base metal, weld metal, and weld heat affected zone (HAZ) metal have been placed in j

the vessel and carf be removed and tested as required An analysis and report will be submitted to the Commission on all such j

surveillance specimens removed'from the reactor vesse in accordance with 10 CFR 50, Appendix H, including information obtained on the level of integrated fast neutron irradiation receive by the specimens and actual vessel material. t Two sets of specimens were contained in the first surveillance capsule M which was removed from the vesselin 198L One set of specimens was f i

tested at this time. The second set was later inserted into a new capsule, and installed in the Prairie Isfand RPV for continued irradiation.

This capsule was removed and tested in 1996. NSP performed calculations per the requirements of Regulatory Guide 1.99, Rev. 2, Position 2.1 to develop new pressure / temperature (P-T) curves. Results of Charpy V-notch impact tests for the two sets of data d were used in developing

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the revised Figures 3.6.1,3.6.2,3.6.3, and 3.6.4. - ,

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146 4/3C/30 l 3.6/4.6 BASES Amendment No. 72,100a j

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'I Exhibit C

' MONTICELLO NUCLEAR GENERATING PLANT License Amendment Request Dated December 31,1998 Revised Monticello Technical Specification Pages

=======================================================

Page v

94 122 133 134 135 136 146 l

1 I

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LIST OF FIGURES Fiaure No. Page 3.4.1 Sodium Pentaborate Solution Volume-Concentre. lion Requirements 97 3.4.2 Sodium Pentaborato Solution Temperature Requirements 98 3.6.1 Core Beltline Operating Limits Curve Adjustment vs. Fluence 133 3.6.2 Minimum Temperature vs. Pressure for Pressure Tests 134 i l 3.6.3 Minimum Temperature vs. Pressure for Mechanical Heatup or i Cooldown Without the Core Critical 135 i

3.6.4 Minimum Temperature vs. Pressure for Critical Core Operation 136 l l 4.6.2 Chloride Stress Corrosion Test Results @ 500*F 137 3.7.1 Differential Pressure Decay Between the Drywell and Wetwell 191 3.8.1 Monticello Nuclear Generating Plant Site Boundary for Liquid Effluents 198g 3.8.2 Monticello Nuclear Generating Plant Site Boundary for Gaseous l Effluents 198h l

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Amendment No. 9,36,47.,74,77,79,404 l

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4.0 SURVEILLANCE REQUIREMENTS 3.0 LIMITING CONDITIONS FOR OPERATION

b. Explode one of two primer assemblies manufactured in the same batch to verify proper function. Then install, as a replacement, ,

the second primer assembly in the explosion valve of the system tested for operation.

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Amendment No. 56,77 .

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS Reactor Vessel Temperature and Pressure . B. Reactor Vessel Temperature and Pressure ,

- B.

During in-service hydrostatic or leak testing, the 1. During in-service hydrostatic or leak testing when 1.

reactor vessel shell temperatures specified in the vessel pressure is above 312 psig, the following 1

4.6.B.1, except for the reactor vessel bottom head, temperatures shall be recorded at least every 15 shall be at or above the temperatures shown on the minutes.

two curves of Figure 3.6.2, where the dashed curve, a. Reactor vessel shell adjacent to shell flange.

"RPV Core Beltline,"is increased by the core

b. Reactor vessel bottom head.

beltline temperature adjustment from Figure 3.6.1. Reactor vessel shell or coolant temperature The reactor vessel bottom head temperature shall c.

representative of the minimum temperature of be at or above the temperatures shown on the solid the beltline region.

curve of Figure 3.6.2,"RPV Remote from Core Beltline," with no adjustment from Figure 3.6.1' 2. Test specimens representing the reactor vessel, base weld, and weld heat affected zone metal shall

2. During heatup by non-nuclear means (except with be installed in the reactor vessel adjacent to the the reactor vessel vented), cooldown following vessel wall at the core midplane level. The material '

nuclear shutdown, or low level physics tests the sample progrcm shall conform to ASTM E 185-66. t reactor vessel shell and fluid temperatures specified l in 4.6.A shall be at or above the higher of the temperatures of Figure 3.6.3 where the dashed curve, "RPV Core Beltline," is increased by the expected shift in RTNOTf rom Figure 3.6.1.

3. During all operation with a critical reactor, other than for low level physics tests or at times when the reactor vessel is vented, the reactor vessel shell and fluid temperatures specified in 4.6.A shall be at or above the higher of the temperatures of Figure 3.6.4 where the dashed curve,"RPV Core Beltline," [

is increased by the expected shift in RTNDTf rom Figure 3.6.1.

t 3.6/4.6 122 Amendment No. 3,73 ,

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L Bases 3.6/4.6 (Continued):

The fracture toughness of all ferritic steels gradually and uniformly decreases with exposure to fast neutrons above a threshold value, and it is prudent and conservative to account for this in the operation of the reactor pressure vessel. Two types of information

are needed in this analysis: 1) A relationship between the changes in fracture toughness of the reactor pressure vessel steel and the neutron fluence (integrated neutron flux), and 2) A measure of the neutron fluence at the point of interest in the reactor pressure vessel wall.

, The relationship of predicted adjustment of reference temperature versus fluence and the copper and nickel content of the core beltline materials given in Regulatory Guide 1.99, Revision 2, was originally used to define the core beltline temperature adjustment l versus fluence shown on Figure 3.6.1.

A relationship between full power years of operation and neutron fluence has been experimentally determined for the reactor vessel.

The vessel pressurization temperatures at any time period can be determined from the thermal energy output of the plant and Figure 3.6.1 used in conjunction with Figure 3.6.2 (pressure tests), Figure 3.6.3 (mechanical heatup or cooldown with a noncritical core), or l Figure 3.6.4 (operation with a critical core). During the first fuel cycle, only calculated neutron fluence values were used. At the first refueling, neutron dosimeter wires which were installed adjacent to the vessel wall were removed to experimentally determine the

neutron fluence versus full power years of operation. This experimental result was updated by testing additional dosimetry removed with the first surveillance capsule.

Reactor vessel material samples are provided, however, to verify the relationship expressed by Figure 3.6.1. Three sets of mechanical test specimens representing the base metal, weld metal, and wald heat affected zone (HAZ) metal have been placed in the vessel and can be removed and tested as required. Two sets of specimens were contained in the first surveillance capsule which was removed from the vessel in 1981. One set of specimens was tested at this time. The second set was later inserted into a new capsule, and installed in the Prairie Island Nuclear Generating Plant RPV for accelerated irradiation. This capsule was removed and tested in 1996. NSP performed calculations per the requirements of Regulatory Guide 1.99, Rev. 2, Position 2.1 to develop new pressure / temperature (P-T) curves. Results of Charpy V-notch impact tests for the two sets of data and from 1997 non-irradiated material test data were used in developing the revised Figures 3.6.1,3.6.2,3.6.3, and 3.6.4. An analysis and report will be submitted to the Commission on all such surveillance specimens removed from the reactor vessel in accordance with 10 CFR 50, Appendix H, including information obtained on the level of integrated fast neutron irradiation received by the specimens and actual vessel material.

146 3.6/4.6 BASES Amendment No. 72,400s

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