ML20080J671

Replacement Pages 111,117 & 118 to Proposed Tech Spec Figure 3.6.1 Re Reactor Vessel Pressure Temp Limits for Operation Through 1.15E8 Mwh(T)

ML20080J671
Person / Time
Site:	Vermont Yankee
Issue date:	02/07/1984
From:	VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20080J658	List: ML20080J657 ML20080J671
References
NUDOCS 8402150166
Download: ML20080J671 (3)

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8402150166 840207 PDR ADOCK 05000271 111 P PDR

VYNPS RASES -

3.6 & 4.6 REACTOR COOLANT SYSTEM f Pressure and Temperature Limitations i A'll components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system i temperature and pressure changen. These cyclic loads are introduced by normal load transients, reactor trips,

and startup and shutdown operations. The various categories of load cycles used for design purposes are l provided in Section 4.2 -of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions i and satisfy the stress limits for cyclic operation.

j. During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary f rom compressive at the inner wall to tensile at the outer wall. These thermal-induced compressive stresses tend to

, alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based j on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall-of the vessel becomes the controlling location. The thermal gradients established during heatup I produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure-induced j tensile stresses which are already present. The thermal-induced stresses at the outer wall of the vessel are j tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower

! bound curve similar to that described for the heatup of the inner wall cannot _ be defined. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures should be within 500F of each other prior to startup of an idle loop.

The reactor vessel materials have been tested to determine their initial nil-ductility transition temoetature 4 (NDTT) of 400F maximum. An additional margin of 200F has been added in order to estimate reference temperature, RTNDT. Rean.or operation and resultant fast neutron (E greater than 1 Mev) irradiation will cause an increase in-the ETNDT. Therefore, an adjusted reference temperature can be predicted using current industry practices (GE SIL No.14, Supplement No.1) based on recent GE ' surveillance data. The reference l

i Amendment No. 62 117 i

1 I

j i

i'

\

o taparttura fcr cirura flenga matcrici 10 d; fined by SRP Sectic 5.3.2, cnd Br nch Tcchnic21 Pritirn MTEB 5-2, .'

"Practure Toughness Requirements for Older Plants". The closure flange is located in a low neutron fluence area .

and therefore no measurable RTNDI shift is expected over the life of the plant. The pressure / temperature

• limit curve Figure 3.6.1 includes predicted adjustments for this shif t in RTNDT for operation through 1.330 x 108 M4H(t), as well as adjustments for possible errors in the pressure and temperature sensing.

instruments.

The actual shif t in NDTT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. In order to estimate the material properties at the 1/4 and 3/4 positions in the vessel plate, the shift in NDTT is assumed to be 62% and 22%, respectively of the irradiation samples properties. The heatup and cooldown curves must te recalculated when the RTNDT determined f rom the surveillance capsule is different f rom the calculated RTMDI for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.6.1 for reactor criticality sad for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided to assure compliance with the requirements of Appendix H to 10CFR Part 50.

Coolant Chemistry A steady state radioiodine concentration limit of 1.1 Ci of I-131 dose equivalent per gram of water in the reactor coolant system can be reached if the gross radioactivity in the gaseous effluents are near the limit as set forth in Specification 3.8.C.l.a or there is a failure or prolonged shutdown of the cleanup demineralizer.

In the event of a steam line rupture outside the drywell, the NRC staff calculations show the resultant radiological dose at the site boundary to be less than 30 rem to the thyroid.

6 Amendment No. 62 118

VEIIMONT YANKEE NUCLEAR POWER CORPORATION Proposed Change No. 118

. RD 5 Box 169. Ferry Road, Brattleboro, VT 05301 nEpty To y y ENGINEERING OFFICE 1671 WORCESTER ROAD FRAMINGHAM, MASSACHUSETTS 01701

{*

• TELE

• HONE 617-872-8100 February 7,1984 FVY 84-9 United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Office of Nuclear Reactor Regulation Mr. D. G. Eisenhut, Director Division of Licensing

References:

(a) License No. DPR-28 (Docket No. 50-271)

(b) Letter, VYNPC to USNRC, FVY 83-45, Proposed Change No.

107, dated May 26, 1983 (c) Final Rule [48FR24008), Fracture Toughness Requirements for Light Water Reactors, dated May 27, 1983

Subject:

Reactor Vessel Pressure Temperature Curves

Dear Sir:

Pursuant to Section 50.59 of the Commission's Rules and Regulations, Vermont Yankee Nuclear Power Corporation hereby proposes the following change to Appendix A of the Operating License.

Proposed Change Replace Pages 111, 117, and 118 of the Vermont Yankee Technical Specifications with the enclosed revised Pages 111, 117, and 118. These pages are intended to supersede the replacement pages previously submitted to you via Reference (b) .

Figure 3.6.1, " Reactor Vessel Pressure Temperature Limits for Operation Through 1.15E8 MWh(t)" has been updated to reflect allowable heatup curves for reactor operation through a power output of 1.330E8 MWh(t). The revised figure also reflects the promulgation of a revision to 10CFR Part 50, Appendix G [ Reference (c)]. Pages 117 and 118 have been revised to reflect a change to the bases section of the Technical Specifications, i

Jnited States Nuclear Regulatory Commission February 7,1984 Attention: Mr. D. G. Eisenhut, Director Page 2 Reason for Change This proposed change will revise our Technical Specifications to accommodate shif ts in transition temperature for the reactor vessel materials that were induced by radiation ef fects. These shif ts are accounted for by revision of our presaure-temperature limits for heating up and cooling down the reactor. Periodic review and adjustment, if necessary, of the curves to account for the effects of increased neutron exposure is required by 10CFR Part 50, Appendices G and H.

This change adjusts the curves of Figure 3.6.1 to compensate for the effects of increased neutron exposure to permit operation to a power level of 1.330E8 MWh(t). This adjustment is necessary because the existing curves are limited to a power output of 1.15E8 MWh(t), a value which is expected to be reached during March 1984.

Basis for Change The basis for this change is discussed in detail in Reference (b). In addition, the recent promulgation of a rule change to 10CFR Part 50, Appendix G [Ref erence (c)] allows for:

1. Removal (for BWRs) of the hydrostatic pressure test temperature limit for criticality. The new temperature limit for criticality is the RTNDfof the closure flange plus 600F. Fo r Ve rmont Yankee, this will be 120 F (the former temperature was 173.50F). This value is applicable at pressu res <220 psig. At pressures > 220, the criticality curves are a continuation of the previous curves based on 10CFR50, Appendix G, which requires that vessel temperature always be 400F above the ASME Code,Section III, Appendix G, calculated curves during criticality.

2. Limiting normal operation and hydrotest to pressures below 220 psig until vessel closure flange temperature is well above RTNDT of closure flange region. Specifically, when pressure exceeds 220 psig, the new hydrotest l temperature is RTNDT of the closure flange plus 900F. This temperature le 1500F. In addition, when pressure exceeds 220 psig, the new normal operation temperature is [RT NDTICF P l us 1200F = 1800F.

Ve rmont Yankee's vessel closure flange is ASME SA 508, C12 material.

Because no f racture toughness test data for this material is available, its reference temperature (RTNDT) is defined by " Standard Review Plan", Section 5.3.2, Pressure Temperature Limits, and Branch Technical Position MIEB 5-2,

" Fracture Toughness Requirements for Older Plants". That temperature is 600F.

4

4

~ United States Nuclear Regulatory Commission February 7,1984 Attention: Mr. D. G. Eisenhut, Director Page 3 I

The closure flange is located in a low neutron fluence area, i.e., out of the " vessel beltline", and therefore no measurable RTNDT shift is expected over plant lif e.

These changes are reflected on the enclosed Pages 111 (Figure 3.6.1),

117, and 118.

- Safety Considerations The safety considerations are discussed in detail in Reference (b). This change haca been reviewed by the Nuclear Safety Audit and Review Committee.

Significant Hazards Consideration The NRC haa provided guidance concerning the application of standards for conclusions regarding "Significant Hazards Consideration" [48FR14870]. The examples of actions involving no significant hazards consideration include:

"A change to make a license conform to changes in the regulations, where the license change results in very minor changes to f acility operations cicarly in keeping with the regulations."

This change to the pressure-temperature limits is similar to the example cited above because 10CFR Part 50, Appendices G and H require the updating of pressure-temperature limits based on the surveillance program. This proposed change will result in a minor change to facility operations clearly in keeping with the regulations.

Based on the above, we have determined that this change does not

constitute a significant hazards consideration, as defined in 10CFR50.92(c).

i Fee Detet1nination ,

c This proposed change requires an approval that involves a single safety issue and .is not deemed to involve an unreviewed safety question. For these-reasons, Vermont Yankee Nuclear Power Corporation proposes this change as a Class III Amendment. - A payment of.$4,000.00 is enclosed.

Schedule of Change.

For reasons discussed above, we request that you expedite your review and approval of this proposed change. This change will be implemented as soon as practicable following receipt of your approval.

l r

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_, . . , . . ...,,._m_,,__._,,_,,._..._- .... , _ . , , _ _ _ _ , , _ _ , _ , , , , _ . . , . , _ , , ,

0-United States Nuclear Regulatory Commission February 7,1984 Attention: Mr. D. G. Eisenhut, Director Pag e 4 i

We trust that this infomation is acceptable; however, should you require additional information, please contact us.

Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION hc Y L. H. Heider Vice President JBS/bal cc: Vermont Department of Public Services 120 State Street Montpelier, Vermont 05602 Attention: Mr. Richard Saudek, Chairman Enclosure COMMONWEALTH OF MASSACHUSETTS)

)ss MIDDLESEX 00UNTY )

Then personally appeared bef ore me, L. H. Heider, who, being duly sworn, did etate that he is a Vice President of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Vermont Yankee Nuclear Power Corporation and that the statements therein are true to the best of his knowledge and belief.

0 M U. Sinclair L. tkA Notary Public My Commission Expires June 1,1984 l' ).5'.'.:.k s ei; .

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VYNPS -

RASES -

3.6 & 4.6 REACTOR COOLANT SYSTEM Pressure and Temperature Limitations All cocponents in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature ar.d pressure changes. These cyclic loads are introduced by normal load transients, reactor trips,  !

and startup and shutdown operations.. The various categories of load cycles used for design purposes are L provided in Section 4.2 of the FSAR. During startup and shetdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions l and satisfy the stress limits for cyclic operation.

. During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary f rom ,

I compressive at the inner wall to tensile at the outer wall. These thermal-induced compressive stresses tend to l alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based ,

on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for 'inite heatup rates when the inner wall of the vessel is treated as the governing location.

j The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure-induced j tensile stresses which are already present. The thermal-induced stresses at the outer wall of the vessel are i j tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower l bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the

, cases in which the outer wall of the vessel becomes the strees controlling location, each heatup rate of j interest must be analyzed on an individual basis.

l In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures should be within 500F of each other prior to startup of an idle loop. i l The reactor vessel materials have been tested to determine their initial nil-ductility transition temperaturc .

j (NDTT) of 400F maximum. An additional margin of 200F has been added in order to estimate reference j temperature, RTNDT. Reactor operation and resultant fast neutron (E greater than 1 Mev) irradiation will i cause an increase in the RTNDT. Therefore, an adjusted reference temperature can be predicted using current

) industry practices (CE SIL No.14, Supplement No.1) based on recent GE surveillance data. The reference l

i Amendment No. 62 117

, I i

j temperature for clocure flanga material io dsfined by SRP Sectica 5.3.2, cad Brzoch Technical Pacitica MTEM 5-2h *

" Fracture Toughness Requirements for Older Plants". The closura flange is' located in a low neutron fluence area and therefore no measurable RTNDT shif t is expected over the life of the plant. The pressure / temperature . -

limit curve Figure 3.6.1 includes predicted adjustments for this shift in RTNDT for operation through 1.330 x 108 MWH(t), as well as adjustments for possible errors in the pressure and temperature sensing instruments.

l The actual shif t in NDTT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shif t for a ,

j sample can be applied with confidence to the adjacent section of the reactor vessel. In order to estimate the material properties at the 1/4 and 3/4 positions in the vessel plate, the shif t in NDTT is assumed to be 62% and l 22%, respectively of the irradiation samples properties. The heatup and cooldown curves must be recalculated l when the RTNDT determined from the surveillance capsule is different from the calculated RTNDT for the ,

equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.6.1 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided to assure compliance with the requirements of Apper. dix H to 10CFR Part 50. ,

Coolant Chemistry i

A steady state radioiodine concentration limit of 1.1 Ci of I-131 dose equivalent per gram of water in the i reactor coolant system can be reached if the gross radioactivity in Che gaseous effluents are near the limit as ,

I set forth in Specification 3.8.C.1.a or there is a failure or prolonged shutdown of the cleanup demineralizer. 1 In the event of a steam line rupture outside the drywell, the NRC staff calculations show the resultant l radiological dose at the site boundary to be less than 30 ren to the thyroid.

I l

l i l

1  !

? i

! Amendment No. 62 118

• i 1

l l \

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