ML19319C862

Tech Spec Change Requests 12a,12b,12c & 12d Re Tech Specs 1.1-1.7,2.1-2.2,3.1,3.4. & 4.4,deleting All Ref to Operation W/Less than Three Reactor Coolant Pumps in Operation & Deleting Ref to Monitoring Condenser Water Boxes

ML19319C862
Person / Time
Site:	Crystal River
Issue date:	11/08/1977
From:	FLORIDA POWER CORP.
To:
Shared Package
ML19319C856	List: ML19319C855 ML19319C862
References
NUDOCS 8003040821
Download: ML19319C862 (24)

Text

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l Technical Specification Change Request No. 12a Replace pages 2-1, 3, 5, 7, 82-2, 5, 7, 3/4 4-1, and B3/4 4-1 with the attached revised pages 2-1, 3, 5, 7, B2-2, 5, 7, 3/4 4-1 and B3/4 4-1.

Proposed Change All reference to operation with less than three reactor coolant pumps in operation have been deleted from the Technical Specifications.

Reason for Prooosed Change l Amendment 1 to Facility Operating License No. OPR-72, Paragraph 2.c.(3) I states: " Florida Power Corporation shall not operate the reactor in Operational Modes 1 and 2 with less than three reactor coolant pumps in operation until safety analyses for less than three pump operation have been submitted by the licensees and approval has been granted by the Commission by amendment to this license." A safety analysis has not been submitted and the removal of the references to less than three pump operation will make it clearer that operation in that mode is not allowed.

Safety Analysis of the Proposed Change No safety analysis of the proposed change is required since two-pump operation was not allowed by the Facility Operating License because a safety analysis for two-pump operation had not been submitted. By deleting the references to less than three-pump operation from the l Technical . Specification, the situation resulting from the plant enduring a trip of two reactor coolant pumps is now covered by Specification 3.0.3.

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2.0 -SAFETY LIMITS AND LIMITING SAFETY SYSTEM St.tiINGS 2.l' SAFETY LI!1ITS REACTOR CORE 2.1.1 - .The combination of the reactor coolant core outlet pressure and outlet temperature shall not exceed the safety limit shown in Figure

, 2.1 -1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of reactor coolant core outlet pressure and outlet temperature has exceeded the safety limit,-

be in HOT STANDBY within one hour.

REACTOR CORE 2.1.2 The combination of reactor THERMAL POWER and AXIAL POWER IMBALANCE shall not exceed the safety limit shown in Figure 2.1-2 for the various combinations of three and four reactor coolant pump operation.

APPLICABILITY : MODE 1.

ACTION Whenever the point defined by the combination of Reactor Coolant System flow, AXIAL POWER If1 BALANCE and THERMAL POWER has exceeded the appropriate safety limit, be in HOT STANDBY within one hour.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The Reactor Coolant System pressure shall'not exceed 2750 psig, hPPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTIg:

MODES 1 and 2 - Whenever the Reactor Coolant System pressure has ex-ceeded 2750 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within one hour.

MODES 3, 4 - Whenever the Reactor Coolant System pressure has and 5 exceeded 2750 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

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- CRYSTAL RIVER - UNIT 3 2-3

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M r-REACTOR PROTECTION SYSTEM INSTRUMENTATI0tl TRIP SETPOINTS

' R' FUNCTI0flAL UNIT TRIP SETPOINT ALLOWABLE VALUEC M

1. Manual Reactor' Trip Not Applicable Not Applicable 7 . .

E 2. Nuclear Overpower < 105.5% of RATED TilERMAL POWER < 105.5% of RATED THERMAL POWER

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4. Nuclear Overpower Trip Setpoint not to Allowable Values not to exceed Based on RCS Flow and exceed the limit line of the limit line of Figure 2.2-1.

AXIAL POWER IMBLANCE (j) Figure 2.2-1.

5. RCS Pressure-Low II) t 1800 psig 1 1800 psig j

6. RCS Pressure-liigh $ 2355 psig 1 2355 psig

7. RCS Pressure-Variable Low (I) 1 (16.25out T F - 7838) psig t (16.25 T out F - 7838) psig

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2. Trip would occur when three reactor coolant pumps are operating if power is > 80% and reactor flow rate is 76.6%, or flow rate is 1 71.8% and power is 75%.

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CRYSTAL RIVER - UNIT 3. B 2-5

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CRYSTAL RIVER - UNIT 3 B.2-7

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3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.4.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation.

APPLICABILITY: As noted below, but excluding MODE 6.*

ACTION:

MODES 1 and 2:

With one reactor coolant pump not in operation, STARTUP and POWER OPERATION may be initiated and may proceed provided THERMAL POWER is restricted to less than 78% of RATED THERMAL POWER and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the setpoints for the following trips have been reduced to the values specified in Specification 2.2.1

, for operation with three reactor coolant pumps operating:

1. Nuclear Overpower See Special Test Exception 3.10.3.  ;

CRYSTAL RIVER - UNIT 3 3/4 4-1

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'3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. With one reactor coolant pump not in operation in one loop, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE, ensuring that the DNBR will be maintained above 1.30 at the maximum possible THERMAL POWER for the number of reactor coolant pumps in operation or the local quality at the~ point of minimum DNBR equal to 15%, whichever is more restrictive.

A single reactor coolant loop provides sufficient heat removal capability for removing the core decay heat while in HOT STANDBY: however, single failure considerations require placing-a DHR loop into operation in the shutdown cooling mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psig. Each safety valve is designed to relieve 317,973 lbs per hour of saturated steam at the valve's l

setpoint.

l The relief capacity of a single safety valve is adequate to relieve any l overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating DHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psig.

The combined relief capacity of all of these valves is greater than the 4

maximum surge rate resulting from any transient.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

i CRYSTAL RIVER - UNIT 3 8 3/4 4-1 l

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Technical Specification Change Rcqu st No. 12b (ETS)

Replace pages 1-1, 2-1, and 2-2 with the attached revised pages 1-1, 2-1, and 2-2.

Proposed Change Revise Specifications 1.4 and 2.1.1 to delete references to monitoring the condenser water boxes as the points of intake and discharge of circulating water of the unit for temperature requirements.

Reason for Procosed Change Specification 2.1.2 allows Crystal River Unit 3 to operate with a AT across the condenser of 17.50F. The intent of the specification is to limit the circulating water discharge to the discharge canal to a temperature = 17.50F above that of tne intake circulating water (Final Environmental Statement, page 3-5). Because instrumentation was in-stalled in the condenser water boxes, the specification was written to reflect this instrument location. Subsequent to commercial operation of the plant, it was discovered that the RTD in the "D" outlet water box was detecting a higher than average discharge temperature. The thermal profile is attached. This can result in an unnecessary derating of the unit. The instrumentation on all four water boxes was declared in-operable and local indication (per Specification 2.1.1) is being em-played to show compliance with the Limiting Condition for Operation.

In order to again use the computer to monitor the unit aT, the outlet RTD's must be moved downstream to a point where a representative dis-charge temperature can be obtained. The Specification then must also be changed to allow for the use of these more representative sample points via the computer.

Environmental Imoact of Procosed Change There will be no impact to the environment as a result of this proposed change as it is changing a monitoring requirement to more adequately monitor the unit AT; it is not changing the Limiting Conditions for Operation Specification for the unit AT.

Benefit-Cost Analysis of the Procosed Change There will be no environmental cost as a result of this proposed change as stated above. The benefit of this change is that the cnit will not be unnecessarily derated because of an unrepresentative unit AT.

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THERMAL PROFILE Conditions: Inlet. Temperature = 72.20F Outlet Temperature (per computer) = 84.2 F Unit at' 50% power Temperature measured each six inches 83.650F 83.88' 83.42 82.60 -

80.97 79.74 -

79.61 Temperature in the center of 179.81 discharge pipe where the flow is greatest 80.06 80.44 80.81 81.91 82.95 83.65 corresponds to computer point 83.91 6

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/'-'5 1.0 Definitions The following terms are defined for uniform interpretation of the Environmantal Technical Specifications for Crystal River Unit 3.

1.1 Frequency - Terms used to specify frequency are defined as follows:

One per shift - At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Daily - At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Weekly - At least once per 7 days.

Monthly - At least once per 31 days.

Quarterly - At least once per 92 days.

Seniannually - At least once per 6 months.

A maximum' allowable extension for each surveillance requirement shall not exceed 25% of the surveillance interval.

1.2 Gross (8,y) Analysis - Radioactivity measurements of gross beth or gross beta in conjunction with gross' gac=a as defined in Regulatory b Guide 1.21.

1.3 Point of Discharge (POD) - The intersection of the discharge canal and the original bulkhead line as shown on Figure 1.1-1.

1.4 AT Across the Condenser - The average te=perature difference between the inlet and outlet of Unit 3.

1.5 Unit 3 Mixing Zone - The enclosed area _of the discharge canal bounded by the eastern end of the canal and the cable chase from Units 1 and

'2 by crossing the canal.

1.6 Emergency Need For Power - Any event causing authorized Federal officials to require or request that the Florida Power Corporation supply electricity to points within or without the State or other emergencies declared by State, County, or Municipal authorities during which an uninterrupted supply of electric power is vital to public health and safety.

1.7 Abnor=al power Ooeration - The operation of Crystal River Unit 3 beyond these technical specifications due to the Emergency Need for Power, g .*

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. 2-1 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 THERMAL Objective (General)

To limit thermal stress to the aquatic ecosystem and control effluent cooling water temperature within prescribed limits l which are consistent with applicable Federal and State regu- l lations in order to minimize adverse thermal effects.

2.1.1 Maximum AT Across Condenser Objective To limit the maximum temperature rise across Unit 3 during normal operation at all power levels.

Soecification l The temperature rise across the unit shall not exceed 17.5 0F for a period of more than 3 consecutive hours or a maximum of 210F unless there is an emergency need for power as defined in Section 1.

Monitoring Recuirement The unit temperature rise shall be monitored by detectors (RTD's 0-200 + 10F) located in the inlet and outlet of Unit 3.

The detector signal will be monitored by the control room computer. The aT will be alarmed at 17.50F and at 21oF l

maximum. '

If the RTD's or computer are inoperative during power opera-tion above 80*.', the unit AT shall be determined every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

+ 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> utilizing local temperature indicators (30 - 130 +

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Bases l When Unit 3 is operated at design capacity, the intake temoera-ture should be elevated by a value AT of 17.50F. When any one shell of the two twin-shelled surface steam condensers is inoperative for maintenance or other reasons, the AT will rise. Each of the 4 condenser sections will require cleaning I every 4 weeks, due to the buildup of marine growth or debris in the pipes and condensers. During the extreme climatic conditions, especially during tropical storms, sea grass-is uprooted from the Gulf of Mexico, requiring temporary shutdown of a circulator to clean grass and other debris which has I

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. e 2-2 accumulated at the intake structure or inside the condenser water boxes. This will cause a temporary increase inLthe AT.

across the unit. Because of these conditions the AT of 17.50F may be exceeded for a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> period with 210F specified as a maximum limit. Monitoring by means of RTD's in the inlet and outlet of Unit 3 will provide reliable values of the AT across the unit.

2.1.2 Maximum Discharge Temperature Objective To limit the maximum temperature of the condenser cooling water discharged from the plant to the environment during normal operation.

Soecification The temperature of the condenser cooling water at the Point of Discharge shall not exceed 1030F for a period of more than 3 consecutive hours or a maximum of 1060F ualess there is an emergency need for power as defined in Section 1.

Monitoring Requirement The temperature at the point of ' discharge shall be monitored once per hour during the power operations of Unit 3. The temperature sensor system has a range of 30-1100F and an accuracy of + 1/20F. A channel check shall be performed once per month.

When the monitor is inoperative the temperature at the point.

of discharge shall be estimated using operating and physical data in conjunction with curves generated by an empirical t

analysis of the Crystal River discharge canal variables.

Bases The effluent temperature limits during normal operations have been established to assure that the affected area within the receivin; waters is minimized. Due to conditions as specified in Section 2.1.1 Bases, the condenser cooling water temperature of 1030F at the point of discharge may be exceedad for a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> period with 1060F specified as a maximum limit.

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' Technical Specification Request No. 12c l

Replace page 3-4 with the attached revised page 3-4. 1 Proposed Change . I Add as a last sentence to the Specification section of Specification 3.1.3: "This Specification shall not be applicable during the period of the intake canal modification." l Reason for the Proposed Change The intake canal at Crystal River Unit 3 is being modified to accept  !

larger coal barges, i.e., it is being dredged to a minimum depth of 20 feet below mean sea level. In addition, the south dike of the intake canal is being lengthened by approximately 1.5 miles. This activity is scheduled to begin in March,1978 and run until August,1979.

The relation of the amount of impingement during this period to a period when no intake canal work is being performed is uncertain at best.

Therefore any data taken during this period would be of questionable value and would in fact be detrimental to the study no matter which way it would weigh the result. For the same reason, the daily monitoring of the screen-wash racks should be suspended.

Because of this, the Specification should be changed so that only data taken in a normal operational mode (i.e, no intake canal work) is  !

included in the impingement study. -

Environmental Imoact of the Prooosed Chance There will be no environmental impact because of this proposed change.

This change alleviates the requirement to sort the fish and shellfish normally impinged on the trash racks. The study will continue after the intake canal modification if the staff has not given approval to ter-minate it.

Benefit-Cost Analysis of the Procosed Chance As was stated above, there will be no environmental cost as a result of the proposed change. The benefits are twofold. First the impingement study will not be influenced by questionable data taken during this period. Second, the' cost to the licensee of approximately $120,000.00 will not be expended to collect this unusable data.

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, 4 Bases The metabolism of the marsh grass is expected to increase with

- increasing temperature. Any decrease indicates a breakdown of structure. If any of these parameters changes beyond 2o (two standard deviation) of that measured during preopera-tional monitoring, the system should be investigated for catastrophic results.

3.1.3 Impingement on' Intake Screens Objective To determine the quantity of impinged fish and shellfish on

~t he intake screens to compare with preoperational data.

Coecification The fish and shellfish collected in the trash racks adjacent to the intake screens of Units 1 and 2 and Unit 3 will be sampled for 24 consecutive hours once weekly. This program shall be conducted for one year after aperation of Unit 3 begins. This program may be terminated after one year period with staff's approval. Samples shall be sorted according to species, length, and wet weight. The screen-wash racks shall be monitored visually daily to determine any abnormal cathces.

This specification shall not be applicable during the period of the intake canal modification.

Recorting Reauirement Results of the data gathered in this program element shall be reported in accordance with Section 5.6.1. Any daily sample with fish and shellfish biomass greater than 50 kg shall be reported as specified in Section 5.6.2.

Bases preopt. rational data indicate that the average normal expected catch is approximately 20 kg per day for Unit No. 3.

3.1.4 General Ecological Survey Objective To detect changes which might occur and would be used to indicate areas requiring more detailed investigation.

Specifications A series of measurements shall be carried out during the operational life of the plant to indicate the general con-dition of the environment. The areas to be monitored are:

. c. c e '

Technical Specification Change Request No.12d-l (Appendix A)

. Replace pages 3/4 4-32 and B3/4 4-13 with the attached revised pages 3/4 4-32 abd B 3/4 4-13.

Proposed Change Change. Specification 4.4.10.1.b to read: "Each internals vent valve shall be demonstrated OPERABLE at .least once per 24 months during shutdown, by:". ,

Reason for Proposed Change Equilibrium fuel cycle refueling outages are generally'twelva to eighteen months apart. First and sometimes second fuel cycles run longer because of the greater reactivity of the core and the initial problems with a new plant causing increased down time. The performance of this surveillance requires that the reactor head be removed which, at any time other than refueling, is very undesirable. Besides the additional exposure to station personnel the lost generation in order to perform this one surveillance is extremely costly to the licensee.

The original surveillance of the eight internals vent valves was per-formed on November 20, 1976 with the present 18 month surveillance interval (plus 25%) extending to October 5,1978. This does not even encompass the first planned refueling outage of November 1, 1978. This change would allow the plant to continue operation and perform the internals vent valve surveillance during the first refueling outage.

Safety Analysis Justifying the Proposed Change The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains suffi-ciently covered. Inspection and manual actuation of the internals vent valves 1) ensures OPERABILITY, 2) insure that the valves are not stuck open during normal operation, and 3) demonstrates that the valves are fully open at the forces equivalent to the differential pressures assumed in the safety analysis.

Even if an internals vent valve was stuck open during normal operation, it would be in its post-accident position and would perform its function.

In addition, for an internals vent valve to be hampered in its operation, at least four of eight loose rotational clearances must either bind or freeze solidly.

Because of the simple design and operation of the internals vent valves, the successful original surveillance, and the fact that no internals vent valve has ever been found inoperable, the extension of the sur-veillance interval from 18 to 24 months is reasonable and does not endanger the health and safety of plant personnel or the public.

4

n .e s

' REACTOR COOLANT SYSTEM,.

SURVEILLANCE REQUIREME'!TS (Continued)

b. 'Each internals vent valve shall be demonstrated OPERABLE at least once per 24 months during shutdown, by:

1. Verifying through visual inspection .that the valve body and valve disc exhibit no abnormal degradation,

2. Verifying the valve is not stuck in an open pgsition,

.and

3. Verifying through manual actuation that the valve is fully open with a force of 1425 lbs (applied vertica: 1y upward).

4 l

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CRYSTAL RIVER - UNIT 3 3/4 4-32 F

- ~ , , . . . . - -

REACTOR COOLANT SYSTEM (Continued)

BASES

- All pressure-temperature limit curves are applicable up to the fifth 4

effective full power year. The protection against non-ductile failure is >

assured by maintaining the coolant pressure below the upper _ limits of Figures 3.4-2, 3.4-3, and 3.4-4.

The pressure and temperature-limits shown on Figures 3.4-2 and 3.4-3 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 10 CFR 50.

1 The number of reactor vessel irradiation surveillance specimens and.

the frequencies for removing and testing these specimens are provided in  :

Table 4.4-3 to assure compliance with the requrements of Appendix H to 10 CFR Part 50.

T

The limitations imposed on pressurizer heatup and cooldown and. spray 4

water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

3/4.4.10 STRUCTURAL INTEGRITY l

The inspection programs for ASME Code Class 1, 2 and 3 components,  :

except steam generator tubas, ensure that the structural integrity of  !

these components will be maintained at an acceptable level throughout j the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

The internals vent valves are provided to relieve the pressure i

generated by steaming in the core following a LOCA so that the core remains sufficiently covered. Inspection and manual actuation of the internals vent valves 1) ensure OPERABILITY, 2) ensure that the valves are not stuck open during normal operation, and 3) demonstrates that the valves are fully open at the forces assumed in the safety

analysis.

l

'l l CRYSTAL RIVER - UNIT.3 B3/4 4-13 l

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r. .

Technical Specification Change R:Quost No.12d-2 ' (A$pendix A) .

- Proposed Change Revise Specification 4.4.10.1.b.3 to read: _" Verifying e.hrough manual

actuation that the valve is fully open with a force of 1 425 lbs.

(appliedverticallyupward)."

d Revise the final sentence of Bases Specification 3/4.4.10 to read:

" Inspection and manual actuation of the internals vent valves 1) ensure OPERABILITY,.2) ensure that the valves are not stuck open during normal operation, and 3) demonstrates that the valve' s are fully open at the forces assumed in the safety analysis."

Reason for Proposed Change This change is being proposed to better define the force needed to fully open the internals vent valves and to delete the requirement to measure the opening force of the valves. This opening force surveillance dele-tion is explained below.

Safety Analysis Justifying Proposed Change i In the event of a LOCA in the reactor inlet piping, the internals vent valves (FSAR Figure 3-58) nave been designed to provide a direct steam relief path from the upper plenum to the break. During the reflooding phase of a LOCA transient, steam generated from ECCS fluid coming into contact with the hot cladding will pressurize the upper plenum, thereby opening the vent valves and allowing steam to flow through them. The_

rate of steam relief is calculated using a conservatively high flow resistance factor (k-factor) of 4.2 as is discussed in Topical Report BAW-10104. Demonstration that the vent valves will be full-open when a pressure differential is applied across them that is lower than that obtained during the LOCA analysis is sufficient to verify that the vent

valves are operational. Large Break LOCA analys4 for B&W 177 FA plants demonstrating the compliance of the ECCS system to 10 CFR50.46 has been j reported in Topical Report BAW-10103. Therefore, the required pressure differential necessary to hold the vent valves full-open must be shown i to be no more than 1.0 psid.

Attached find the force equivalent calculation for a 1.0 psid driving force necessary to hold the vent valves full-open (Attachment A). ' As can be seen by this calculation, the value specified in the proposed Technical Specification ensures that the valves will be open during the reflooding phase. Also attached, for informational purposes only, find

the calculation for derivation of the force equivalent needed for the vent valves to start to open (Attachment B). As BAW-10103 takes credit for the valves in the full-open position only and by virtue of the fact a lesser force equivalent is' required for start-to-open, the requirement for testing of the valves " start-to-open" force is deemed to be unnecessary.

A f

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The report was also reviewed to examine the effect on peak clad tem:: era-ture. It was determined that, with the exception of the first 0.2 seconds of the reflooding phase of the accident, a pressure differential in excess of 1.0 psid was maintained across the vent v11ves. Therefore, the vent valves would remain fully open during reflood.r'g, except for the first 0.2 seconds, even if the proposed 1.0 psid fo, ce was required.

For the analysis in BAW-10103, the ven' valves were asstmed always to be fully open. Thus, the proposed change 1:ild only have a 0.2 second effect. Even if it is assumed that the cladding would undergo an adiabatic heatup, over the 0.2 second interval, peak cladding temperature would increase by only 70F. In actuality the 0.2 second interval would be decreased by the rapid increase in upper plenum pressure that would occur with the vent valves closed. Additionally, over the 0.2 second delay, the heatup would not be adiabatic. The expected increase in cladding temperature would be.only 30F. Thus, the cladding temperatures reported in BAW-10103 are essentially unaffected by the proposed change.

. , , 3-  ;

- ATTACHMENT A

Purpose:

Calculate equivalent vent valve exercise force to " Hold-0 pen" pressure of 1.0 psi.

Analysis:

The valve face makes an angle of 26* with the vertical in the open position. A one (1) psi pressure drop acting over the 15 inch exposed i disc face has a 9-3/8 inch moment arm with the hinge.

Moment = Pressure Drop X Area X Moment Arm

= 1.0 psi X n (15 in.)2 X 9.375 in.

1 T

= 1656.7 in-lbs.

l Since the exercising force is a vertical force applied to the exercising pin, its moment arm is the projected horizontal distance of its radial

,' location with respect to the valve hinge.

Pin Radial Location is ( (4)2 + (5/8)2)1/2 = 4.048 in.

The angle the pin makes with the horizontal when the valve is open is the angle the valve makes with the vertical (26*)'minus the angle of the pin from horizontal when the valve is vertical (9):

Angle = 26* - Tan-l S/8 4

= 17.1

• Moment of the exercising force:

M force = Force X Pin Radial Location X Cosine Angle

= F X 4.048 X cos 17.1

= 3.87F

! Setting the force moment equal to the pressure moment, the equivalent

vent valve exercising force is solved

3.87F = 1656.7 F = 428.2 lbs.

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ATTACHMENT B

Purpose:

Calculate equivalent vent valve exercising force to " Start-To-Open" pressure.

Analysis:

To simplify parameterizations, the calculations will be based on 1.0 psi driving pressure. For any other " start-to-open" pressure, the equivalent exercise force will be in direct proportion; i.e., for a 1/2 psi differ-ential pressure, the exercise force is 1/2 the calculated value given below.

In the " closed" position the valve hangs 5' from the vertical. The moment arm of the exercising pin when the valve is closed is:

L=((4)2 X (5/8)2)l/2 cos (Tan-l 5/8 -5*)

4

= 4.039 in.

The pressure mcment of the pressure acting over the 14 inch face of a closed valve is:

M = 1.0 X 1(14)2 X 9.375 4

= 14412 in-lbs.

Setting the force and pressure moments equal:

4.039F = 1443.2 F = 357.3 lbs.

. . .