ML18087A526
| ML18087A526 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/05/1982 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18087A525 | List: |
| References | |
| NUDOCS 8210190515 | |
| Download: ML18087A526 (16) | |
Text
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PROPOSED CHANGE SALEM UNIT NO. 1 TECHNICAL SPECIFICATIONS Description of Change e* Ref. LCR 80-05 The boron concentration and contained volume limits for the refueling water storage tank as set forth in Unit No. 1 Technical Specifications should be changed to agree with the corresponding limits in Unit No. 2 Technical Specifications.
Similarly, the Sodium Hydroxide (NaOH) concentration and con-tained volume limits for the containment spray additive tank as listed in Unit No. 1 Technical Specifications should be changed to match those delineated in Unit-No.* 2 Technical':"_,_,_.:....
Specifications.
The requested changes are necessary to satisfy the conditions of NRC IE Bulletin 77-04, POST-LOCA CONTAINMENT SUMP pH, which required that des*ign conditions in the containment sump be met for all circumstances allowed by Technical Specifications.
Safety *E*valuation NRC Bulletin 77-04 required all licensees of nuclear plants to establish a maximum boron concentration in the RWST in order to maintain proper pH in th.e containment sump during a postu-lated LOCA.
Due to interrelationships between RWST level set-points, operator response ti-me, boron concentration, sump pH, and injection requirements for adequate RHR Pump NPSH, an analysis was per"formed to determine *the "best fit" for all factors.
The analysis revealed that a maximum boron concen-tration of 2200 ppm would satisfy all the criteria *** provided the volume of the spray additive tank was at least 2310 gallons of 30% NaOH solution and the RWST low-level setpoint was 150,SQQ gallons as measured from the tank bottom.
The proposed Technical Specification changes will bring Unit No. 1 limits into agreement with both Unit No. 2 Technical Specifications and with the results of the. analysis performed.
9210190515 821005 PDR ADOCK 05000272 p
REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLE:
- a.
A boric acid storage system and associated heat tracing with:
- b.
- 1.
A minimum contained volume of 5106 gallons,
- 2.
Between 20,100 and 21,800 ppm of boron, and
- 3.
A minimum solution temperature of 145°F.
The
- 1.
- 2.
refueling water storage tank with:
~ 4 0
0 I briwee"' 3"'t, 5"00 oo., O A miRim~m contained volume ofAd58;899 gallons of water, be..Jw~ :Z.Ooo ~
- t.2.00 A.m:iflim~m boron concentration oi~ ppm, and
- 3.
A minimum solution temperature of 35°F.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
- a.
With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at least 1% ~k/k at 200°F; restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- b.
With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE:
SALEM - UNIT l 3/ 4 1-16 I:
EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:
be.fw1!.~ 3G~, 500 "11cf.. 400, ooo
- a.
Aill"i~im~m contained volume of.@§:8+8~ gallons of borated water
- ~
be.\\-v.Je.tM A.000 -d iL?.OO
- b.
A4Yli nimc11,, boron concentration of ~ppm, and
- c.
A minimum water temperature of 35°F.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
With the refueling water storage tank inoperable, restore the tank to OPERABLE status within l hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:
- a.
At least once per 7 days by:
- 1.
Verifying the water level in the tank, and
- 2.
Verifying the boron concentration of the water.
- b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST.temperature when the outside air temperature is< 35°F.
SALEM - UNIT l 3/4 5-9 I
"'-. /'.. - -
CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The spray additive system shall be OPERABLE with:
cd-.e.tw e..e..n 2.S" 8 a.vJ 4000 A spray additive tank containing a volume ofV~ 1 Hsiibs~88
- a.
gallons of Ai&~}e&stt~tS& percent by weight NaOH solution, and be~ee."" ~o a....J.~2..
- b.
Two spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a containment spray system pump flow..
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
With the spray additive system inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the spray additive system to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.2.2 The spray additive system shall be demonstrated OPERABLE:
- a.
At least once per 31 days also by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b.
At least once per 6 months by:
- 1.
Verifying the solution level in the tank, and
- 2.
Verifying the concentration of the NaOH solution by chemical analysis.
- c.
At least once per 18 months -during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment High-High pressure test signal.
- d.
At least once per 5 years by verifying a NaOH solution flow rate of 7.3 + 0.7 gpm from the spray additive tank through sample valve-1CS61 with the spray additive tank at 2.5 + 0.5 psig.
SALEM - UNIT 1 3/4 6-10
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PROPOSED CHANGES SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS TECHNICAL SPECIFICATION SALEM UNIT NOS. 1 and 2 Description of Changes Ref. LCR 82-05
- 1.
Change in core limits as 2,250 and 2,400 psia, the proposed core limits are slightly more limiting in the areas bounded by the quality and typical cell DNB limits.
Reason for Changes This change is required to justify the proposed change to the F~H Technical Specification.
PROPOSED CHANGE NUCLEAR ENTHALPY HOT CHANNEL FACTOR -F ~N Description of Change Modify the limit of F~ at fractional thermal power by changing the multiplier from O.~ to 0.3 as follows:
F~N
< 1.55 [l.0+/-0.3(1-P] [1-RBP (BU)]
Reason for Change Allow optimization of core loading patterns at hot full power conditions without incurring a restrictive F ~H limit at low
- power, This could lead to increased fuel utilization, since fresh or depleted burnable poison rod assemblies are often used to flatten power at reduced power level at the expense of cycle lifetime.
Safety Evaluation The effect of the increased F ~H limit at reduced thermal power on the core limits and on axial offset DNB limits has been reviewed.
The results show that the core limits at 2,250 and 2,400 psia and the axial offset DNB limit are more restrictive than the current limits.
Changing these limits to the more restrictive ones (see proposed changes to the Safety Limits and Limiting Safety System Settings Technical Specification) justifies the proposed change in the F ~H limit.
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SAFETY LiMITS.
BASES N
The curves are based 9n an enthalpy hot channel factor, F H' of 1.55 and a reference. cosine with a peak of 1.55Nfor axial power sha~e. An allowance is included for an increase in F6H at reduced power based on the expression:
0.3 F~H = 1.55 [1+~ (1-P)]
where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions ~re higher th~n those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod. insertion assuming the axial-power imbalance is within the limits *of the f1 (6!). function.of the Overtemperature trip.
When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature 6T trips will reduce the setpoints to provide protection consistent with core safety limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects "the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are desjgned to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure ~f 110% (2735 psig) of design pressure.
The Reactor Coolant System piping and fitti-ngs are designed to ANSI B 31.1 1955 Edition while the valves are designed to ANSI B 16.5, MSS*SP-66-1964, or ASME Section III-1968, which permit maximum transient pressures of up to 120% (2985 psig) of component design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3107 psig, 125~
of design pressure, to demonstrate integrity prior to initial *operation.
SALEM - UNIT 1 B 2-2
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- e.
- -POWER DISTRIBUTION LIMITS*
NUCL~R ENTHALPY HOT CHANNEL FACTOR -
F~H LIMITING CONDITION FOR OPERATION 3.2.3
~H sha11 be limited by the fo11~wing relationship:
JJ 0.3 F6H i 1.55 [1.0 + ~
(1-P)J [1-RBP (BU)J Whe~e* p _ THERMAL POWER
, and I
-RATED THERMAL POWER RBP(BU)= Rod Bow Penalty as a function of region -average.
burnup as shown in Figure 3.2-3, where a region is defined as those assemblies with the same loading date (reloads) or enrichment (first core).
APPLICABI"'t{TY:
MODE 1 ACT!ON:
With F~H exceeding its 1imit:
- a.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints to< 55~ of RATED THERMAL POWER within the next-4
-hours,
- b.
,... Demonstrate thru in..;core mapping that ~H is within its 1 imit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the 1im1t or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and
- c.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a. or b. abRve; subsequent POWER OPERATION may proceed provided that ~~H is demonstrated through in-core mapping to be within its limit at a nominal 50% of RArED THERMAL POWER prior to exceeding this iHERMAL POWER, *at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95~ or greater RATED THERMAL POWER.
SALEM - UNIT 1 3/4 2-9 Amendmeiilt No.
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SALEM - UNIT 2 Rz = R,1(1 -
RBP<BU) 1 Figure 3.2*3 RCS TOTAL FLOWRATE VERSUS R -
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- r POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND R LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor-Coolant System (RCS) total flow rate and R1, R? shall be maintained within the region of allowable operation shown on F1gure 3.2-3 for 4 loop operation.
Where:
- a.
- b.
- c.
N F~H Rl = L49 [l.O +M""(l.O - P)]
0.3 Rl R2 = [l-RBP(BU)],.
_. THERMAL POWER P - RATED THERMAL POWER ' and d;
F~H = Measured val~es of F~H obtained.by using the movab.le *incor;-e -
detectors to obtain a power distribution map.. The measured values of F~H shall be used to calculate R since Figure 3.2-3 includes measurement uncertainties of 3.5% for flow and 4% for
. incore measurement of F~H*
- e.
RBP (BU) = Rod Bow.Penalty as a function of region average burnup as shown in Figure 3.2-4, where a region is defined as those assemblies with the same loading date (reloads) or enrich-ment (first core).
APPLICABILITY:
MODE l.
ACTION:
With the combination of RCS total flow rate and R1, R2 outside the region of acceptable operation shown on Figure 3.2-3:
- a.
Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
- l.
Either restore the combination of RCS total flow rate and R1, R2 to within the above limits, *or
- 2.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to less
~han.or equal to 55% of RATED THERMAL POWER within the next.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SALEM - UNIT 2 3/4 2-9
r
- ~.
\\.. *.
- 2. 1 SAFETY LIMITS BASES
- 2. 1. 1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in.excessive cladding temperatures because of the onset of departure from nucleate *bailing (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation.
The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local ONB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cau~e DNB at a particular core location to the local heat flux, is indicative of the margin to ONB.
The minimum value of the DNBR during steady state operation, normal operational transients~ and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that ONB will not occur and is chosen as an appropriate margin to ONB for all operating conditions.
The curves of Figures 2. 1-1 and 2. 1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
The curves are bas:ed on an entha 1 py hot channe 1 factor, FN, of 1. 55 and a reference cosine with a pea~ of 1.55 for axial power shape. ~n allowance is included for an increase_ in F.6.H at reduced power based on the expression: J N
0.3 F.6.H = 1. 55 [l + ~(1-P)]
where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f1(delta I) function of the Overtemperature trip.
When the axial power SALEM - UNIT 2 B 2-1
I I
PROPOSED LICENSE CHANGE SALEM GENERATING STATION UNIT NOS. l AND 2 DESCRIPTION OF CHANGE Ref. LCR 82-14 Change paragraph 4.4.6.2d for Unit l and 4.4.7.2d for Unit 2 to read as follows:
- d.
Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> *during steady state operation.
- The provisions of specification 4.0.4 are not applicable for entry into Mode 4.
- Addition to Unit 2
- Addition to Unit l REASON FOR CHANGE This change is re*quired to provide consistency between the unit Tech Specs.
The additional wording indicated for each unit is presently in the existing paragraph for the other unit.
SAFETY EVALUATION This change does not involve an unreviewed safety question since it merely calls for the wording of each Tech Spec paragraph to be consolidated into a consistent wording for each unit.
1l
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License Change Request Attachment Ref. LCR 82-15 DESCRIPTION OF CHANGE Add the following note to Tech. Spec. &ectiori 3.3.2.l Table -3.3.3 Section 4b:
The automatic actuation logic includes the redundant solenoid operated vent valves for each Main Steam Isolation Valve.
One vent valve on any one Main Steam Isolation Valve may be isolated without affecting the function of the automatic actuation logic provided the remaining seven solenoid vent valves remain operable.
For this condition entry into the Action Statement (No. 13 on Unit 1, No. 20 on Unit 2) is not required.
REASON FOR CHANGE This change is being submitted to more clearly define the bounds of the automatic actuation logic and to delineate the restrictions for operating with one solenoid vent valve isolated.
SAFETY EVALUATION
- a.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report will not be increased.
Eight solenoid valves are provided for four Main Steam Isolation Valves (MSIV' s).
The una".tailability of one solenoid valve will not render the associated MSIV inoperable.
An indepe.ndent redundant solenoid valve is provided for control of the affected MSIV.
Given one inoperable solenoid valve and subsequent failure of the redundant logic train, three of the four MSIV's will close thus providing the safety function.
Therefore, the loss of one out of eight solenoid valves will not impact plant safety.
Should more than one solenoid valve be ~noperable at any given time, plant safety could be impacted resulting in an unreviewed safety question.
- 0.
The possibility of an accident or malfunction of a di ffera!'tt.
type than any evaluated previously in the safety analysis report will not be created.
The subject solenoid valves affect only MSIV operation and interface with no.other systems or equipment other than the motive power systems necessary for operation.
This LCR is not related to a plant modification.
r:
f
/
License Change Request Attachment (MSIV' s)
- c.
The margin for safety as defined in the basis for ~ny technical specification is not reduced.
The original technical specification related to this equipment was intended to apply to the two redundant SSPS logic trains whereas four MSIV's exist, each with redundant controls.
Three of the four MSIV's are required to close for safety purposes.
STATE OF NEW JERSEY)
) ss:
COUNTY OF SALEM COUNTY OF SALEM
)
Ref. LCR 80-05 LCR 82-05 LCR 82-14 LCR 82-15 RICHARD A. UDERITZ, being duly sworn according to law disposes and says:
I am a Vice President of Public Service Electric and Gas Company, and as such, I find the matters set forth in our Request for Amendment dated October 5, 1982, are true to the best of my knowledge, information and belief.
Subscribed and sworn to before me this
'JT'4 day of QC.TOBE.2-
, 1982.
RUDOLPH l. von FISCHER JR.
Notary Public of New Jersey My Commission Expires Sept. 10, 1988 My commission expires of