ML20197K104

Revised Proposed TS Bases Reflecting Previous Removal of Condenser Low Vacuum Scram Function from TS as Well as Plant Design & Changes to EDG Ratings,Design Basis Load Limits & Loading Profiles

ML20197K104
Person / Time
Site:	Nine Mile Point
Issue date:	12/04/1998
From:	NIAGARA MOHAWK POWER CORP.
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ML20197J729	List: ML20197J726 ML20197K104
References
NUDOCS 9812160086
Download: ML20197K104 (10)

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ATTACHMENT A NIAGARA MOHAWK POWER CORPORATION .

I LICENSE NO. DPR-63 DOCKET NO. 50-220 i Channes to the Technical Soecification Bases Replace the existing pages 250,251, and 258 with the attached corresponding revised l pages. The revised replacement pages have been retyped in their entirety, incorporating :

the changes, and include marginal markings (revision bars) to indicate the changes. l l

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1 9912160086 901204 ff PDR ADOCK 05000220 u P PDR .

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BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION  ;

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a. The set points included in the tables are those used in the transient analysis and the accident analysis. The high flow set point for

• the main steam line is 105 psi differential. This represents a flow of approximately 4.4x10 6 lb/hr. The high flow set point for the i emergency cooling system supply line is s 11.5 psi differential. This represents a flow of approximately 9.8x105lb/hr at rated conditions. _

i The automatic initiation signals for the emergency cooling systems have to be sustained for more than 12 seconds to cause opening -

of the return valves. If the signals last for less than 12 seconds, the emergency cooling system operating wi!! not be automatically j initiated.

The high level in the scram discharge volume is provided to assure that there is still sufficient free volume in the discharge system to t receive the control rod drives discharge. Following a scrarr, bypassing is permitted to allow draining of the discharge volume and resetting of the reactor protection system relays. Since all control rods are completely inserted following a scram and since the l bypass of this particular scram initiates a control rod block, it is permissible to bypass this scram function. The scram trip associated with the shutdown position of the mode switch can be reset after 10 seconds. ,

t The condenser low-low-low vacuum and the main steam line isolation valve position signals are bypassed in the startup and refuel. l positions of the reactor mode switch when the reactor pressure is less than 600 psig. These are bypassed to allow warmup of the main steam lines and to provide a heat sink during startup. l ,

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-i AMENDMENT NO. //d,149 250

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t BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The set points on the generator load rejection and turbine stop valve closure scram trips are set to anticipate and minimize the = '

l consequences of turbine trip with failure of the turbine bypass system as described in the bases for Specification 2.1.2. Since the severity  ;

of the transients is dependent on the reactor operating power level, bypassing of the scrams below the specified power level is permissible.

Although the operator will set the setpoints at the values indicated in Tables 3.6.2.a-1, the actual values of the various set points can differ eppreciably from the value the operator is attempting to set. The deviations include inherent instrument error, operator setting error and -  :

drift of the set point. These errors are compensated for in the transient analyses bv conservatism in the controlling parameter assumptions  !

es discussed in the bases for Specification 2.1.2. The deviations associated with the set points for the safety systems used to mitigate -

accidents have negligible effect on the initiation of these systems. These safety systems have initiation times which are orders of ,

magnitude greater than the differcace in time between reaching the nominal set point and the worst set point due to error. The maximum tilowable set point deviations are listed t:elow:

N2utron Flux i APRM Scram,12.3% of rated neutron flux (analytical limit is 120% of rated flux)  ;

, APRM Rod Block, i2.3% of rated neutron flux (analytical limit is 110% of rated flux)  ;

IRM.12.5% of rated neutron flux  ;

The APRM downscale rod block setpoint has been derived based on GE setpoint methodology as outlined in NEDC-31336, "GE  ;

instrumentation Setpoint Methodology." In this methodology, tho setpoint is defined as three values, Nominal Trip Setpoint, Allowable Vclue, and Analytical Limit. Table 3.6.2g shows the nominal trip setpoints. The corresponding allowable value is as follows

APRM Downscale Rod Block, allowable value is 2:[4.24/1251 divisions of full scale Recirculation Flow Upscale, i1.6% of rated recirculation flow (analytical limit is 107.1% of rated flow)

R:: circulation Flow Comparator, i2.09% of rated recirculation flow (analytical limit is 10% flow differential)

R:;ector Pressure,115.8 psig Containment Pressure,10.053 psig R: actor Water Level, 2.6 inches of water  ;

i Main Steam Line Isolation Valve Position, 2.5% of stem position l Scram Discharge Volume, +0 and -1 gallon i

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AMENDMENT NO. /dd,153 251 .

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BASES FOR 3.6.3 AND 4.6.3 EMERGENCY POWER SOURCES

Other than tho' Station turbine generator, the Station is supplied by four independent sources of a-c power; two 115 kv transmission lines,.

and two diesel-generators. : Any one of the required power sources will provide' the power required for a LOCA. Engineenng calculations ,

! show that a LOCA concurrent with a loss of offsite power and the smgle failure of one of the diesel-generators results in a loading for the l remaining diesel-generator that is below the unit's 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> / year rating. This loading is. greater than that required during a Station i shutdown condition. The monthly test run paralleled with the system is based on the manufacturer's recommendation for thess units in this . ;

type of service. The testing during operating cycle will simulate the accident conditions under which operation of the diesel-generators is  !

required. ' The major equipment comprising the maximum diesel-generator loading is given in Figure IX-6*. j l

i' As mentioned above, a single diesel-generator is capable of providing the required power to equipment following a LOCA. Two fuel oil .

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I ctorage tanks are provided with -jping interties to permit supplyin0 either diesel-generator. A two-day supply will pr& vide adequate time to arrange for fuel makeup if needed. The full capacity of both tanks will hold a four-day supply. ~

i it has been demonstrated in Section XV.B.3.23' that even with complete d-c loss the reactor can be safely isolated and the emergency . .l-cooling system will be operative with makeup water to the emergency cooling system shells maintained manually. Having at least one d-c'  :

battery available will permit: automatic makeup to the shells rather than manual, closing of the d-c actuated isolation valve on all lines from .

3 the primary system and the suppression chamber, maintenance of electrical switching functions in the Station and providing nmergency i lighting and communications power.

A battery system shalt have a minimum of 106 volts at the battery terminals to be considered operable.

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'FSAR  ;

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, AMENDMENT NO.142 258'

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ATTACHMENTB NIAGARA MOHAWK POWER CORPORATION LICENSE NO. DPR-63 DOCKET NO. 50 220 Suonortina information for the Technical Soecification Bases Chanaes I. Succortino information for the Chanaos to the "Basas for 3.6.2 and 4.6.2 Protective Instrumentation" INTRODUCTION This revision to the Nine Mile Point Unit 1 (NMP1) Technical Specification (TS) " Bases for 3.6.2 and 4.6.2 Protective Instrumentation" updates the Bases to reflect the previous  ;

removal of the Condenser Low Vacuum scram function from the TSs, as well as from the plant design. These Bases are further revised to include changes that correct and clarify the information regarding a Reactor Protection System (RPS) bypass feature.

The Condenser Low Vacuum scram at 23 in. Hg was provided to anticipate a main steam line isolation on loss of condenser vacuum at 7 in. Hg (Condenser Low-Low-Low Vacuum),

thereby providing protection to balance of plant equipment and also limiting the reactor pressure transient. However, the main steam line isolation valve closure transient analyses were based on solenoid-actuated relief valve actuation without the anticipatory Condenser l

Low Vacuum scram. Since the results of these analyses showed the reactor pressure j transients to be within permissible limits,it was concluded that the Condenser Low 1 Vacuum scram is not necessary for reactor protection, and therefore, not subject to the reactor protection criteria. As a result, it was determined that the Condenser Low Vacuum scram function is not necessary for reactor protection and TS Amendment No. 37 deleted the function from Tables 3.6.2a and 4.6.2a, " Instrumentation that initiates a Scram." ,

l Modification N1-88-053 subsequently disconnected and removed the RPS circuitry I associated with the Condenser Low Vacuum scram signal and eliminated non-coincident <

scram bypass circuitry in order to reduce unnecessary plant scrams. Although TS Amendment No. 37 and Modification N188-053 removed the Condenser Low Vacuum scram function from the TSs and plant design, two references to the Conder.ser Low Vacuum scram signal still remained in the " Bases for 3.6.2 and 4.6.2 Protective t Instrumentation." The Bases are being revised to remove these references.

The main steam line isolation trip on Condenser Low Low-Low Vacuum is automatically l bypassed when the reactor mode switch is in the "Startup" or " Refuel" positions and l

reactor pressure is less than 600 psig. The Main Steam isolation Valve (MSIV) Position scram is automatically bypassed when the reactor mode switch is in the "Startup,"

" Refuel," or " Shutdown" positions and reactor pressure is less than 600 psig it is necessary to bypass the main steam line isolation trip and scram functions to allow warmup of the main steam lines and to provide a heat sink during reactor startup. The

" Bases for 3.6.2 and 4.6.2 Protectivo instrumentation" identify the Condenser Low

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Vacuum and Condenser Low-Low Vacuum signals as incorporating this automatic bypass  ;

feature. The Condenser Low Vacuum scram signal should no longer be identified as having the bypass feature since the signal was previously removed from the TSs and plar.t design, 1 of 6

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as discussed in the previous paragraph. Furthermore, the Cm 'enser Low-Low Vacuum signal does not include the automatic bypass feature. The Condenser Low-Low Vacuum signal initiates closure of the main turbine stop valves at 20 in. Hg. The bypass feature is not necessary for this protective function since the turbine stop valves would not normally be open when the reactor mode switch is in the "Startup" or " Refuel" positions, with reactor pressure less than 600 psig. The Bases are being revised to reflect the correct information regarding the automatic bypass feature. in addition, an editorial change is included which clarifies the reasons for the bypass feature.

On page 250, the fourth paragraph of the current TS " Bases for 3.6.2 and 4.6.2 Protective Instrumentation" reads as follows:

The condenser low vacuum, low-low vacuum and the main steam line isolation valve position signals are bypassed in the startup and refuel positions of the reactor mode switch when the reactor pressure is less than 600 psig. These are bypassed to allow warmup of the main steam lines and a heat sink during startup.

Niagara Mohawk Power Corporation (NMPC) is revising the paragraph to read as follows:

l The condenser /ow-low-low vacuum and the main steam line isolation valve position l signals are bypassed in the startup and refuel positions of the reactor mode switch when the reactor pressure is less than 600 psig. These are bypassed to allow warmup of the main steam lines and to provide a heat sink during startup.'

l On page 251, the current TS " Bases for 3.6.2 and 4.6.2 Protective Instrumentation" contains a list of selected RPS parameters and provides their maximum allowable setpoint I deviations (tolerances). The last parameter and setpoint tolerance on the list reads as '

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1 Condenser Low Vacuum, 0.5 inches of mercury I

NMPC is removing the Condenser I.ow Vacuum parameter and its setpoint tolerance of l 10.5 inches of mercury from the list. \

l EVALUATION On page 250, the fourth paragraph of the current TS " Bases for 3.6.2 and 4.6.2 Protective Instrumentation" provides a list of TS required RPS signals that are bypassed in the ,

"Startup" and " Refuel" positions of the reactor mode switch when the reactor pressure is l less than 600 psig. The three RPS signals listed are the Condenser Low Vacuum (scram) signal, the Condenser Low-Low Vacuum signal, and the MSIV Position (scram) signal.

Currently, of these three signals, only the MSIV signal is correctly identified. The Condenser Low Vacuum scram signal and the Condenser Low-Low Vacuum signal should be deleted from the list and the Condenser Low-Low-Low Vacuum signal should be added l to tne list. The Bases are being revised to incorporate these corrections. The reasons for the changes are discussed below j

• The Condenser Low Vacuum scram signal (23 in. Hg) should be deleted from the

list of RPS signals provided in the fourth paragraph of the TS Bases on page 250.

j As previously described, the Condenser Low Vacuum scram signal is not necessary i

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for reactor protection and was removed from TS Tables 3.6.2a and 4.6.2a by Amendment No. 37. A modification subsequently disconnected and removed the associated RPS trip circuitry. The Condenser Low Vacuum scram signal is being deleted to update the Bases information to reflect its previous removal from the TSs, as well as from the plant design.

The list of RPS signals provided in the fourth paragraph of the TS Bases on page 250 contain only TS sequired RPS instrumentation signals that are automatically bypassed when the reactor mode switch is in the "Startup" or " Refuel" podier 3 and reactor pressure is less than 600 psig. The reference to the Condens r u w-Low Vacuum signal (20 in. Hg) is being dele.3d from the list since it is n Lod never was) a TS required RPS instrumentation signal, nor does it (and never de incorporate the described bypass feature. The list is being revised to correctly identify the Condenser Low-Low Low Vacuum signal as incorporating the bypass feature. The Condenser Low-Low Low Vacuum signal is idemified as Parameter (6) in TS Tables 3.6.2b and 4.6.2b, " Instrumentation that Initiates Primary Coolant System or Containment isolation," and Note (a) for the parameter identifies it as having the bypass feature. Descriptions of the Condenser Low-Low Low Vacuum isolation function and bypass feature are provided in Section Vill of the NMP1 UFSAR. This change restores consistency between the TSs, the TS Bases, and the UFSAR. In addition, an editorial change inserts the words "to provide" in the second sentence (fourth paragraph on page 250) to clarify the reason for the bypass feature.

On page 251, the current TS " Bases for 3.6.2 and 4.6.2 Protective Instrumentation" contains a list of selected RPS parameters and provides their maximum allowable setpoint tolerances. The Condenser Low Vacuum signal and its associated setpoint tolerance comprise the last item on the list. As previously discussed, the Condenser Low Vacuum scram signal is not necessary for reactor protection and was removed from TS Tables 3.6.2a and 4.6.2a by Amendment No. 37. A modification subsequer.tly disconnected and removed the associated RPS circuitry. Thus, the Condenser Low Vacuum scram signal and the associated trip and bypass logic circuitry no longer exist in the plant design or in the TSs. Accordingly, the Condenser Low Vacuum scram signal and its associated setpoint tolerance are being deleted from the TS Bases.

CONCLUSIONS The Condenser Low Vacuum scram function and associated setpoint tolcrance value are l being removed from the TS " Bases for 3.6.2 and 4.6.2 Protective Instrumentation." '

Analyses show that the reactor pressure transients resulting from a main steam line isolation on loss of condenser vacuum will remain within permissible limits without the Condenser Low Vacuum scram. As a result, Amendment No. 37 removed the scram function from the TSs and a modification subsequently disconnncted and removed the associated RPS circuitry. The Condenser Low Vacuum scram function and associated setpoint tolerance calue are being removed from the TS Bases consistent with the current TSs and pCant design. These are informational changes, and as such, will r ot increase the probability or conseqLences of an accident or malfunction previously analyzed, create a new or different kind of accident or malfunction, or reduce a margin of safety. It is, therefore, concluded thst the changes will not adversely affect nuclear safety or the health and safety of the public. j I

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. 1 The Condenser Low-Low Vacuum signal is incorrectly included on the TS Bases list of RPS signals that are bypassed in the"Startup" and " Refuel" positions of the reactor mode switch when the reactor pressure is less than 600 psig. Based on the TSs and UFSAR, it is the Condenser Low Low-Low Vacuum signal that has this bypass feature incorporated hito its design. Accordingly, the list is being revised to identify the Condenser Low-Low-Low Vacuum signal as incorporating the b/ pass feature, in addition, an editorial change clarifies the reason for.the bypass feature. These changes are informational in nature and have no safety significance since there are no actual changes to the facility. It is, therefore, concluded that the changes will rst adversely affect public health and safety. j

11. Suooortino information for the Channes to the " Bases for 3.6.3 and 4.6.3 Emeroency Power Sources" INTRODUCTION This revision to the NMP1 TS " Bases for 3.6.3 and 4.6.3 Emergency Power Sources,"

updates the Bases to be consistent with the UFSAR. The Bases are being revised to reflect changes to'the Emergency Diesel Generator (EDG) ratings, design basis load limits, and loading profiles which resulted from the reconstitution of the EDG design basis loading analysis. In addition, references to the Final Safety Analysis Report (FSAR) are replaced with their corresponding UFSAR references.

The EDG system includes two redundant and independent EDGs (102 and 103) which provide onsite emergency power to their respective Power Boards (102 and 103). The EDG system is described in Section IX.B.4.1 of the NMP1 UFSAR. The primary safety- l related function of the EDG system is to provide sufficient electrical power to operate the l loads needed for accident mitigation and safe shutdown when offsite power is not i available. The emergency loads supp!ied fmm each Power Board comprise an independent l load group and each load group is capable of satisfying the accident mitigation and safe l shutdown requirements of the plant. Each EDG is designed to independently start and carry the maximum anticipated emergency load supplied from its Power Board. Therefore, one EDG can provide sufficient electrical power to satisfy the primary safety-related function of the EDG system. With concurrent loss of offsite power and LOCA signals, the EDGs automatically start and load onto their Power Boards. This is followed by automatic sequential loading of the engineered safeguards loads. Additional loads required for accident mitigation are manually added. Each EDG is rated for continuous operation at 2586 KW and 10% ovstload operation at 2845 KW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. In addition, each EDG unit has a 2000 hr/yr rating of 2838 KW and a 7 day /yr emergency rating of 2945 KW. These ratings are limited by engine output capability.

The 2000 br/yr rating (2838 KW) is used as the design basis load limit in the EDG loading an6 lysis (Engineering Calculation No. 4.16KVAC-EDG-ES) and the results are presented in Figure IX-6 of the NMP1 UFSAR. The Figure demonstrates that the maximum anticipated continuous load following a LOCA is maintained within the 2000 hr/yr rating. Use of the 2000 hr/yr rating as the design load limit is somewhat conservative with respect to the guidance provided in Revision 2 of Regulatory Guide 1.9, " Selection, Design, and Qualification of Diesel-Generator. Units As Standby (Onsite) Electric Power Systems At Nuclear Power Plants," which was in effect at the time Calculation 4.16KVAC-EDG-ES was originally issued. The Regulatory Guide states that predicted loads should not exceed the short time rating of the diesel-generator unit as defined in Section 3.7.2 of IEEE Std 387-4 of 6

1 1977. The short time rating (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period) is identified as the 10%  !

overload rarting by the diesel-generator vendor and, for the NMP1 EDGs, this rating is 2845 i KW.

On page 258, the first paragraph of the " Bases for 3.6.3 and 4.6.3 Emergency Power Sources" reads as follows:

Other than the Station turbine generator, the Station is supplied by four independent sources of a c power; two 115 kv transmission lines, and two diesel generators.

Any one of the required power sources will. provide the power required for the worst loss-of coolant accident. The required loads of 2500 kva and 2750 kva for the ,

loss-of-coolant are calculated in detail in the First Supplement to the FSAR. This loading is greater than that required during a Station shutdown condition. The '

monthly test run paralleled with the system is based on the manufacturer's i recommendation for these units in this type of service. The testing during operating cycle will simulate the accident conditions under which operation of the diesel-generators is required. A detailed tabulation of the equipment comprising the maximum diesel-generator load is given in the answer to Question V 10 of the First Supplement to the FSAR.

NMPC is revising the first paragraph to read as follows:

Other than the Station turbine generator, the Station is supplied by four independent sources of a-c power; two 115 kv transmission lines, and two diesel-generators.

Any one of the required power sources will provide the power required for a LOCA.

Engineering calculations show that a LOCA concurrent with a loss of offsite power and the single failure of one of the diesel-generators results in a loading for the remaining diesel-generator that is below the unit's 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> / year rating. This loading is greater than that required during a Station shutdown condition. The monthly test run paralleled with the system is based on the manufacturer's recommendation for these units in this type of service. The testing during operating cycle will simulate the accident conditions under which operation of the diesel-generators is required. The major equipment comprising the maximum diesel-generator loading is given in Figure IX 6',

Also on page 258, NMPC is revising the second and third paragraphs of the Bases as follows:

At the end of the first sentence of the second paragraph, the words major accident are being replaced with LOCA.

In the first sentence of the third paragraph, the reference to Appendix E-l.3.21 of the FSAR is being replaced with its corresponding UFSAR reference of Section XV.B. 3. 23.

EVALUATION The changes to the first paragraph of the Bases reflect the reaults of the current EDG loading analysis based on a LOCA concurrent with a hss of offsite power and single failure of one of the EDGs. The previously specified LOCA load values of "2500 KVA and 2750 5 of 6

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KVA" are being replaced with a statement that the calculated loading is below an EDG unit's 2000 hr/yr rating. Maintaining the EDG loading below the load limit constitutes the

. basis for determining that the current design is acceptable.

The changes to the first paragraph also replace the reference to "the answer to Question V 10 of the First Supplement to the FSAR" with its corresponding UFSAR reference of

" Figure IX 6." The referenced UFSAR Figure presents the current EDG loading information.

The Figure identifies the major equipment comprising the LOCA loading and demonstrates that the loads are within the EDG unit's 2000 hr/yr rating of 2838 KW. 1 The change to the first sentence of the second paragraph replaces the words " major accident" with "LOCA" for consistency with the first paragraph. The second paragraph establishes the basis for the four day EDG fuel supply by reiterating the basis established in the first paragraph that a single EDG is capable of satisfying the emergency power source requirements for a LOCA. The technical content of the second paragraph is not altered as the only " major accident" mentioned in the first paragraph is a "LOCA."

The change to the first sentence of the third paragraph, replaces the FSAR reference to

" Appendix E-l.3.21" with its corresponding UFSAR reference of "Section XV.B.3.23." The referenced UFSAR section describes the capability of the plant to withstand various DC interruptions. The first sentence of the Bases paragraph describes the plant's response to a loss of both main DC battery busses coupled with a loss of all AC power. The change in references doss not technically alter the paragraph since both the FSAR and UFSAR references contain essentially identical descriptions of the results of the event.

l CONCLUSIONS l

The changes to the " Bases for 3.6.3 and 4.6.3 Emergency Power Sources" reflect the results of the current EDG loading analysis and update the Bases consistent with the NMP1 UFSAR. The EDG loading analysis, as presented in the UFSAR, demonstrates that the maximum anticipated continuous load following a LOCA is maintained within an EDG's 2000 br/yr rating. This provides assurance that the EDG system is capable of performing its primary safety related function, which is to provide sufficient electrical power to operate the loads needed for accident mitigation and safe shutdown when offsite power is not available. The change updating the FSAR reference describing the plant response to a complete DC loss does not technically alter the Bases information. Accordingly, the changes will not increase the probability or consequences of an accident or malfunction previously analyzed, create a new or different kind of accident or malfunction, or reduce a margin of safety. It is, therefore, concluded that the changes will not adversely affect the public health and safety.

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