ML17059A067

Corrected Pages 18,21,26 & 197 to Amend 143 to License DPR-63

ML17059A067
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Site:	Nine Mile Point
Issue date:	10/04/1993
From:	Office of Nuclear Reactor Regulation
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ML17059A068	List: ML17059A067
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NUDOCS 9310120137
Download: ML17059A067 (12)

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-OO UdO VIUE BASES FOR 2.1.2 FUEL CLADDING - LIMITINGSAFETY SYSTEM SETTING However, in response to expressed beliefsl I that variation of APRM flux scram with recirculation flow is a prudent measure to assure safe plant operation during the design confirmation phase of plant operation, the scram setting will be varied with recirculation flow.

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity safety limitis reached.

The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation.

Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses.

Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity safety limityet allows operating margin that.

reduces the possibility of unnecessary scrams.

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of FRTP and CMFLPD. The scram setting is adjusted in accordance with Specification 2.1.2a when the core maximum fraction of limiting power density exceeds the fraction of rated thermal power.

Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.

The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at a constant recirculation flow rate, and thus to protect against the condition of a MCPR less than the SLCPR. This rod block trip setting, which is automatically varied with recirculation flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal.

The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range.

The margin to the safety limitincreases as the flow decreases for the specified trip setting versus fiow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 1 1 0'k of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in t core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the core maximum fraction of limiting power density exceeds the fraction of rated thermal power, thus, preserving the APRM rod block safety margin.

b.

Normal operation of the automatic recirculation pump control will be in excess of 30% rated flow; therefore, little operation below 30% flow is anticipated.

For operation in the startup mode while the reactor is at low pressure, the IRM scram setting is 12% of rated neutron flux. Although the operator willset the IRM scram trip at 12% of rated neutron flux or less, the actual scram setting can be as much as 2.5% of rated neutron fluxgreater.

This includes the margins discussed above.

This provides adequate margin between the setpoint and the safety limitat 25% of rated power.

The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.

There are a few possible sources of rapid reactivity input to the system in the low power flow condition.

Effects of increasing pressure at zero or low void content are AMENDMENTNO. Ak, 143 18

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BASES FOR 2.1.2 FUEL CLADDING-LIMITINGSAFETY SYSTEM SETTING f-g.

The low pressure isolation of the main steam lines at 850 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel.

Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram.

Thus, the combination of main steam line isolation on reactor low pressure and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.

With the sera~

set at a10% valve closuie, there is no increase in neutron flux and peak pressure if the vessel dome is limited to 1141 psi~

(8, 9, 10)

The operator willset the pressure trip at greater than or equal to 850 psig and the isolation valve stem position scram setting at less than or equal to 10%%d of valve stem position from full open.

However, the actual pressure set point can be as much as 15.8 psi lower than the indicated 850 psig and the valve position set point can be as much as 2.5% of stem position greater.

These allowable deviations are due to instrument error, operator setting error and drift with time.

In addition to the above mentioned Limiting Safety System Setting, other reactor protection system devices (LCO 3.6.2) serve as a secondary backup to the Limiting Safety System Setting chosen.

These are as follows:

High fission product activity released from the core is sensed in the main steam lines by the high radiation main steam line monitors.

These monitors provide a backup scram signal and also close the main steam line isolation valves.

The scram dump volume high level scram trip assures that scram capability willnot be impaired because of insufficient scram dump volume to accommodate the water discharged from the control rod drive hydraulic system as a result of a reactor scram (Section X-C.2.10)'.

h.

The generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast I

closure of the turbine control valves due to the worst case transient of a load rejection and subsequent failure of the bypass.

In fact, analysis 'hows that heat flux does not increaye from its initial value at all because of the fast action of the load rejection scram; thus, no significant change in MCPR occurs.

The turbine stop valve closure scram is provided for the same reasons as discussed in h above.

With a scram setting of

<10%%d valve closure, the resultant transients are nearly the same as for those described in h above; and, thus, adequate margin exists.

oUFSAR AMENDMENTNO. )ME, 143 21

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BASES FOR 2.2.2 REACTOR COOLANT SYSTEM - LIMITINGSAFETY SYSTEM SETTING C.

As shown in Sections XV-B.3.1 and 3.5", rapid Station transients due to isolation valve or turbine trip valve closures result in I

coincident high-flux and high-pressure transients.

Therefore, the APRM trip, although primarily intended for core protection, also serves as backup protection for pressure transients.

Although the operator willset the scram setting at less than or equal to that required by Specification 2.1.2a, the actual neutron flux setting can be as much as 2.7 percent of rated neutron flux above the specified value.

This includes the errors discussed above.

The flow bias could vary as much as one percent of rated recirculation flow above or below the indicated point.

In addition to the above-mentioned Limiting Safety System Setting, other reactor protection system devices (LCO 3.6.2) serve as secondary backup to the Limiting Safety System Setting chosen.

These are as follows:

The primary containment high-pressure scram serves as backup to high reactor pressure scram in the event of lifting of the safety valves.

As discussed in Section VIII-A.2.1',a pressure in excess of 3.5 psig due to steam leakage or blowdown to the drywell willtrip a scram well before the core is uncovered.

The scram dump volume high-level scram trip assures that scram capability willnot be impaired because of.insufficient scram dump volume to accommodate the water discharge from the control-rod-drive hydraulic system as a result of a reactor scram (Section X-C.2.10) '.

In the event of main-steam-line isolation valve closure, reactor pressure will increase.

A reactor scram is, therefore, providedt on main-steam-line isolation valve position and anticipates the high reactor pressure scram trip.

'UFSAR AMENDMENTNO. )Mk, 143 26

Docket No. 50-220 Hr.

B. Ralph Sylvia Executive Vice President, Nuclear Niagara Mohawk Power Corporation 301 Plainfield Road

Syracuse, New York 13212 October 4, 1993

SUBJECT:

CORRECTED TECHNICAL SPECIFICATION PAGES FOR NINE MILE POINT NUCLEAR STATION UNIT NO. 1, LICENSE AMENDMENT NO.

143 (TAC NO. H85074)

By letter dated September 24, 1993, you informed us that four of the technical specification (TS) pages (TS pages 18, 21, 26, and 197) Niagara Mohawk Power Corporation had provided us on June 30, 1993, for issuance with Nine Mile Point Nuclear Station Unit No.

1, License Amendment No.

143 contained some minor errors.

These errors were not detected prior to the issuance of License Amendment No.

143 on July 26, 1993.

Enclosed are corrected TS pages 18, 21, 26, and 197 which should be used to replace the subject pages that were issued on July 26,

1993, as part of License Amendment No.

143 to Nine Mile Point Unit No.

1 Operating License DPR-63.'incerely,'riginal signed by:

Donald S. Brinkman, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

Nine Mile Point Unit No.

1 Technical Specification Pages 18, 21, 26, and 197 cc w/enclosures:

See next page LA:P

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93 OFFICIAL RECORD COPY FILENAME: NM185074. LTR

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LIMITINGCONDITION FOR OPERATION SURVEILLANCEREQUIREMENT

.b.

During operation with.the Core Maximum Fraction of Limiting Power Density (CMFLPD) greater than the Fraction of Rated Thermal Power (FRTP), either:

(1)

The APRM scram and rod block settings shall be reduced to the values given by the equations in Specification 2.1.2a; (2)

The APRM gain shall be adjusted in accordance with Specification 2.1.2a; or c.

During reactor power operation at a25 percent rated thermal power, the Core Maximum Fraction of Limiting Power Density (CMFLPD) shall be checked daily and the flow-referenced APRM scram and rod block signals shall be adjusted, if necessary, as specified by Specification 2.1.2a.

(3)

The power distribution shall be changed such that the CMFLPD no longer exceeds FRTP.

AMENDMENTNO. Qk, 143 197

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~ j Docket No. 50-220 Hr. B. Ralph Sylvia Executive Vice President, Nuclear Niagara Mohawk Power Corporation 301 Plainfield Road

Syracuse, New York 13212 October 4, 1993

SUBJECT:

CORRECTED TECHNICAL SPECIFICATION PAGES FOR NINE MILE POINT NUCLEAR STATION UNIT NO. 1, LICENSE AMENDMENT NO.

143 (TAC NO. H85074)

By letter dated September 24,

1993, you informed us that four of the technical specification (TS) pages (TS pages 18,. 21, 26, and 197) Niagara Mohawk Power Corporation had provided us on June 30, 1993, for issuance with Nine Mile Point Nuclear Station Unit No.

1, License Amendment No.

143 contained some minor errors.

These errors were not detected prior to the issuance of License Amendment No.

143 on July 26, 1993.

Enclosed are C

corrected TS pages 18, 21, 26, and 197 which should be used to replace the subject pages that were issued on July 26,

1993, as part of License Amendment No.

143 to Nine Mile Point Unit No.

1 Operating License DPR-63.

Sincerely, Original signed by:

Donald S.

Brinkman, Senior Project Manager Project Directorate I-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation

Enclosures:

Nine Mile Point Unit No.

1 Technical Specification Pages 18, 21, 26, and 197 cc w/enclosures:

See next page LA:P

-1 CV a""

93

'PH:PDI-1 DBrinkm n:smm 93 D:PDI-1 RACa ra ~gC S

093 OFFICIAL RECORD COPY FILENAME: NH185074. LTR

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