ML20239A181
| ML20239A181 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/31/1998 |
| From: | Andersen J NRC (Affiliation Not Assigned) |
| To: | Bowling M NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TAC-MA2305, TAC-MA3299, NUDOCS 9809080292 | |
| Download: ML20239A181 (11) | |
Text
- _ _ - _ _ _ _ _
lt
. Mr. M:rtin L.- Bowling, Jr.
Agat 31,' 1998
~ Recovtry Officer-Technical Strvices
'[
Northeast Nuclear Energy Comp:ny
~ /o Ms. Patricia A. Loftus c
Director-Regulatory Affairs P. O. Box 128
- Waterford, Connecticut 06385
SUBJECT:
MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 - REVISION TO TECHNICAL SPECIFICATIONS BASES (TAC NOS. MA2305 AND MA3299)
Dear Mr. Bowling:
By letters dated July 17 and 30,1998, Northeast Nuclear Energy Company (NNECO) provided the NRC with changes to Technical Specification (TS) Bases Sections 3/4.4.9 and 3/4.6.1.1.
NNECO provided the TS Bases page to the NRC for information only.
~
' As you are aware, the TS Bases are not part of the TSs as defined by 10 CFR 50.36. Changes
. to the TS Bases may voluntarily be made in accordance with the provisions of 10 CFR 50.59 Should the proposed change involve an unreviewed safety question pursuant to 10 CFR 50.59(a)(2), or involve a change in the interpretation of implementation of the TS (i.e., constitute a TS change), then the proposed change is to be provided to the staff pursuant to the provisions
)
of 10 CFR 50.59(c) and 10 CFR 50.90 for prior NRC review and approval.
I The TS Bases you provided are hereby retumed to you and should be inserted in the TS to ensure that the NRC staff and NNECO have identical TS Bases pages.. The staff did not perform l
. an evaluation of your TS Bases revisions and staff concurrence with the revisions is not implied by this letter. The staff may review the evaluations that support these TS Bases revisions during the next inspection of Millstone Unit 3's implemenaition of 10 CFR 50.59.
Sincere!,
sig ed by-James W. Andersen, Project Manager Special Projectsyffice - Licensing Office of Nuclear Reactor Regulation Docket No. 50-423
Enclosure:
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Mr. Martin L Bowling, Jr.
Recovery Officer - Technical Services Northeast Nuclear Energy Company clo Ms. Patricia A. Loftus Director-Regulatory Affairs P. O. Box 128 Waterford, Connecticut 06385
SUBJECT:
MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 - REVISION TO TECHNICAL SPECIFICATIONS BASES (TAC NOS. MA2305 AND MA3299)
Dear Mr. Bowling:
By letters dated July 17 and 30,1998, Northeast Nuclear Energy Company (NNECO) provided the NRC with changes to Technical Specification (TS) Bases Sections 3/4.4.9 and 3/4.6.1.1.
NNECO provided the TS Bases pages to the NRC for information only.
As you are awcre, the TS Bases are nn part of the TSs as defined by 10 CFR 50.36. Changes to the TS Bases may voluntarily be mr.de in accordance with the provisions of 10 CFR 50.59.
Should the proposed change involve tan unreviewed safety question pursuant to 10 CFR 50.59(a)(2), or involve a change in the interpretation of implementation of the TS (i.e., constitute
- a TS change), then the proposed enange is to be provided to the staff pursuant to the provisions of 10 CFR 50.59(c) and 10 CFR 50.90 for prior NRC review and approval.
i The TS Bases you provided are hereby retumed to you and should be inserted in the TS to j
ensure that the NRC staff and NNECO have identical TS Bases pages. The staff did not perform
.j sn evaluation of your TS Bases revisions and staff concurrence with the revisions is not implied by this letter. The staff may review the evaluations that support these TS Bases revisions during the next inspection of Millstone Unit 3's implementation of 10 CFR 50.59.
Sincerely, m s W. Andersen, Project Manager pecial Projects Office - Licensing Office of Nuclear Reactor Regulation Docket No. 50-423 j
Enclosure:
As stated i
l:
cc w/ encl: See next page
o.
REVISED TECHNICAL SPECIFICATIONS BASES FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following pages of the Appendix A, Technical Specifications Bases, with the attached pages. The revised pages contain verticallines indicating the areas of change.
Remove Insert B 3/4 4-16 B 3/4 416 B 3/4 4-16a B 3/4 4-20 B 3/4 4-20 B 3/4 4-22 B 3/4 4-22 B 3/4 6-1 B 3/4 6-1 B 3/4 6-ia B 3/4 6-1a l
l Enclosure l
l L-t-
l
' BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
The use of,a PORV for Cold Overpressure Protection is limited to those conditions when no more than one RCS loop is isolat.ed from the reactor vessel and, whenever an RCP is running, the temperature sign'al from the isolated loop is removed from the PORV opening logic. When two or more loops are isolated, Cold Overpressure Protection must be provided by either the two RHR suction relief valves or a depressurized and vented RCS.
The reactor vessel material is less tough at low temperatures than at normal op'erating temperature.. As the vessel neutron exposure accumulates, the material t-oughness decreases and becomes less resistant to stress at low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is I
increased only as temperature is increased.
I The potential for vessel overpressurization is most acute when the RCS is water solid, occurring while shutdown; a pressure fluctuation can occur more quickly than an operator can react to relieve the condition.
Exceeding the RCS P/T limits by a significant amount could cause nonductile cracking of __the reactor vessel.
LCO 3.4.9.1, " Pressure / Temperature Limits - Reactor Coolant System,"
requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the limits provided in Figures 3.4-2 and 3.4-3.
This LC0 provides RCS overpressure protection by limiting mass input capability and requiring adequate pressure relief capacity. Limiting mass input capability requires all Safety Injection (SIH) pumps and all but one centrifugal charging io pump to be incapable of injection into the RCS.
The pressure relief capacity requires either two redundant relief valves or'a depressurized RCS and an RCS vent of sufficient size.
One relief valve or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.
With minimum mass input capability, the ability to provide core coolant addition is restricted. The LC0 does not require the makeup control system deactivated or 'the safety injection (SI) actuation circuits blocked._ Due to the lower l
pressures in the Cold Overpressure Protection MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve.
PORV Requirements As designed, the PORV Cold Overpressure Protection (COPPS) is signaled to open if the RCS pressure approaches a limit determined by the COPPS actuation logic.
The COPPS actuation logic monitors both RCS temperature and RCS pressure and determines when the nominal setpoint of Figure 3.4-4a or Figure 3.4-4b is approached.-
The wide range RCS temperature indications are auctioneered to select the lowest temperature signal.
MILLSTONE - UNIT 3 8 3/4 4-16 Amendment No. pp, pp, 157, Revised by NRC letter dated C 3 1 W
REACTOR COOLANT SYSTEN BASES l
OVERPRESSURE PROTECTION SYSTEMS (continued)
PORV Requirements (continued)
The lowest temperature signal is processed through a function generator that calculates a pressure setpoint for that temperature.
The calculated pressure l
setpoint is then compared with the indicated RCS _ pressure from a wide range l
pressure channel.
If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to open.
The wide range RCS temperature indicators monitor loop temperature in the portion of the RCS loop that can be isolated. _ If a loop is isolated, the temperature signal from the isolated loop may be significantly lower than the temperature signals from the unisolated loops. This would result in a calculated pressure l
setpoint _ for the PORV below that anticipated by the operator based on the l
temperature in the unisolated portion of the RCS. Since this could result in a
'significantly lower calculated PORV setpoint, RCP operation is not permitted j
under these conditions unless the temperature input from the isolated loop is removed from the auctioneered circuit. This restriction will ensure that the #1 RCP seals are not challenged as a result of PORV undershoot. Since the PORV mass and heat injection transients have only been analyzed for a maximum of one loop isolated, the use of the PORVs is restricted to three and four RCS loops unisolated.
l' If one loop is isolated without removing its temperature input from the PORV calculated setpoint auctioneered circuit and at least one RCP is in operation,
- 4 or if more than one loop is isolated, then the PORVs must have their block valves closed or COPPS must be blocked. For these cases, Cold Overpressure Protection must be provided by either the two RHR suction relief valves or a depressurized RCS and an RCS vent.
The use of the PORVs is restricted to three and four RCS loops unisolated; for a loop to be considered isolated, both RCS loop stop valves must be closed. If only one loop stop valve in an RCS loop is closed for an extended +eriod, the PORVs must have their block valves closed or COPPS must be blocked. A single RCS loop stop valve can be stroked for short time periods for surveillance or other purposes and not affect the use of the PORVs for Cold Overpressure Protection.
The RHR suction relief valves have been qualified for all mass injection transients for any combination of isolated loops.
In addition, the heat
-injection transients not prohibited by the Technical Specifications have also been considered in the qualification of the RHR suction relief valves.
j-Figure 3.4-4a and Figure 3.4-4b present the PORV setpoints for COPPS.
Above l
Il0*F, the setpoints are staggered so only one valve opens during a low l
NILLSTONE - UNIT 3 8 3/4 4-16a Amendment No. pp, pp, J g I I D Revised by NRC letter dated
a.
REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
The cold overpressure transient analyses demonstrate that either one relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when RCS letdown is isolated and only one centrifugal charging pump is operating.
Thus, the LCO allows only one centrifugal charging pump capable of injecting when
-cold overpressure protection is required.
The. cold overpressure protection enabling temperature is conservatively i
established at a value 1275*F based on the criteria described in Branch j
Technical Position RSB 5-2 provided in the Standard Review Plan (NUREG-0800).
l PORV Performance The 10CFR50 Appendix G analyses show that the vessel is protected against non-ductile failure when the PORVs are set to open at the values shown in Figures 3.4-4a and 3.4-4b within the tolerance allowed for the calibration accuracy. The curves'are derived by analyses for both three and four RCS loops unisolated that
- model the performance of the PORV cold overpressure protection system (COPP assuming the limiting mass and heat transients of one centrifugal charging pump injecting into.the RCS, or the energy addition as a result of starting an RCP l
with temperature asymmetry between the RCS and the steam generators.
These analyses consider pressure overshoot and undershoot beyond the PORV opening and l
closing, resulting from signal processing and valve stroke times.
i l
The PORV setpoints-in Figures 3.4-4a and 3.4-4b will be updated when the P/T l
P limits conflict with the cold overpressure analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement.
Revised limits are determined using neutron fluence projections and the results of testing of the reactor vessel material irradiation l
surveillance specimens. The Bases for LC0 3.4.9.1, " Pressure / Temperature Limits
- Reactor Coolant System (Except the Pressurizer)," discuss these evaluations.
The PORVs are considered active components.
Thus, the-failure of ene PORV is assumed to represent the worst case, single active failure.
RHR Suction Relief Valve Performance i
The RHR suction relief valves do not have variable pressure and temperature lift setpoints as do the PORVs. Analyses show that one RHR suction relief valve with a setpoint at or between 426.8 psig and 453.2 psig will pass flow greater than that required for the limiting cold overpressure transient while maintaining RCS l
pressure less than the isothermal P/T limit curve. Assuming maximum relief flow l
requirements during the limiting cold overpressure event, an RHR suction relief valve will maintain RCS pressure to 5110% of the nominal lift setpoint.
Although each RHR suction relief valve.is a passive spring loaded device, which meets single failure criteria, its location within the RHR System precludes meeting single failure criteria when spurious RHR suction isolation valve or RHR suction valve closure is postulated.- Thus the loss of an RHR suction relief NILLSTONE - UNIT 3 B 3/4 4-20 Amendment No. JJ/,
Revised ny NRC letter dated M 81 %
I
REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued) c.
When COPPS is armed, PORY undershoot is analyzed for mass injection transients limited to one charging pump.
" Reactor Coolant System - Overpressure Protection Systems," provides this protection by requiring both safety injection pumps and all but one charging pump to be incapable of injection into the RCS.
In order to provide protection for the RCP #1 seal, a PORV setpoint of 1595 psia for temperatures 1160 degrees F must be met. This minimum setpoint is derived by adding the applicable train uncertainty and valve undershoot to the required minimum RCS pressure required for seal integrity.
Due to the differing instrument uncertainties for the two trains of PORV COPPS, the train with the highest uncertainty is paired to the high setpoint curve.
LC.Q This LC0 requires that cold overpressure protection be OPERABLE and the maximum mass input be limited to one. charging pump. Failure to meet this LCO could lead to the loss of low temperature overpressure mitigation and violation of the j
Reference 1 isothermal limits as a result of an operational transient.
To limit the mass input capability, the LC0 requires a maximum of one centrifugal charging pump capable of injecting into the RCS.
The elements of the LC0 that provides low temperature overpressure mitigation F
through pressure relief are:
1.
Two OPERABLE PORVs; or A PORV is OPERABLE for cold overpressure protection when its block valve is open, its lift setpoint is set to the nominal setpoints provided for both three and four loops unisolated by Figure 3.4-4a or 3.4-4b and when the surveillance requirements are met.
For three loops unisolated, the temperature input from the isolated loop must be removed from the COPPS auctioneered circuitry whenever any RCP is in operation.
2.
Two OPERABLE RHR suction relief valves; or An RHR suction relief valve is OPERABLE for cold overpressure protection when its isolation valves from the RCS are open and when its setpoint is at or between 426.8 psig and 453.2 psig, as verified by required testing.
3.
One OPERABLE PORV and one OPERABLE RHR suction relief valve; or 4.
A depressurized RCS and an RCS vent.
I An RCS vent is OPERABLE when open with an area of 2 5.4 square inches.
Each of these methods of overpressure prevention is capable of mitigating the limiting cold overpressure transient.
MILLSTONE - UNIT 3 B 3/4 4-22 Amendment No.
Revised by NRC letter dated 81 W
s 4
1 b
3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restric-l tion, in conjunction with the leakage rate limitation, will limit the SITE l
BOUNDARY radiation doses to within the dose guidelines of 10 CFR Part 100 l
during accident conditions and the control room operators dose to within the guidelines of GDC 19.
i The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative _ control includes the following considerations:
(1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing this operator to close' l
these ~ valves in an accident situation, and (3) assuring that environmental l
conditions will not preclude access to close the valves and that this action will i
prevent the release of radioactivity outside the containment.
When the Residual Heat Removal (RHR) System is placed in service in the plant cooldown mode of operation, the RHR suction isolation remotely operated valves 3RHS*MV8701A and 3RHS*MV8701B, and/or 3RHS*MV8702A and 3RHS*MV87028 are opened. These valves are normally operated from the control room. They do not receive an automatic containment isolation closure signal, but are interlocked to prevent their opening if Reactor Coolant System (RCS) pressure is greater than l
approximately 412.5 psia. When any of these valves are opened, either one of the
.two required licensed (Reactor Operator) control room operators can be credited as the operator required for administrative control. It is not necessary to use l
a separate dedicated operator.
3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total l
containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P.
As an added conservatism, the measured overall integrated leakage rate is further limited to less than 0.75 L, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.
The Limiting Condition for Operation defines the limitations on containment leakage rates for compliance with 10CFR50, Appendix J.
The leakage rates are verified by surveillance testing in accordance with the requirements - of Appendix J.
Although the LCO specifies the leakage rates at accident pressure, P.,. it is not feasible to perform a test at such an exact value for pressure.
I Consequently, the surveillance testing is performed at a pressure greater than
[
'or equal to P, to account for test instrument uncertainties and stabilization changes. This conservative test pressure ensures that the measured leakage rates i
MILLSTONE - UNIT 3 B 3/4 6-1 AmendmentNo.pp,pp,JJ/,g,I1E Revised by NRC letter dated
3/4.6 CONTAINMENT SYSTEMS
'e BASES 3/4.6.1.2 CONTAINMENT LEAKAGE (continued) are representative of those which would occur at accident pressure while meeting the intent of the LCO.
This test methodology is consistent with the guidance provided in ANSI /ANS 56.8-1981 for meeting the requirements set forth in Appendix J.
The surveillance testing for measuring leakage rates are consistent with' the requirements of Appendix J of 10 CFR Part 50. A partial exemption has been granted from the requirements of'10CFR50, Appendix J,Section III.D.l(a). The exemption removes the requirement that the third Type A test for each 10-year period be conducted when the plant is shut down for the 10-year plant inservice inspection (Reference License Amendment No. 111).
The enclosure building bypass leakage paths are listed in Operating Procedure 3273, " Technical Requirements - Supplementary Technical Specifica-tions."
The addition or deletion of the enclosure building bypass leakage paths shall be made in accordance with Section 50.59 of 10CFR50 and approved
[by the Plant Operation Review Committee.
l 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment i
leak rate.
Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
While the leakage rate limitation is specified at accident pressure, P., the actual surveillance testing is performed by applying a pressure greater than or equal to P,.
This higher pressure accounts for test instrument uncertainties and test volume stabilization I
changes which occurs under actual test conditions. This method of performing surveillance testing is consistent with the guidance provided in ANSI 56.8-1981 and ensures that the leakage rate measured meets the intent of the LCO and Appendi.x J.
l 3/4.6.1.4 and 3/4.6.1.5 AIR PRESSURE and AIR TEMPERATURE The limitations on containment pressure and average air temperature ensure-that:
(1) the containment structure is prevented from exceeding its design negative pressure of 8 psia, and (2) the containment peak pressure does not exceed the design pressure of 60 psia during LOCA conditions.
Neasure-ments shall be made at all listed locations, whether by fixed or portable instruments The limits on the pressure, prior to determining the average air temperature.
and average air temperature are consistent with the assumptions of the safety analysis.
The minimum total containment pressure of 10.6 psia is. determined by summing the minimum permissible air partial pressure of 8.9 psia and the maximum expected vapor pressure of 1.7 psia (occurring at the maximum permissible containment initial temperature of 120'F).
'NILLSTONE - UNIT 3 B 3/4 6-la Amendment No. J7, pp, Jpf, Ob @ @@ hy 2 0 %@ %@P M @d AUG 311
- e Millstone Nuclear Power St: tion
.UnR3 cc:
Lillian M. Cuoco, Esquire Joseph R. Egan, Esquire
- Senior Nuclear Counsel Egan & Associates, P.C.
Northeast Utilities Service Company 2300 N Street, NW P. O. Box 270 Washington, DC 20037 i
Hartford, CT 06141-0270 Mr. F. C. Rothen Mr. Kevin T. A. McCarthy, Director Vice President - Work Services Monitoring and Radiation Division Northeast Utilities Service Company Department of Environmental Protection P. O. Box 128 79 Elm Street Waterford, CT 06385 Hartford, CT 06106-5127 Emest C. Had!ey, Esquire Regional Administrator, Region i 1040 B Main Street U.S. Nuclear Regulatory Commission P.O. Box 549 475 Allendale Road West Wareham, MA 02576 King of Prussia, PA 19406 Mr. John Buckingham First Selectmen Department of Public Utility Control Town of Waterford Electric Unit Hall of Records 10 Liberty Square 200 Boston Post Road New Britain, CT 06051 Waterford, CT 06385 Mr. James S. Robinson, Manager Mr. Wayne D. Lanning Nuclear Investments and Administration Deputy Director of Inspections New England Power Company Special Projects Office 25 Research Drive 475 Allendale Road Westborough, MA 01582 King of Prussia, PA 19406-1415 Mr. John Streeter Mr. M. H. Brothers Recovery Officer - Nuclear Oversight Vice President - Millstone Unit 3 Northeast Utilities Service Company i
Northeast Nuclear Energy Company P. O. Box 128 P.O. Box 128 -
Waterford, CT 06385 Waterford, CT 06385 Deborah Katz, President Mr. M. R. Scully, Executive Director Citizens Awareness Network Connecticut Municipal Electric P.O. Box 83 Energy Cooperative Shelbume Falls, MA 03170 30 Stott Avenue Norwich, CT 06360 Mr. Atlan Johanson, Assistant Director F
Office of Policy and Management Mr. David Amerine Policy Development and Planning Vice President - Human Services Division Northeast Utilities Service Company 450 Capitol Avenue - MS# 52ERN P. O. Box 128 P. O. Box 341441 Waterford, CT 06385 Hartford, CT 06134-1441 I
-______i____________._._.____ _ _. _ _ _ _ _. _. _. _ _ _ _ _ _ _
i T
t 3
Millstone Nucle r Power Striion Unit 3 Mr. William D. Meinert Nuc; ear Engineer l
cc:
Massachusetts Municipal Wholesale The Honorable Terry Concannon E!<ctric Company Nuclear Energy Advisory Council P.O. Box 426 Room 4035 Ludlow, MA 01056 Legislative Office Building
- Capitol Avenue Attomey Nicholas J. Scobbo, Jr.
Hartford, CT 06106 Ferriter, Scobbo, Caruso, Rodophele. PC 1 Beacon Street,11th Floor Mr. Evan W. Woollacott Boston, MA 021G3 Co-Chair Nuclear Energy Advisory Council Citizens Regulatory Commission 128 Tony's Plain Road ATTN Ms. Susan Perry Luxton Simsbury, CT 03P 180 Great Neck Road Waterford, CT 06385 Mr. John W. Beck, President Little Harbor Consultants, Inc.
Millstone -ITPOP Project Office P.O. Box 0630 Niantic, CT 06357-0630 Mr. B. D. Kenyon (.\\cting)
Chief Nuclear Officer-Millstone Northaast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385 Mr. Daniel L. Curry Project Director Parsons Power Group Inc.
2675 Morgantown Road Reading, PA 19607 Mr. Don Schopfer Verification Team Manager Sargent & Lundy 55 E. Monroe Street Chicago,IL 60603 Mr. P. D. Hinnenkamp Director-Unit 3 Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385
- Senior Resident inspector Millstone Nuclear Power Station clo U.S. Nuclear Regulatory Commission i~
. P. O. Box 513
' Niantic, CT 06357 i
Ie