ML20236W020

Proposed Tech Specs Bases Section 3/4.6.1.1,clarifying Administrative Controls for RHR Isolation Valves When RHR Sys Is in Svc for Core Cooling

ML20236W020
Person / Time
Site:	Millstone
Issue date:	07/30/1998
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20236W014	List: B17046, Application for Amend to License NPF-49,clarifying Administrative Controls for RHR Isolation Valves When RHR Sys Is in Svc for Core Cooling ML20236W020
References
NUDOCS 9808050030
Download: ML20236W020 (6)

Text

.._-__

Docket No. 50-423 B17046-Millstone. Nuclear Power Station, Unit No. 3 Marked Up Technical Specification Page July 1998 9909030030 990730 "

-yDR ADOCK 05000423 PDR}a.

___m___.____._m

_m_ _. __...._

12/185-i i

3/4.5 CONTAINMENT SYSTENS 1

BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 -CONTAINMENT INTEGRITY Primary CONTAINMENT kNTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be rastricted to those leakage paths and associated leak rates assumed in the safety analyses. This restric-tion, in conjunction with the leakage rate limitation, will'linit the SITE m7.'

BOUNDARY radiation doses to within the dose guidelines of 10 CFR Part 100 during accident conditions and.the control room operators dose to within the guidelines of GDC 19.

The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative' control includes the following i

considerations:

(1) stationing an operator, who is in constant communication

.with control room, at the valve controls, (2) instructing this operator to close these valves in an a'ccident situation,.and (3) assuring that environmental cond. ions will not preclude access to clost the valves and that this action will prevent the release of radioactivity outside the cp tainment.

1

\\

INrfK T 3/4.6.1.2 CONTAINMENT LEAKAGE 1

The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the~ peak accident pressure, P,.

As an added conservatism, the i-measured overall integrated leakage rate is further limited to less than 0.75 L 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 lea' kage ' rates for compliance with 10CFR50, Appendix J.

The leakage rates are verified byJ 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.

j Conseqtiently, the surveillance. testing is performed at a pressure greater than i

or ' equal to P, to account for test instrument uncertainties and stabilization changes. This conservative test pressure ensures that the measi!*ed leakage rates I

are representative of those which wou?d occur at acci&nt pressure while meeting 1

the intent of the LC0.. This test methodology is corisistent with the guidance provided in ANSI /ANS 56.8-1981' for meeting the requirements set. forth in Appendix J.

l 3

1 The surveillance testing for measuring leakage rates are cor.,istent with 1

' the requirements of Appendix J of 10 CFR Part 50. ' A partial exemption'has been j

granted from the requirements of.10CF!!50, Appeadix J.Section III.D.1(a).. The exemption removes the requirement that the third Type A test for each 10-year V, ) -

period be conducted when the plant is shut ~down for the 10-year plant inservice I

inspection '(Reference License Amendment No.111).

4 MILLSTONEL-tlNIT 3l B 3/4.6-1 Amendment No. J7, 77, JJJ,154 f',

st I

MCRAT When the Residual Heat Removal (RHR) System is placed in service in the plant cooldown mode of operation, the PJiR suction isolation remotely operated valves 3RHS*MV8701A and 3RHS*MV8701B, and/or 3RHS*MV8702A and 3RHS*MV8702B 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 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 a separate' dedicated operator.

1 L

Docket No. 50-423 B17046 i

i

l l

1 l

l 1

Millstone Nuclear Power Station, Unit No. 3 1

i i

Retyped Technical Specification Page j

l I

l l

l 1

1 i

July 1998 l

l.b-e _- _ _ _ _ _ _ _. _

I

~.

i 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 assochted leak rates assumed in the safety analyses. This restric-tion, in conjunction with the leakage rate limitation, will limit the SITE B0UNDARY radiation doses to within the dose guidelines of 10 CFR Part 100 during accident conditions and the control room operators dose to within the guidelines of GDC 19.

j 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 tk valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will 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*MV8702B 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 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 a separate dedicated operator.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total 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 i

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.

[

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 l

l NILLSTONE - UNIT 3 B 3/4 6-1 Amendment No. 57, 77, JJJ, JJJ.

(

ma f.

, 3/4.6 CONTAINMENT SYSTEMS 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 methoda ugy 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.1(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.

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 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 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 LC0 and Appendix J.

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.

Measure-ments shall be made at all listed locations, whether by. fixed or portable instruments, prior to determining the average air temperature.

The limits on the pressure 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).

MILLSTONE - UNIT 3 8 3/4 6-la Amendment No. 77, pp, #7, 0692 L_-_-_-__-_-_____________-____-______-____-____-______--_-__-__

_ _ _ - - _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - - _ -