ML20247E959
| ML20247E959 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/08/1989 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247E962 | List: |
| References | |
| NUDOCS 8905260383 | |
| Download: ML20247E959 (14) | |
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UNITED STATES l-8 NUCLEAR REGULATORY COMMISSION o
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WASHINGTON, D. C. 20666
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VIRGINIA ELECTRIC AND POWER COMPANY OCD DOMillION ELECTRIC COOPERATIVE p0CKETNO.50-338 NORTH ANNA POWER STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 116 License No. NPF-4 I
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated October 19, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in er 'qrmity with the pplication, the provisions of the Act, and tne rules and regulations of the_ Commission; C.
There 1. reasonable assurance (1) that the activities authorized
.by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance v,' this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8905260383 890508 PDR ADOCK 05000338 P
PNU.
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. 2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.- NPF-7 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
116, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as'of the date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION erbert N. Berkow, Director Project Directorate II-2 Division of Reactor' Projects-I/II Office of Nuclear Reactor Regulation 1
Attachment:
Changes to the Technical Specifications Date of Issuance: May 8, 1989 l
1-_-_-_--_
ATTACHMENT TO LICENSE AMENDMENT NO. 116 TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 Replace the following pages of the Appendix "A" Technical Specifications.
with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Page 3/4 6-1 B3/4 6-1
___.----___-___-.---.am___m_____._.____.m.m.-
-am__
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PrimaryCONTAINNENTINTEGRITYshallbemaintained.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
1 l
Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY 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 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 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
a.
At least once per 31 days by verifying that all penetrations
- not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.3.1., and b.
By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.
c.
After each closing of the equipment hatch, by leak rate testing the equipment hatch seals with gas at Pa, greater than or-equal to 40.6 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L -
a d.
Each time containment integrity is established after vacuum has been broken by pressure testing the butterfly isolation valves in the containment purge lines and the containment vacuum ejector line.
- Except valves, blind flanges and deactivated automatic valves which are located inside the containment and are locked sealed or otherwise sealed in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.
NORTH ANNA - UNIT 1 3/4 6-1 Amendment No.116 i
\\
i CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:**
l a.
An overall integrated leakage rate of:
1.
s L, 0.1 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at$,,a44.1psig,or l
b.
A combined leakage rate of s 0.60 L, pressurized to P, a 44.1 ps for all penetrations and valves subject to Type B and C tests, when l
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L or (b) with the measured combined leakage. rate for all penetrations and, valves subject to Type B and C tests exceeding 0.60 L restore the leakage rate (s) to within the limit (s) prior to increasing t$e, Reactor Coolant System temperature above 200*F.
SURVEILLANCE RE0VIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be deters'ned in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of either ANSI N45.4-1972 for leakage rate point data. analysis or ANSI /A515-56.8-1987 for mass point data analysis with a minimum test duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.**
(
a.
Three Type A tests (Overall Integrated Containment Lehkage Rate) l shall be conducted at 40 + 10 month intervals during shutdown at Pa 2. 44.1 psig during eacii 10-year service period. The third test
[
of each set shall be conductsd during the shutdown for the 10-year plant inservice inspection.*
- The third test of the first 10-year service period shall be conducted during the 1989 Refueling /10-Year ISI Outage.
- For Specification 3/4.6.1.2 only, P shall be 40.6 psig.until completion a
of the Cycle 7 to 8 refueling outage.
Following this outage, P shall a
be 44.1 psig.
5
!! ORTH A!!NA - UNIT 1 3/4 6-2 Amendment 'lo. 105,705,110
y j
l 3/4.6 CONTAINMENT SYSTEMS 4
BASES 1
3/4. 6.1 CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY 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 accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
Leakage integrity tests in the containment purge lines and the containment vacuum ejector system lines is to identify excessive degradation of the resilient seats of these valves. These tests will be performed in addition to the Type C tests required by 10 CFR Part 50, Appendix J and will not relieve the responsibility to conform with Appendix J.
3/4. 6.1. 2 CONTAINMENT LEAKAGE The limitations on containment leakage tests ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La during performance of the periodic test to account for po:;sible degradation of the containment leakage barriers between leakage tests.
The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50. Due to the increased accuracy of the mass-point method for containment integrated leakage testing, the mass-point method referenced in ANSI /ANS 56.8-1987 can be used in lieu of the methods described in ANSI N45.4-1972.
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 I
during the intervals between air lock leakage tests.
3/4.6.1.4 and 3/4.6.1.5 INTERNAL PRESSURE AND TEMPERATURE l
The limitations on containment internal pressure and average air temperature i
ensure that
- 1) The containment pressure is prevented from reaching the containment lower design pressure of 5.5 psia for an inadvertent containment spray actuation, NORTH ANNA - UNIT 1 B 3/4 6-1 Amendment No. JpS, 116
CONTAINMENT SYSTEMS BASES 2)
That the peak clad fuel temperature will remain less than 2200*F for a LOCA and 3)
That for either a LOCA or MSLB; a)
The peak containment pressure will be limited to the upper containment design pressure of 45 psig, b)
The containment internal pressure can be returned sub-atmospheric within 60 minutes, and c)
Safety related equipment within the containment will not experience temperatures greater than those to which they have previously been qualified.
d)
It is a design criteria that the containment internal pres-sure remain subatmospheric after 60 minutes.
The limits shown in Figure 3.6-1 and Specification 3.6.1.5 are con-sistant with the assumptions of the accident analyses which included consideration of instrument loop uncertainties.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the con-tainment will be maintained comparable to the original design standards for-the life of the facility. Structural integrity is required to-ensure that the containment will withstand the design. pressure of 45 psig.
The visual examination of the concrete and liner and the Type A leakage tests are sufficient to demonstrate this capability.
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.0.2.1 and 3/4.6.2.2 CONTAINMENT QUENCH AND RECIRCULATION SPARY SYSTE'b The OPERABILITY of the containment spray systems ensures that containment depressurization and subsequent return to subatmospheric pressure will occur in the event of a LOCA. The pressure reduction and resultant termination of containment leakage are consistent with the assumptions used in the accident analyses.
l NORTH ANNA - UNIT 1 B 3/4 6-2 Amendment No. 110
prar f
UNITED STATES y
g NUCLEAR REGULATORY COMMISSION r,
j WASMNGTON, D. C. 20566
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VIRGINIA ELECTRIC AND PCWER COMPANY OCD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 99 License No. NPF-7 1.
The Nuclear Regulatory Commission (the Comission) has found that:
'l A.
The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated October 19, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon i
defense and security or to the health and safety of the public;
~
and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
I
4 2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment,
)
and paragraph 2.D.(2) of Facility Operating License No. NPF-4 is hereby amended to read as follows:
(2) Technical Specifications l
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 99, are hereby incorporated in the license. The licensee shall operate the facility in i
accordance with the Technical Specifications.
4 l
3.
This license amendment is effective as of the date of issuance and shall be implemented within 30 days.
FOR TH NUCLEAR REGULATORY COMMISSION
\\
bert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 8, 1989 1
1 A
+
1 ATTACHMENT TO LICENSE AMENDMENT NO. 99 TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 j
L Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of' change.
.l The corresponding overleaf pages are also provided to maintain document completeness.
Page 3/4 6-1 B3/4 6-1 l
n,.,
l 3/4.6 CONTAINMENT' SYSTEMS 3/4.6.1 -CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION i
3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STAND 8Y 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 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 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
a.
At least once per 31 days by verifying that all penetrations
- not-capable'of being closed by OPERABLE containment automatic isolation valves and required to'be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 3.6.3.1., and b.
By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.
c.
After each closing of the equipment hatch, by leak rate testing the equipment hatch seals with gas at Pa, greater than or equal to 40.6 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2.d for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 La.
d.
Each time containment integrity is established after vacuum has been broken by pressure testing the butterfly isolation valves in the containment purge lines and the containment vacuum ejector line.
- Except valves, blind flanges and deactivated automatic valves which are located inside the containment and are locked sealed or otherwise sealed in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.
NORTH ANNA - UNIT 2 3/4 6-1 Amendment No. 99
^
CONTAINMENT SYSTEMS
- CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:**
An overall integrated leakage' rate of:
a.
1.
Less than or equal to L,
0.1 percent by weight of the I
containment air per 24 hou!s at P,, greater than or equal to 44.1 psig, or b.
A combined leakage rate of less than or equal to 0.60 L for all penetrations and valves subject to Type B and C telts, when pressurized to P,, greater than or equal to 44.1 psig.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
)
With either (a) the measured overall integrated containment leakage rate I
exceeding 0.75 L or (b) with the measured combined leakage rate for all a
penetrations and valves subject to Type B and C tests exceeding 0.60 L restore the overall integrated leakage rate to less than 0.75 L - and t$e, combined leakage rate for all penetrations subject to Type B and (a tests to less than or equal to 0.60 L, prior to increasing the Reactor Coolant System temperature above 200'F.
SURVEILLANCE RE0VIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified'in Appendix J of 10 CFR 50 using the methods and provisions'of either ANSI' N45.4-1972 for leakage rate point data analysis or At!SI/ANS-56.8-1987 for mass point data analysis with a minimum test duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.**
.l Three Type A tests (Overall Integrated Containment Leaichge Rate) a.
shall be conducted at 40 + 10 month intervals during shutdown at Pa greater than or equal to 44.1 psig during each 10-year service l
period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection.*
- The second test of the first 10-year service period shall be conducted during the 1989 Refueling Outage.
- For. Specification 3/4.6.1.2 only, P shall be 40.5 psig until completion of the Cycle 6 to 7 refueling outage. aFollowing this outage, P shall be 44.1 psig.
a MORTH ATIA - UtlIT 2 3/4 6-2 Amendment No. 97,9A, 96
(
I 1
3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 CONTAINMENT I
3/4.6.1.1 CONTAINMENT INTEGRITY
{
l t
CONTAINMENT INTEGRITY ensures that the release of radioactive materials from j
the containment atmosphere will be restricted to those leakage paths and j
associated leak rates assumed in the accident analyses. This restriction, in l
conjunction with the leakage rate limitation, will limit the site boundary j
radiation doses to within the limits of 10 CFR 100 during accident conditions.
j l
Leakage integrity tests in the containment purge lines and the containment vacuum ejector system lines is to identify excessive degradation of the resilient seats of these valves. These tests will be performed in addition to the Type C tests required by 10 CFR Part 50, Appendix J and will not relieve the responsibility to conform with Appendix J.
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 accident analyses As an added conservatism, the measured at the peak accident pressure, P.
overall i.egrated leakage rate is further limited to less than or equal to i
0.75 La during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.
The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50.
Due to the increased accuracy of the mass-point method for containment integrated leakage testing, the mass-point method referenced in ANSI /ANS 56.8-1987 can be used in lieu of the methods described in ANSI N45.4-1972.
f 3/4.6.1.3 CONTAINMENT AIR LOCKS i
Tne limitations on closure and leak rate for the containment air locks l
are required to meet the restrictions on CONTAINMENT INTEGRITY and containment l
leak rate.
Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal I
l damage during the intervals between air lock leakage tests.
3/4.6.1.4 and 3/4.6.1.5 INTERNAL PRESSURE AND TEMPERATURE i
The limitations on containment internal pressure and average air temperature ensure that 1
j i
i NORTH ANNA - UNIT 2 B 3/4 6-1 Amendment No. H,99 l
{
4 4
4 CONTAINMENT SYSTEMS cuaEs 1)
The containment pressure is prevented from reaching the containment lower design pressure of 5.5 psia for an inadvertent containment spray actuation, 2)
That the peak clad fuel temperature will remain less than 2200*F for a.:CA and 3)
That for either a LOCA or MSLB:
a)
The peak containment pressure will be limited to the upper
. containment design pressure of 45 psig, b)
The containment internal pressure can be returned subatmospheric within 60 minutes, and c)
Safety related equipment within the containment will not experier temperatures greater than those to which they have previously been qualified.
d)
It is a design criteria that the containment internal pressure remain subatmospheric after 60 minutes.
I The limits shown in Figure 3.6-1 and Specification 3.6.1.5 are consistent with the assumptions of the accident analyses which included consideration of instru~ent loop uncertainties.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures.that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.
Structural integrity is required to ensure that the containment will withstand the design pressure of 45 psig.
The visual examination of the concrete and liner and the Type A leakage tests are sufficient to demonstrate this capability.
1 NORTH ANNA - UNIT 2 B 3/4 6-2 Amendment !!o. 96 DEC 1.41983
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