ML20106D604
| ML20106D604 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 10/08/1984 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20106D571 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 GL-83-37, NUDOCS 8410260069 | |
| Download: ML20106D604 (10) | |
Text
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' ATTACHMENT A-
'1.-
Remove pages:
3'/4-3-51, 3/4 3-52, 3/4 6-20, 6-13 2.
- Insert pages
- 13/4 3-51, 3/4 3-52, 3/4'4-32,
- B 3/4 4-l_1,. 3/4 6-20, 6-13 '
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TABLE. 3.3-11!
9
. ACCIDENT MONITORING' INSTRUMENTATION E
.1.
Pressurizer Water Level
'3 2
E
.p
'2.
Auxiliary Feedwater F1ow. Rate 1.per' steam 1 perl steam generator.-
generator 3.
Reactor Coolant System Subcooling Margin Monitor' 1
0 4.
PORV'Accoustical Detector Position Indicator' 2/ valve
- 1/ valve-3 5.
PORY Limit Switch Position Indicator.
1/ valve' 0/ valve SR 6.
PORV Block Valve Limit Switch Position Indicator
' 1/ valve 0/ valve-O*
x-Y 7.
Safety Valve Accoustical Detector Position Indicator 2/ valve
- 1/ valve SS E
8.
Safety Valve Temperature. Detector Position. Indicator 1/ valve 0/ valve E
9.
PORV Control Pressure Channels (PT-RC-444, 445) 2-1
- 10. Containment Sump Wide Range Water Level 2
1 One Detector Active, Second Detector Passive
v
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co j
TABLE 4.3-7:
m
[
' ACCIDENT MONITORING. INSTRUMENTATION SURVEILLANCE REQUIREMENTS'
~
2F
. C!
e
. CHANNEL' CHANNEL g
INSTRUMENT CHECK
' CALIBRATION 4
'1.
Pressurizer Water' Level M
R 2.
Auxiliary Feedwater Flow Rate S/U(1)'
R-3.
Reactor Coolant System Subcooling Margin M
R 4.
PORY Accoustical Detector Position Indicator M
R ow y'A
-5.
PORV Limit Switch Position. Indicator M
R' 6.
PORV Block Valve Limit Switch Position Indicator M
R' y
7.
Safety Valve Accoustical Detector Position Indicator M
R 8.
Safety Valve Temperature Detector Position Indicator M
R 9.
PORV Control Pressure Channels (PT-RC-444, 445)
M R
- 10. Containment Sump Wide Range Water Level M
R j
(1) Channel check to be performed in conjunction with Surveillance Requirement 4.7.1.2.a.9 following an extended plant outage.
I
-REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.12 ' At least one' reactor coolant system vent path consisting of at least two valves in series powered from emergency buses shall be OPERABLE
- and closed **
at each of the following locations:
a.
Reactor Vessel head b.
Pressurizer. steam space c.
Pres:urizer Relief Tank inlet suppiy d.
Containment supply APPLICABILITY:
MODES 1, 2, 3 and 4 ACTION:
a.
With one of the above reactor coolant system vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed or isolated with power removed from the inoperable valve.
Power operation may continue until the next scheduled outage, at which time all vent paths shall be OPERABLE prior to entry into Mode 1.
The provisions of Specification 3.0.4 are not applicable.
b.
With both Train-A and Train B vent paths inoperable; maintain the inoperable vent paths closed or isolated with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at least one of the inoperable vent paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT-STANDBY within 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 4.4.12 Each' reactor. coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by:
1.
Verifying all manual isolation valves in each vent path are locked or sealed in the open position.
'2.
Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING.
3.
Verifying flow through the reactor coolant vent system vent path to the i
Pressurizer Relief Tank during venting during COLD SHUTDOWN or REFUELING.
For purposes of this specification an inoperable vent valve is defined as; a valve which exhibits excessive leakage, or cannot be opened and closed on demand, or does not have its normal emergency power supply 0PERABLE.
i
- The valves may be operated for venting operations and leak testing.
BEAVER VALLEY - UNIT 1 3/4 4-32 PROPOSED WORDING
REACTOR CCOLANT SYSTDI
, BASES 3/4.4.12 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or. steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head, the pressurizer steam space, the pressurizer relief tank inlet supply, and the containment supply ensures the capability exists to perform this function.
The valve redundancy of the reactor coolant system vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the reactor coolant system vent systems are consistent with the requirenents of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements", flovember 1980.
f L
l BEAVER VALLEY - UtlIT 1 B 3/4 4-11 l
PROPOSED WORDING
CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HJDR0 GEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two separate and independent wide range containment hydrogen analyzers j
she.11 be OPERABLE.
APPLICABILITY: MODES 1 and 2.
ACTION:~
With one wide range hydrogen analyzer inoperable, restore the a.
inoperable analyzer to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.
The provisions of Specification 3.0.4 are not applicable.
l SURVEILLANCE REQUIREMENTS 4.6.4.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by:
Performing a CHANNEL CALIBRATION using sample gases containing:
a.
1.
One volume percent hydrogen, balance nitrogen, and 2.
Four volume percent hydrogen, balance nitrogen.
l BEAVER VALLEY - UNIT 1 3/4 6-20 PROPOSED WORDING i--
ADMINISTRATIVE CONTROLS
.6.8.3 Temporary changes to procedures of 6.8.1 abovs may be made provided:
a.
The intent of the original procedure is not altered.
b.
The change is approved by two (2) members of the plant management staff, at least one (1) of whom holds a Senior Reactor Operator's License on the unit affected.
c.
The change is documented, reviewed by the OSC and approved by the Plant Superintendent within 14 days of implementation.
6.8.4 A Post-Accident monitoring program shall be established, implemented, and maintained:
A program which will provide the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples following an accident.
The program shall include the following:
(i)
Training of personnel, (ii)
Procedures for sampling and analysis, and (iii)
Provisions for maintenance of sampling and analysis equipment.
6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Pegional Office of Inspection and Enforcement unless otherwise noted.
STARTUP REPORTS 6.9.1.1 A summary report of plant startup and power escalation testing will be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
6.9.1.2 The startup report shil address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.
Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details requested in license conditions based-on other commitments shall be included in this report.
6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.
If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.
BEAVER VALLEY - UNIT 1 6-13 PROPOSED WORDING s
r ATTACHMENT B Safety Evaluation Proposed Change Request No. 94 amends the Beaver Valley Power Station, Unit I
No.1 Technical Specifications, Appendix A by incorporating the applicable I
NUREG-0737 requirements specified by Generic Letter 83-37, t
j Description and Purpose of Change 1.
'page 3/4 3-51, Table 3.3-11 Accident Monitoring Instrumentation has been revised by adding instrument No.
9, PORV Control Pressure Channels (PT-RC-444,445) and instrument No. 10, Containment Sump Wide Range Water I
Level.
Instrument No. 9 was added to this table to be consistent with the change to Table 4.3-7 incorporated by Amendment No. 45.
Instrument No. 10 was added to comply with the requirements of NUREG-0737 by completing the list of accident monitoring instrumentation.
2.
page 3/4 3-52, Table 4.3-7 Accident Monitoring Instrumentation Surveillance Requirements has been revised by adding instrument No. 10, Containment Sump Wide-Range Water Level to reflect the addition to Table 3.3-11.
Note (2) has been deleted since the channel calibration referenced has been completed and this note is no longer applicable.
3.
page 3/4 4-32, Section 3.4.12 Reactor Coolant System Vents has been added to specify the limiting conditions for operation and associated surveillance requirements for the reactor coolant vest system.
4.
page B 3/4 4-11, Bases Section 3/4.4.12 Reactor Coolant System Vents has been added to provide the basis for the above specification.
5.
page 3/4 6-20, Section 3.6.4.1 Hydrogen Analyzers has been revised to apply specifically to the wide range hydrogen analyzers installed to meet the requirements of NUREG-0737. An exception to specification 3.0.4 has been added to the Action statement to permit the plant to change modes when one hydrogen analyzer is inoperable.
6.
page 6-13, Administrative Control Section 6.8.4 has been added to require a post-accident monitoring program be established, implemented and maintained to meet the reauirements of NUREG-0737.
Basis for Proposed No Significant Hazards Consideration Determination The proposed changes impose new requirements on portions of systems previously not governed by the Technical Specifications and to reflect additional administrative controls and plant modifications implemented to comply with NUREG-0737 for mitigating the consequences of an accident.
The Commission has provided guidance concerning the application of these standards by providing certain examples (48 FR 14870). One of these, Example (ii), involving no significant hazards consideration is "A change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications."
The new requirements match this example. Therefore, based on the above example, it is proposed that the change be characterized as involving no significant hazards consideration.
m
L Safzty Evaluation 1A-94
[
Page 2 Basis 1.
Is the probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated in the UFSAR-increased?
N_o.
o Reason 1.
Table 3.3-11 and Table 4.3-7 were revised by adding an additional instrument, the Containment Sump Wide Range Water Level. This will not increase the probability of an accident occurring since the instrument only provides indication of sump level and this change only adds them to the table. -The addition of this instrument to the tables reflects the equipment presently installed to comply with NUREG-0737 ana is consistent with UFSAR Section 7.3.1.3.1.
2.
The Reactor Coolant Vent System was installed to comply with the requirements of NUREG-0737, Item II.B.1.
The proposed specification will provide additional assurance that the vent paths from the reactor
-vessel and/or the pressurizer are available during and following an accident to vent noncondensible gases from the RCS. The system design is not required to meet single failure criteria, therefore, no immediate action toward a plant shutdown is required unless both trains of one vent path are inoperable. The changes are consistent with the RCS vent description in UFSAR Section 4.2.11.
3.
The Hydrogen Analyzer specification has been revised to apply to the containment wide-range hydrogen analyzers since the wide-range analyzers were installed to comply with the requirements of NUREG-0737.
The exception to specification 3.0.4 added to the Action statement will allow mode changes since the operability of.the hydrogen analyzers is not extremely dependent on the plant operating mode; they are only required for post accident use. These changes are consistent with the UFSAR Section 6.5.2 description of the containment hydrogen analyzers.
4.
The addition of Administrative Controls Section 6.8.4, Post Accident Monitoring Program is an administrative change requiring that a program be established, implemented and maintained.
This change complies with the requirements of NUREG-0737 and is consistent with UFSAR Section 9.6.
2.
Is the probability for an accident or malfunction of a different type than previously evaluated in the UFSAR created? No Reason 1.
The changes to the Accident Monitoring Instrumentation Tables provide additional assurance that the equipment will be operable during accident conditions, therefore, no new accident will be created.
2.
The Reactor Coolant Vent System piping is small enough so that its loss would constitute a small break LOCA, for which the plant is already analyzed, therefore, no new accidents will be created.
3.
The changes to the hydrogen analyzer specification will not create a new accident, since the UFSAR currently addresses the loss of the hydrogen analyzers while operating.
- Saf;ty Evaluati:n 1A-94
,i Page 3 4.
Incorporating the Post Accident Sampling Program requirements is an administrative change and does not affect the UFSAR accident analyses.
3.
Is the margin of safety as defined in the basis for any technical specification reduced? N,{o Reason 1.
The addition of the containment sump wide-range water level instrument to the accident monitoring tables complies with the requirements of NUREG-0737 and will increase the margin of safety by providing additional information to plant operators during accident conditions.
2.
An additional bases section has been added to provide the bases for the RCS vent specification and the margin of safety for this system is consistent with the margin of safety provided for the other specifications.
3.
The hydrogen analyzers are used during accident conditions regardless of whether or not the plant is in Mode 1 or 2, therefore, the margin of safety for this system will not be reduced.
4.
The addition of the Post-Accident Sampling System to the Administrative Controls section is an administrative change and will not affect the basis for any technical specification.
4.
Based on the above, is an unreviewed safety question involved? fio 5.
Is a change to the UFSAR required? No Conclusion The proposed changes are administrative in nature since they reflect administrative and plant component changes installed as required by NUREG-0737.
These changes do not involve additional physical changes to plant safety-related systems, components or structures, will not increase the likelihood of a malfunction of safety related equipment, increase the consequence of an accident previously analyzed, nor create the possibility of a malfunction different than previously evaluated in the UFSAR.
Based on the considerations addressed above, the proposed revision's have been determined to be safe and do not involve an unreviewed safety question.
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