Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability" and Generic Issue 94, "Additional Low-Temperature Over Pressure Protection for Light-Water Reactors" Pursuant to 10 CFR 50.54(f) (Generic Letter 90-06)
June 25, 1990
TO: ALL PRESSURIZED WATER REACTOR LICENSEES AND CONSTRUCTION PERMIT
HOLDERS
SUBJECT: RESOLUTION OF GENERIC ISSUE 70, "POWER-OPERATED RELIEF
VALVE AND BLOCK VALVE RELIABILITY," AND GENERIC ISSUE 94,
"ADDITIONAL LOW-TEMPERATURE OVERPRESSURE PROTECTION FOR
LIGHT-WATER REACTORS," PURSUANT TO 10 CFR 50.54(f)
(GENERIC LETTER 90-06)
The purpose of this generic letter is to advise pressurized water reactor
(PWR) licensees and construction permit (CP) holders of the staff positions
delineated in Enclosures A and B to this letter. Enclosure A presents the
staff position resulting from the resolution of Generic Issue 70 (GI-70) and
is applicable to all Westinghouse and Babcock and Wilcox (B&W)-designed
plants and Combustion Engineering (CE)-designed plants with power-operated
relief valves (PORVs). Enclosure B presents the staff position resulting
from the resolution of Generic Issue 94 (GI-94) and is applicable to all
Westinghouse-designed and CE-designed plants whether or not they have PORVs
and block valves. Enclosure B does not apply to B&W-designed plants.
The technical findings and the regulatory analysis related to GI-70 are
discussed in NUREG-1316, "Technical Findings and Regulatory Analysis Related
to Generic Issue 70--Evaluation of Power-Operated Relief Valve and Block
Valve Reliability in PWR Nuclear Power Plants" (Enclosure C). In Enclosure
D, the staff prepared a regulatory analysis for GI-94 based on the work
performed by Battelle Pacific Northwest Laboratory (PNL) and reported in
NUREG-1326, "Regulatory Analysis for the Resolution of Generic Issue 94,
Additional Low-Temperature Overpressure Protection for Light-Water
Reactors."
On the basis of technical studies for GI-70, the staff requests that to
enhance safety, actions identified in Section 3 of Enclosure A be taken by
all PWR licensees and CP holders that use or could use PORVs to perform any
of the safety-related functions identified in Section 2 of Enclosure A.
These actions result from the staff interpretation of safety-related
equipment (see 10 CFR 50.49 and 10 CFR Part 100, Appendix A).
On the basis of technical studies for GI-94, the staff also requests that to
enhance safety, actions identified in Section 3 of Enclosure B be taken by
all Combustion Engineering and Westinghouse PWR licensees and CP holders.
These actions result from the staff interpretation of General Design
Criteria 15 and 31 in 10 CFR Part 50, Appendix A. The information requested
by this letter is directed at addressing these concerns.
Note that the staff's requests are based on the performance of PORVs and
PORV block valve designs used to date on U.S. power reactors. Currently,
certain valve manufacturers are developing modified designs with the goal of
improving reliability. The use of more reliable valves should result in
less frequent corrective maintenance and can result in longer inservice
testing intervals as delineated in Section XI of the ASME Boiler and
Pressure Vessel Code.
9006200120
.
Generic Letter 90-06 - 2 -
Accordingly, pursuant to Section 182 of the Atomic Energy Act and
10 CFR 50.54(f), you, as a PWR licensee or CP holder, are required to advise
the NRC staff under oath or affirmation, within 180 days of the date of this
letter, of your current plans relating to PORVs and block valves and to
low-temperature over-pressure protection, in particular whether you intend to
follow the staff positions included in Enclosures A and B as applicable, or
propose alternative measures, and your proposed schedule for implementation.
For PWR plants with an operating license, staff positions 1 and 2 in Section
3.1 of Enclosure A should be implemented by the end of the first refueling
outage that starts 6 months or later from the date of this letter. Requests
for the technical specification modifications in staff position 3 in Section
3.1 of Enclosure A and in Section 3 of Enclosure B should be submitted by
the end of the first refueling outage that starts 6 months or later from the
date of this letter.
For PWR CP holders, staff positions 1 and 2 in Section 3.1 of Enclosure A
should be implemented before initial criticality or within 6 months of the
date of this letter, whichever is later. The technical specification
modifications in staff position 3 in Section 3.1 of Enclosure A and in
Section 3 of Enclosure B should be submitted by the end of the first
refueling outage that starts 6 months or later from the date of this letter.
If the applicable schedule cannot be met, the licensee or the CP holder
shall advise the staff of a proposed revised schedule, justification for any
delay, and any planned compensating measures to be taken during the interim.
Alternatives to schedules and the guidance provided herein will be evaluated
on their merits on an individual case basis. Based on its review and the
acceptability of these responses, the staff will close out GI-70 and GI-94
for each plant.
Your response shall include the following specific items.
1. A statement by licensees and CP holders as to whether they will commit
to incorporate improvements 1, 2, and 3 in Section 3.1 of Enclosure A.
With respect to improvement 3 in Section 3.1 of Enclosure A, licensees
and CP holders shall state whether they will commit to use those
modified limiting condi-tions of operation of PORVs and block valves in
the technical specifica-tions for Modes 1, 2, and 3 in Attachment A-1 of
Enclosure A for Westinghouse-designed and CE-designed plants with two
PORVs, or in Attachment A-2 of Enclosure A for Westinghouse-designed
plants with three PORVs, or in Attachment A-4 of Enclosure A for
B&W-designed plants.(1) In addition to this 10 CFR 50.54(f) request,
if the licensees and the CP holders commit to implement these
recommended technical specifications, it is requested that they submit
modifications to their current technical specifications in a license
amendment in accordance with the schedule noted above.
(1) Plants that already have staff-issued technical specifications
consistent with these requirements need merely state this in their response.
No further action will be required for this aspect of the Commission's
position.
.
Generic Letter 90-06 - 3 -
2. A statement by licensees and CP holders as to whether they will submit
a license amendment request to modify the technical specifications and
commit to use the modified technical specifications for the
low-temperature overpressure protection system concerning the limiting
conditions of operation in Modes 5 and 6 as identified in Attachment
B-1 of Enclosure B to this generic letter for Westinghouse-designed or
CE-designed plants, as appropriate. In addition to this
10 CFR 50.54(f) request, if the licensees and CP holders commit to
implement these recommended technical specifications, it is requested
that they submit modifications to their current technical
specifications in a license amendment in accordance with the schedule
noted above.
The actions to incorporate technical specification (TS) requirements for the
resolution of GI-70 and GI-94 are considered to be consistent with the
Commission's Policy Statement on Technical Specification Improvements. This
policy statement captures existing requirements under Criterion 3
(Mitigation of Design-Basis Accidents or Transients) or under the provisions
to retain requirements that operating experience and probabilistic risk
assessment show to be important to the public health and safety. Although
it is recognized that PORVs for older plants may not have been classified as
safety-related components that are used to mitigate a design-basis accident
and, therefore, may not have been included in TS as part of the plant's
licensing basis, this is not an acceptable basis for not implementing the
proposed actions to incorporate TS requirements for PORVs consistent with
the guidance provided. Likewise, such requirements would be retained in TS
when implementing improvements in TS consistent with the Commission policy
statement on the basis of Criterion 3 or risk considerations noted above.
Backfit Discussion
For GI-70, the actions proposed by the NRC staff to improve the reliability
of PORVs and block valves, as identified in Section 3 of Enclosure A,
represent new staff positions for some licensees and CP holders, and this
request is considered a backfit in accordance with NRC procedures. This
backfit is a cost-justified safety enhancement. Therefore, an analysis of
the type described in 10 CFR 50.109(a)(3) and 50.109(c) was performed, and a
determination was made that there will be a substantial increase in overall
protection of the public health and safety and that the attendant costs are
justified in view of this increased protection. The analysis and
determination will be made available in the Public Document Room with the
minutes of the 167th and 168th meetings of the Committee to Review Generic
Requirements.
It is noted that most of the recommended actions for GI-70 may already be
implemented by those plants that have received operating licenses in recent
years and would, therefore, represent less of a backfit than for older PWR
plants that currently do not include PORVs and block valves in the ASME
Section XI Inservice Testing Program and do not have technical
specifications for PORVs and block valves or that operate with the block
valves closed because of leaking PORVs.
.
Generic Letter 90-06 - 4 -
For GI-94, the actions proposed by the NRC staff to improve the availability
of the low-temperature overpressure protection (Ltop) system, as identified
in Section 3 of Enclosure B, represent a new interpretation of existing
requirements for some licensees and CP holders, and this request is
considered a backfit in accordance with NRC procedures. This backfit is a
cost-justified safety enhancement. Therefore, an analysis of the type
described in 10 CFR 50.109(a)(3) and 50.109(c) was performed, and a
determination was made that there will be a substantial increase in overall
protection of the public health and safety and that the attendant costs are
justified in view of this increased protection. The analysis and
determination will be made available in the Public Document Room with the
minutes of the 167th and 168th meetings of the Committee to Review Generic
Requirements.
This request is covered by Office of Management and Budget Clearance Number
3150-0011, which expires January 31, 1991. The estimated average burden
hours is 320 person-hours per licensee response, including assessment of the
new recommendations, searching data sources, gathering and analyzing the
data, and preparing the required reports. Comments on the accuracy of this
estimate and suggestions to reduce the burden may be directed to the Office
of Management and Budget, Room 3208, New Executive Office Building,
Washington, D.C. 20503, and the U.S. Nuclear Regulatory Commission,
Information and Records Management Branch, Office of Information Resources
Management, Washington, D.C. 20555.
Sincerely,
James G. Partlow
Associate Director for Projects
Office of Nuclear Reactor Regulation
Technical Contact: George A. Schwenk
(301) 492-0878
Enclosures:
A. Staff Positions Resulting from Resolution of Generic Issue 70
B. Staff Positions Resulting from Resolution of Generic Issue 94
C. NUREG-1316, "Technical Findings and Regulatory Analysis Related
to Generic Issue 70--Evaluation of Power-Operated Relief Valve and Block
Valve Reliability in PWR Nuclear Power Plants"
D. NUREG-1326, "Regulatory Analysis for the Resolution of Generic
Issue 94, Additional Low-Temperature Overpressure Protection for Light-
Water Reactors"
.
Enclosure A to Generic Letter 90-06
Staff Positions Resulting from
Resolution of Generic Issue 70 -
PORV and Block Valve Reliability
1. BACKGROUND
Generic Issue 70 (GI-70), "Power-Operated Relief Valve and Block Valve
Reliability," involves the evaluation of the reliability of power-operated
relief valves (PORVs) and block valves and their safety significance in PWR
plants. The technical findings and regulatory analysis related to GI-70 are
discussed in NUREG-1316, "Technical Findings and Regulatory Analysis Related
to Generic Issue 70--Evaluation of Power-Operated Relief Valve and Block
Valve Reliability in PWR Nuclear Power Plants" (Enclosure C). This report
identifies those safety-related functions that may be performed by PORVs and
also identifies potential improvements to PORVs and block valves. In
support of the resolution of GI-70, the Oak Ridge National Laboratory (ORNL)
performed a study of PORV and block valve operating experience. A report,
prepared by ORNL, was issued as NUREG/CR-4692, "Operating Experience Review
of Failures of Power Operated Relief Valves and Block Valves in Nuclear
Power Plants," dated October 1987.
Traditionally, the PORV and its block valve are provided for plant
operational flexibility and for limiting the number of challenges to the
safety-related pressurizer safety valves. The operation of the PORVs has
not been classified as a safety-related function, i.e., one on which the
results and conclusions of the safety analysis are based and that invokes
the highest level of quality and construction. For overpressure protection
of the reactor coolant pressure boundary (RCPB) at normal operating
temperature and pressure, the operation of PORVs has not been explicitly
considered as a safety-related function. Also, an inadvertent opening of a
PORV or safety valve has been analyzed in the Final Safety Analysis Reports
as an anticipated operational occurrence with acceptable consequences. For
these reasons, most PWRs, particularly those licensed prior to 1979, do not
classify PORVs as safety-related components.
The Three Mile Island Unit 2 (TMI-2) accident focused attention on the
reliability of PORVs and block valves since the malfunction of the PORV at
TMI-2 contributed to the severity of the accident. On other occasions,
PORVs have stuck open when called upon to function. Also, there are PORVs
in many operating plants that have leakage problems so that the plants must
be operated with the upstream block valves in the closed position. The
technical specifications governing PORVs on most operating PWRs, which deal
with closing the block valve and removing power, were developed to allow
continued plant operation with degraded PORVs, but did not consider the need
for the PORVs to perform the safety functions discussed below.
Following the TMI-2 accident, the staff began to examine transient and
accident events in more detail, particularly with respect to required
operator actions and equipment availability and performance. As a result,
the staff initiated an evaluation of the role of PORVs to perform certain
safety-related functions.
.
A-2
2. SAFETY FUNCTIONS OF PORVs AND BLOCK VALVES
The staff, in its evaluation, determined that over a period of time the role
of PORVs has changed such that PORVs are now relied upon by many
Westinghouse, B&W, and CE designed plants with PORVs to perform one, or
more, of the following safety-related functions:
1. Mitigation of a design-basis steam generator tube rupture
accident,
2. Low-temperature overpressure protection of the reactor vessel
during startup and shutdown, or
3. Plant cooldown in compliance with Branch Technical Position RSB
5-1 to SRP 5.4.7, "Residual Heat Removal (RHR) System."
Where PORVs are used or could be used to perform one, or more, of the
safety-related functions identified above or to perform any other
safety-related function that may be identified in the future, it is
appropriate to reconsider the safety classification of PORVs and the
associated block valves. For certain PWR plants receiving an operating
license in recent years, the staff has required these valves to be
classified as safety-related components if they perform one, or more,
safety-related functions.
For operating PWR plants, the staff has concluded that it is not cost
effective to replace (backfit) existing non-safety-grade PORVs and block
valves (and associated control systems) with PORVs and block valves that are
safety grade even when they have been determined to perform any of the
safety-related functions discussed above. Subsequent to the TMI-2 accident,
a number of improvements were required of PORVs and block valves, such as
requirements to be powered from Class IE buses and to have valve position
indication in the control room. For operating plants, the greatest
immediate benefits can be derived from implementing items 1 through 3
identified below, which can increase the reliability of these components and
provide assurance they will function as required.
3. IMPROVEMENTS TO ALL PORVs AND BLOCK VALVES
3.1 Operating PWR Plants and Construction Permit Holders
Based on the analysis and findings for GI-70, the staff concludes that the
following actions should be taken to improve the reliability of PORVs and
block valves:
1. Include PORVs and block valves within the scope of an operational
quality assurance program that is in compliance with 10 CFR Part
50, Appendix B. This program should include the following
elements:
a. The addition of PORVs and block valves to the plant
operational Quality Assurance List.
b. Implementation of a maintenance/refurbishment program for
PORVs and block valves that is based on the manufacturer's
recommendations
.
A-3
or guidelines and is implemented by trained plant maintenance
personnel.
c. When replacement parts and spares, as well as complete
components, are required for existing non-safety-grade PORVs
and block valves (and associated control systems), it is the
intent of this generic letter that these items may be
procured in accordance with the original construction codes
and standards.
2. Include PORVs, valves in PORV control air systems, and block
valves within the scope of a program covered by Subsection IWV,
"Inservice Testing of Valves in Nuclear Power Plants," of Section
XI of the ASME Boiler and Pressure Vessel Code. Stroke testing of
PORVs should only be performed during Mode 3 (HOT STANDBY) or Mode
4 (HOT SHUTDOWN) and in all cases prior to establishing conditions
where the PORVs are used for low-temperature overpressure
protection. Stroke testing of the PORVs should not be performed
during power operation. Additionally, the PORV block valves
should be included in the licensees' expanded MOV test program
discussed in NRC Generic Letter 89-10, "Safety-Related Motor
Operated Valve Testing and Surveillance," dated June 28, 1989.
3. For operating PWR plants, modify the limiting conditions of
operation of PORVs and block valves in the technical
specifications for Modes 1, 2, and 3 to incorporate the position
adopted by the staff in recent licensing actions. Attachments A-1
through A-3 are provided for guidance. The staff recognizes that
some recently licensed PWR plants already have technical
specifications in accordance with the staff position. Such plants
are already in compliance with this position and need merely state
that in their response. These recent technical specifications
require that plants that run with the block valves closed (e.g.,
due to leaking PORVs) maintain electrical power to the block
valves so they can be readily opened from the control room upon
demand. Additionally, plant operation in Modes 1, 2, and 3 with
PORVs and block valves inoperable for reasons other than seat
leakage is not permitted for periods of more than 72 hours.
.
A-4 Generic Issue 70
Enclosure A to Generic Letter 90-06
Attachment A-1
Modified Standard Technical Specifications
for Combustion Engineering and Westinghouse Plants
REACTOR COOLANT SYSTEM
3/4.4.4 RELIEF VALVES
LIMITING CONDITION FOR OPERATION
The following is to be used when two PORVs are provided:
3.4.4 Both power-operated relief valves (PORVs) and their associated block
valves shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
a. With one or both PORVs inoperable because of excessive seat
leakage, within 1 hour either restore the PORV(s) to OPERABLE
status or close the associated block valve(s) with power
maintained to the block valve(s); otherwise, be in at least HOT
STANDBY within the next 6 hours and in HOT SHUTDOWN within the
following 6 hours.
b. With one PORV inoperable due to causes other than excessive seat
leakage, within 1 hour either restore the PORV to OPERABLE status
or close its associated block valve and remove power from the
block valve; restore the PORV to OPERABLE status within the
following 72 hours or be in HOT STANDBY within the next 6 hours
and in HOT SHUTDOWN within the following 6 hours.
c. With both PORVs inoperable due to causes other than excessive seat
leakage, within 1 hour either restore at least one PORV to
OPERABLE status or close its associated block valve and remove
power from the block valve and be in HOT STANDBY within the next 6
hours and in HOT SHUTDOWN within the following 6 hours.
d. With one or both block valves inoperable, within 1 hour restore
the block valve(s) to OPERABLE status or place its associated
PORV(s) in manual control. Restore at least one block valve to
OPERABLE status within the next hour if both block valves are
inoperable; restore any remaining inoperable block valve to
operable status within 72 hours; otherwise, be in at least HOT
STANDBY within the next 6 hours and in HOT SHUTDOWN within the
following 6 hours.
.
A-5 Generic Issue 70
e. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV
shall be demonstrated OPERABLE at least once per 18 months by:
a. Operating the PORV through one complete cycle of full travel
during MODES 3 or 4, and
.
A-6 Generic Issue 70
Enclosure A To Generic Letter 90-06
Attachment A-2
Modified Standard Technical Specifications
for Westinghouse Plants with Three PORVs
REACTOR COOLANT SYSTEM
3/4.4.4 RELIEF VALVES
LIMITING CONDITION FOR OPERATION
The following is to be used when three PORVs are provided:
3.4.4 All power-operated relief valves (PORVs) and their associated block
valves sha11 be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
a. With one or more PORVs inoperable because of excessive seat
leakage, within 1 hour either restore the PORV(s) to OPERABLE
status or close the associated block valve(s) with power
maintained to the block valve(s); otherwise, be in at least HOT
STANDBY within the next 6 hours and HOT SHUTDOWN within the
following 6 hours.
b. With one or two PORVs inoperable due to causes other than
excessive seat leakage, within 1 hour either restore the PORV(s)
to OPERABLE status or close the associated block valve(s) and
remove power from the block valve(s); restore the PORV(s) to
OPERABLE status within the following 72 hours or be in HOT STANDBY
within the next 6 hours and in HOT SHUTDOWN within the following 6
hours.
c. With three PORVs inoperable due to causes other than excessive
seat leakage, within 1 hour either restore at least one PORV to
OPERABLE status or close the block valves and remove power from
the block valve(s) and be in HOT STANDBY within the next 6 hours
and in HOT SHUTDOWN within the following 6 hours.
d. With one or more block valves inoperable, within 1 hour restore
the block valve(s) to OPERABLE status or place its associated PORV
in manual control. Restore at least one block valve to OPERABLE
status within the next hour if three block valves are inoperable;
restore any remaining inoperable block valve(s) to operable status
within 72 hours; otherwise, be in HOT STANDBY within the next 6
hours and in HOT SHUTDOWN within the following 6 hours.
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A-7 Generic Issue 70
e. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV
shall be demonstrated OPERABLE at least once per 18 months by:
a. Operating the PORV through one complete cycle of full travel
during MODES 3 or 4, and
b. Where applicable, operating solenoid air control valves and check
valves on associated air accumulators in PORV control systems
through one complete cycle of full travel for plants with
air-operated PORVs, and
c. Performing a CHANNEL CALIBRATION of the actuation instrumentation.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92
days by operating the valve through one complete cycle of full travel unless
the block valve is closed in order to meet the requirements of ACTION b, or
c in Specification 3.4.4.
4.4.4.3 The emergency power supply for the PORVs and block valves shall be
demonstrated OPERABLE at least once per 18 months by:
a. Manually transferring motive and control power from the normal to
the emergency power bus, and
b. Operating the valves through a complete cycle of full travel.
WESTINGHOUSE PLANTS
.
A-8 Generic Issue 70
Enclosure A to Generic Letter 90-06
Attachment A-3
Applicable to Combustion Engineering and Westinghouse Plants
3/4.4.4 RELIEF VALVES
Bases of the Limiting Condition for Operation (LCO) and Surveillance
Requirements:
The OPERABILITY of the PORVs and block valves is determined on the basis of
their being capable of performing the following functions:
A. Manual control of PORVs to control reactor coolant system pressure.
This is a function that is used for the steam generator tube rupture
accident and for plant shutdown. This function has been classified as
safety related for more recent plant designs.
B. Maintaining the integrity of the reactor coolant pressure boundary.
This is a function that is related to controlling identified leakage
and ensuring the ability to detect unidentified reactor coolant
pressure boundary leakage.
C. Manual control of the block valve to: (1) unblock an isolated PORV to
allow it to be used for manual control of reactor coolant system
pressure (Item A), and (2) isolate a PORV with excessive seat leakage
(Item B).
D. Automatic control of PORVs to control reactor coolant system pressure.
This is a function that reduces challenges to the code safety valves
for overpressurization events.
E. Manual control of a block valve to isolate a stuck-open PORV.
Surveillance Requirements provide the assurance that the PORVs and block
valves can perform their functions. Specification 4.4.4.1 addresses PORVs,
4.4.4.2 the block valves, and 4.4.4.3 the emergency (backup) power sources.
The latter are provided for either PORVs or block valves, generally as a
consequence of the TMI ACTION requirements to upgrade the operability of
PORVs and block valves, where they are installed with non-safety-grade power
sources, including instrument air, and are provided with a backup
(emergency) power source. The block valves are exempt from the surveillance
requirements to cycle the valves when they have been closed to comply with
the ACTION requirements. This precludes the need to cycle the valves with
full system differential pressure or when maintenance is being performed to
restore an inoperable PORV to operable status.
Surveillance requirement 4.4.4.1.b has been added to include testing of the
mechanical and electrical aspects of control systems for air-operated PORVs.
.
A-9 Generic Issue 70
Testing of PORVs in HOT STANDBY or HOT SHUTDOWN is required in order to
simulate the temperature and pressure environmental effects on PORVs. In
many PORV designs, testing at COLD SHUTDOWN is not considered to be a
representative test for assessing PORV performance under normal plant
operating conditions.
The Modified Standard Technical Specification (STS) requirements include the
following changes from prior STS guidance:
1. Clarify the statement of LCO by replacing "All" with "Both" where the
design includes two PORVs.
2. ACTION statement a. includes the requirement to maintain power to closed
block valve(s) because removal of power would render the block valve(s)
inoperable and the requirements of ACTION statement c. would apply. Power
is maintained to the block valve(s) so that it is operable and may be
subsequently opened to allow the PORV to be used to control reactor
pressure. Closure of the block valve(s) establishes reactor coolant
pressure boundary integrity for a PORV that has excessive seat leakage.
(Reactor coolant pressure boundary integrity takes priority over the
capability of the PORV to mitigate an overpressure event.) However, the
APPLICABILITY requirements of the LCO to operate with the block valve(s)
closed with power maintained to the block valve(s) are only intended to
permit operation of the plant for a limited period of time not to exceed the
next refueling outage (MODE 6) so that maintenance can be performed on the
PORVs to eliminate the seat leakage condition. The PORVs should normally be
available for automatic mitigation of overpressure events and should be
returned to OPERABLE status prior to entering STARTUP (MODE 2).
3. ACTION statements b. and c. include the removal of power from a closed
block valve as additional assurance to preclude any inadvertent opening of
the block valve at a time in which the PORV may not be closed due to
maintenance to restore it to OPERABLE status. (In contrast, ACTION
statement a. is intended to permit continued plant operation for a limited
period of time with the block valves closed, i.e., continued operation is
not dependent on maintenance at power to eliminate excessive PORV leakage,
and, therefore, ACTION statement a. does not require removal of power from
the block valve.)
4. ACTION statements a., b., and c. have been changed to terminate the
forced shutdown requirements with the plant being in HOT SHUTDOWN rather
than COLD SHUTDOWN because the APPLICABILITY requirements of the LCO do not
extend past the HOT STANDBY mode.
5. ACTION statement d. has been modified to establish remedial measures
that are consistent with the function of the block valves. The prime
importance for the capability to close the block valve is to isolate a
stuck-open PORV. Therefore, if the block valve(s) cannot be restored to
operable status within 1 hour, the remedial action is to place the PORV in
manual control to preclude its automatic opening for an overpressure event
and to avoid the potential for a stuck-open PORV at a time that the block
valve is inoperable. The time allowed to restore the block valve(s) to
operable status is based upon the remedial action time limits for inoperable
PORVs per ACTION statements b. and c. since the PORVs
.
A-10 Generic Issue 70
are not capable of mitigating an overpressure event when placed in manual
control. These actions are also consistent with the use of the PORVs to
control reactor coolant system pressure if the block valves are inoperable
at a time when they have been closed to isolate PORVs that have excessive
seat leakage. The modified ACTION statement does not specify closure of the
block valves because such action would not likely be possible when the block
valve is inoperable. Likewise, it does not specify either the closure of
the PORV, because it would not likely be open, or the removal of power from
the PORV. When the block valve is inoperable, placing the PORV in manual
control is sufficient to preclude the potential for having a stuck-open PORV
that could not be isolated because of an inoperable block valve. For the
same reasons, reference is not made to ACTION statements b. and c. for the
required remedial actions.
6. Surveillance requirement 4.4.4.2 has been modified to remove the
exception for testing the block valves when they are closed to isolate an
inoperable PORV. If the block valve is closed to isolate a PORV with
excessive seat leakage, the operability of the block valve is of importance,
because opening the block valve is necessary to permit the PORV to be used
for manual control of reactor pressure. If the block valve is closed to
isolate an otherwise inoperable PORV, the maximum allowable outage time is
72 hours, which is well within the allowable limits (25 percent) to extend
the block valve surveillance interval (92 days). Furthermore, these test
requirements would be completed by the reopening of a recently closed block
valve upon restoration of the PORV to operable status, i.e., completion of
the ACTION statement fulfills the required surveillance requirement.
.
A-11 Generic Issue 70
Enclosure A to Generic Letter 90-06
Attachment A-4
Modified Technical Specifications
for Babcock and Wilcox Plant
REACTOR COOLANT SYSTEM
3/4.4.4 RELIEF VALVE
LIMITING CONDITION FOR OPERATION
3.4.4 The power-operated relief valve (PORV) and its associated block valve
shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
a. With the PORV inoperable because of excessive seat leakage, within
1 hour either restore the PORV to OPERABLE status or close the
associated block valve with power maintained to the block valve;
otherwise, be in at least HOT STANDBY within the next 6 hours and^C
in HOT SHUTDOWN within the following 6 hours.
b. With the PORV inoperable due to causes other than excessive seat
leakage, within 1 hour either restore the PORV to OPERABLE status
or close the associated block valve and remove power from the
block valve, and be in HOT STANDBY within the next 6 hours and in
HOT SHUTDOWN within the following 6 hours.
c. With the block valve inoperable, within 1 hour restore the block
valves to OPERABLE status or place the associated PORV in manual
control and restore the block valve to operable status within the
next hour; otherwise, be in HOT STANDBY within the next 6 hours
and in HOT SHUTDOWN within the following 6 hours.
d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
4.4.4.1 In addition to the requirements of Specification 4.0.5, the PORV
shall be demonstrated OPERABLE at least once per 18 months by:
a. Operating the PORV through one complete cycle of full travel
during MODES 3 or 4, and
b. Performing a CHANNEL CALIBRATION of the actuation instrumentation.
.
A-12 Generic Issue 70
4.4.4.2 The block valve shall be demonstrated OPERABLE at least once per 92
days by operating the valve through one complete cycle of full travel unless
the block valve is closed in order to meet the requirements of ACTION b in
Specification 3.4.4.
4.4.4.3 The emergency power supply for the PORV and block valve shall be
demonstrated OPERABLE at least once per 18 months by:
a. Manually transferring motive and control power from the normal to
the emergency power bus, and
b. Operating the valve through a complete cycle of full travel.
BABCOCK & WILCOX PLANTS
.
A-13 Generic Issue 70
Enclosure A to Generic Letter 90-06
Attachment A-5
Applicable to Babcock and Wilcox Plants
3/4.4.4 RELIEF VALVE
Bases of the Limiting Condition for Operation (LCO) and Surveillance
Requirements:
The OPERABILITY of the PORV and block valve is determined on the basis of
their being capable of performing the following functions:
A. Manual control of the PORV to control reactor coolant system pressure.
This is a function that is used for the steam generator tube rupture
accident and for plant shutdown. This function has been classified as
safety related for more recent plant designs.
B. Maintaining the integrity of the reactor coolant pressure boundary.
This is a function that is related to controlling identified leakage
and ensuring the ability to detect unidentified reactor coolant
pressure boundary leakage.
C. Manual control of the block valve to: (1) unblock an isolated PORV to
allow it to be used for manual control of reactor coolant system
pressure (Item A), and (2) isolate the PORV with excessive seat leakage
(Item B).
D. Automatic control of the PORV to control reactor coolant system
pressure. This is a function that reduces challenges to the code
safety valves for overpressurization events.
E. Manual control of a block valve to isolate a stuck-open PORV.
Surveillance Requirements provide the assurance that the PORV and block
valve can perform their functions. Specification 4.4.4.1 addresses the
PORV, 4.4.4.2 the block valve, and 4.4.4.3 the emergency (backup) power
source. The latter is provided for either the PORV or block valve,
generally as a consequence of the TMI ACTION requirements to upgrade the
operability of PORVs and block valves, where they are installed with
non-safety-grade power sources, including instrument air, and are provided
with backup (emergency) power sources. The block valve is exempt from the
surveillance requirements to cycle the valve when it has been closed to
comply with the ACTION requirements. This precludes the need to cycle the
valve with full system differential pressure or when maintenance is being
performed to restore an inoperable PORV to operable status.
.
A-14 Generic Issue 70
Testing the PORV in HOT STANDBY or HOT SHUTDOWN is required in order to
simulate the temperature and pressure environmental effects on the PORV. In
many PORV designs, testing at COLD SHUTDOWN is not considered to be a
representative test for assessing PORV performance under normal plant
operating conditions.
The Modified Standard Technical Specification (STS) requirements include the
following changes from prior STS guidance:
1. ACTION statement a. includes the requirement to maintain power to the
closed block valve, because removal of power would render the block valve
inoperable and the requirements of ACTION statement c. would apply. Power
is maintained to the block valve so that it is operable and may be
subsequently opened to allow the PORV to be used to control reactor
pressure. Closure of the block valve establishes reactor coolant pressure
boundary integrity for a PORV that has excessive seat leakage. (Reactor
coolant pressure boundary integrity takes priority over the capability of
the PORV to mitigate an overpressure event.) However, the APPLICABILITY
requirement of the LCO to operate with the block valve closed with power
maintained to the block valve is only intended to permit operation of the
plant for a limited period of time not to exceed the next refueling outage
(MODE 6) so that maintenance can be performed on the PORV to eliminate the
seat leakage condition. The PORV should normally be available for automatic
mitigation of overpressure events and should be returned to OPERABLE status
prior to entering STARTUP (MODE 2).
2. ACTION statement b. includes the removal of power from the closed block
valve as additional assurance to preclude any inadvertent opening of the
block valve at a time in which the PORV may not be closed due to maintenance
to restore it to OPERABLE status. (In contrast, ACTION statement a. is
intended to permit continued plant operation for a limited period of time
with the block valve closed, i.e., continued operation is not dependent on
maintenance at power to eliminate excessive PORV leakage, and, therefore,
ACTION statement a. does not require removal of power from the block valve.)
3. ACTION statements a. and b. have been changed to terminate the forced
shutdown requirements with the plant being in HOT SHUTDOWN rather than COLD
SHUTDOWN because the APPLICABILITY requirements of the LCO do not extend
past the HOT STANDBY mode.
4. ACTION statement c. has been modified to establish remedial measures
that are consistent with the function of the block valves. The prime
importance for the capability to close the block valve is to isolate a
stuck-open PORV. Therefore, if the block valve cannot be restored to
operable status within 1 hour, the remedial action is to place the PORV in
manual control to preclude its opening for an overpressure event and to
avoid the potential for a stuck-open PORV at a time that the block valve is
inoperable. The time allowed to restore the block valve to operable status
is based upon the remedial action time limits for an inoperable PORV per
ACTION statement b. since the PORV is not capable of mitigating an
overpressure event when placed in manual control. This action is also
consistent with the use of the PORV to control reactor coolant system
pressure if the block valve is inoperable at a time when it was
.
A-15 Generic Issue 70
closed to isolate a PORV that has excessive seat leakage. The modified
ACTION statement does not specify closure of the block valve because such
action would not likely be possible when the block valve is inoperable.
Likewise, it does not specify either the closure of the PORV, because it
would not likely be open, or the removal of power from the PORV. When the
block valve is inoperable, placing the PORV in manual control is sufficient
to preclude the potential for having a stuck-open PORV that could not be
isolated because of an inoperable block valve. For the same reasons,
reference is not made to ACTION statement b. for the required remedial
action.
5. Surveillance requirement 4.4.4.2 has been modified to remove the
exception for testing the block valve when it is closed to isolate an
inoperable PORV. If the block valve is closed to isolate a PORV with
excessive seat leakage, the operability of the block valve is of importance,
because opening the block valve is necessary to permit the PORV to be used
for manual control of reactor pressure. If the block valve is closed to
isolate an otherwise inoperable PORV, the maximum allowable outage time is
72 hours, which is well within the allowable limits (25 percent) to extend
the block valve surveillance interval (92 days). Furthermore, these test
requirements would be completed by the reopening of a recently closed block
valve upon restoration of the PORV to operable status, i.e., completion of
the ACTION statement fulfills the required surveillance requirement.
.
B-1
Enclosure B to Generic Letter 90-06
Staff Positions Resulting from
Resolution of Generic Issue 94 -
Additional Low-Temperature Overpressure Protection
For Light-Water Reactors
1. BACKGROUND
Generic Issue 94 (GI-94), "Additional Low-Temperature Overpressure
Protection for Light-Water Reactors," addresses concerns with the
implementation of the requirements set forth in the resolution of Unresolved
Safety Issue (USI) A-26, "Reactor Vessel Pressure Transient Protection
(Overpressure Protection)." In support of GI-94, the Battelle Pacific
Northwest Laboratories (PNL) performed a study based on actual operating
reactor experiences to determine the risks associated with current
low-temperature overpressure protection (Ltop) systems. A report, prepared
by PNL, has been issued as NUREG/CR-5186, "Value/Impact Analysis of Generic
Issue 94, Additional Low Temperature Over-pressure Protection for Light-Water
Reactors," dated November 1988. The staff has prepared a regulatory analysis
for GI-94 based on the work performed by PNL and reported in NUREG-1326,
"Regulatory Analysis for the Resolution of Generic Issue 94, Additional
Low-Temperature Overpressure Protection for Light-Water Reactors" (Enclosure
D).
Low-temperature overpressure protection (Ltop) was designated as Unresolved
Safety Issue A-26 in 1978 (NUREG-0371). PWR licensees implemented
procedures to reduce the potential for overpressure events and installed
equipment modifications to mitigate such events based on the staff
recommendations from the USI A-26 evaluations, under Multi-Plant Action Item
B-04 (NUREG-0748). Current staff guidelines for Ltop are in Standard Review
Plan Section 5.2.2, "Overpressure Protection," and in its attached Branch
Technical Position (BTP) RSB 5-2, "Overpressure Protection of Pressurized
Water Reactors While Operating at Low Temperatures" (NUREG-0800).
The administrative controls and procedures that were identified as part of
Multi-Plant Action Item B-04 include the following items:
1. Minimize the time the reactor coolant system (RCS) is maintained in
a water-solid condition.
2. Restrict the number of high-pressure safety injection pumps
operable to no more than one when the RCS is in the Ltop condition.
3. Ensure that the steam generator to RCS temperature difference is
less than 50 Deg F when a reactor coolant pump (RCP) is being started
in a water-solid RCS.
4. Set the PORV setpoint (if the particular plant relies on this
component for Ltop) to a plant-specific analysis supported value, and
have surveillance that checks the PORV actuation electronics and
setpoint.
.
B-2
Twelve PWR overpressure transients were reported during the period from 1981
to 1983 after completion of USI A-26. Two of these events, at Turkey Point
Unit 4, exceeded the pressure/temperature limits of the technical
specifications. During this same timeframe, there were 37 reported
instances when at least one Ltop channel was out of service. In 12 of these
cases, both Ltop channels were inoperable.
The continuation of overpressure transient events, and the unavailability of
Ltop protection channels, suggested the need to reevaluate the current
overpressure protection requirements, or their implementation, to determine
whether additional measures are warranted.
Major overpressurization of the reactor coolant system while at low
temperature, if combined with a critical crack in the reactor pressure
vessel welds or plate material, could result in a brittle fracture of the
pressure vessel. Failure of the pressure vessel could make it impossible to
provide adequate coolant to the reactor core and result in major core damage
or a core melt accident.
The safety significance of these continuing low-temperature overpressure
transients was designated as Generic Issue 94, "Additional Low Temperature
Overpressure Protection." The concerns of GI-94 are applicable to all PWR
plants regardless of the features used to mitigate a low-temperature
overpressure event or of any measures to preclude events that would
challenge these features or exceed the design basis for Ltop.
The implementation of the requirement for an Ltop system (the resolution of
USI A-26) has been found to be essentially uniform for the Combustion
Engineering (CE) and Westinghouse (W) PWRs. With the exception of a few
plants,* the Ltop protection systems consist of either redundant PORVs or
redundant safety relief valves (SRVs) in the residual heat removal (RHR)
system and in general meet the guidance set forth in Branch Technical
Position RSB 5-2, "Overpressurization Protection of Pressurized Water
Reactors While Operating at Low Temperatures."
Variability in meeting IEEE-279 requirements, equipment environmental
qualification, and in meeting the guidance of Regulatory Guide 1.26,
"Quality Group Classification and Standards for Water-, Steam-, and
Radioactive-Waste-Containing Components of Nuclear Power Plants," exists.
As part of the NRC staff acceptance of Ltop protection system designs for
the implementation of the resolution of USI A-26, it was concluded that the
costs associated with upgrading existing systems to meet the guidance of
Regulatory Guide 1.26 were not
* CE - San Onofre Units 2 and 3 rely on a single RHR (SDCS) SRV for Ltop.
With the SRV inoperable, depressurize and vent within 8 hours.
- Maine Yankee relies on two PORVs when pressure is above 400 psig
and two RHR SRVs when pressure is below 400 psig.
W - DC Cook Units 1 and 2 rely on either two PORVs or one PORV and one
RHR SRV.
- Yankee Rowe relies on one PORV and two RHR SRVs.
- Newer Westinghouse plants allow either two PORVs or two RHR SRVs.
.
B-3
justifiable. Further evaluations performed for GI-94 have also concluded
that it is not cost beneficial to upgrade these systems to fully
safety-grade standards.
2. CURRENT STANDARD TECHNICAL SPECIFICATION REQUIREMENTS
The section of the Standard Technical Specifications (STS) covering the Ltop
protection system is entitled Overpressure Protection System, Section
3.4.10.3 for CE plants and Section 3.4.9.3 for W plants. The Ltop system
setpoints are established based on additional restrictions for the restart
of an idle reactor coolant pump and on the number of high-pressure safety
injection pumps and/or coolant charging pumps allowed to be operable when
Ltop is required. These additional restrictions define the initial
conditions for the plant-specific transient analyses performed to establish
the Ltop system setpoints. The additional restrictions are provided
regarding the restart of inactive reactor coolant pumps in Sections 3.4.1.3
(Hot Shutdown) and 3.4.1.4 (Cold Shutdown). High-pressure safety injection
pump operability restrictions are provided in Section 3/4.5.3 (ECCS
Subsystems). In addition to these administrative restric-tions, the
transient analyses are based on a dual-channel system being operable to
satisfy the single failure criterion of 10 CFR Part 50, Appendix A, for a
system that performs a safety function. Therefore, the Overpressure
Protection System TS is consistent with Criterion 2 of the Commission's
Policy Statement on Technical Specification Improvements for Nuclear Power
Plants. The TS also satisfied Criterion 3 of the policy statement in that
the Ltop system is the primary success path for the mitigation of
low-temperature overpressure transients that present a challenge to a
fission product barrier, in this case, the reactor pressure vessel.
PORVs are relied on, by most Westinghouse designed plants and about one-half
of the Combustion Engineering plants, to provide Ltop protection. In
addition to PORVs, the RHR SRVs are also relied on to provide Ltop
protection for some W plants and for the CE plants that do not have PORVs.
Newer W plants have TS that require either two PORVs or two RHR SRVs for
Ltop protection.
The current STS ACTION requirements for the Ltop system include a 7-day
allowable outage time (AOT) to restore an inoperable Ltop channel to
operable status before other remedial measures would have to be taken. In
addition, ACTION d. states that the provisions of Specification 3.0.4 are
not applicable. Therefore, the plant may enter the modes for which the
Limiting Conditions for Operation (LCO) apply, during a plant shutdown or
placement of the head on the vessel following refueling, when an Ltop
channel is inoperable. In this situation, the 7-day AOT applies for
restoring the channel to operable status before other remedial measures
would have to be taken. This is the same manner in which the ACTION
requirements apply when an Ltop channel is determined to be inoperable while
the plant is in a mode for which the Ltop system is required to be operable.
Based on the NRC evaluation of the Ltop system unavailability, it is
concluded that additional restrictions on operation with an inoperable Ltop
channel are warranted when the potential for a low-temperature overpressure
event is the
.
B-4
highest, and especially when the plant is in a water-solid condition.
Furthermore, it is concluded that the additional restrictions regarding the
restart of inactive reactor coolant pumps and regarding the operability of
high-pressure safety injection pumps should be implemented in the TS, as
indicated in the STS, and licensees should verify that these administrative
restrictions have been implemented. Finally, it is concluded that these
additional measures will help to emphasize the importance of the Ltop
system, especially while operating in a water-solid condition, as the
primary success path for the mitigation of overpressure transients during
low-temperature operation.
3. IMPROVEMENTS IN PROTECTION SYSTEM AVAILABILITY
The staff has determined that Ltop protection system unavailability is the
dominant contributor to risk from low-temperature overpressure transients.
The staff has further concluded that a substantial improvement in
availability when the potential for an overpressure event is the highest,
and especially during water-solid operations, can be achieved through
improved administrative restrictions on the Ltop system.
In developing the staff position on the resolution of the low-temperature
overpressure protection generic issue, a number of factors have been taken
into consideration.
The staff has considered the conditions under which a low-temperature
overpressure transient is most likely to occur. While Ltop protection is
required for all shutdown modes, the most vulnerable period of time was
found to be MODE 5 (Cold Shutdown) with the reactor coolant temperature less
than or equal to 200 Deg F, especially when water-solid, based on the
detailed evaluation of operating reactor experiences performed in support of
GI-94. Ltop transients that have challenged the overpressure protection
system have occurred with reactor coolant temperatures in the range of 80
Deg F to 190 Deg F. In addition, a review of the STS for containment
integrity indicates that there are no specific requirements imposed during
MODE 5, when the reactor coolant temperature is below 200 Deg F. Industry
responses to Generic Letter 87-12, "Loss of RHR While RCS Partially Filled,"
dated July 9, 1987, also indicate that containment integrity during MODE 5
is often relaxed to allow for testing, maintenance, and the repair of
equipment.
In addition, the staff takes note of the fact that, in all instances when
pressure/temperatures limits in the TS have been exceeded, one Ltop
protection channel was removed from service for maintenance-related
activities. During these events the redundant Ltop protection channel
failed to mitigate the overpressure transient as a result of a
system/component failure that had not been detected.
The reported Ltop transients have occurred in MODE 5 with RCS temperatures
ranging from 80 Deg F to 190 Deg F. Since this temperature range includes
MODE 6, RCS temperature less than 140 Deg F but with k-eff less than 0.95 as
compared to k-eff less than 0.99 for MODE 5, the staff concludes that the
additional administrative restriction for the single channel AOT is
applicable to MODE 5 and MODE 6 (with the reactor pressure vessel head on).
.
B-5
The staff concludes that the Ltop system performs a safety-related function
and inoperable Ltop equipment should be restored to an operable status in a
shorter period of time. The current 7-day AOT for a single channel is
considered to be too long under certain conditions. The staff has concluded
that the AOT for a single channel should be reduced to 24 hours when
operating in MODE 5 or 6 when the potential for an overpressure transient is
highest. The operating reactor experiences indicate that these events occur
during planned heatup (restart of an idle reactor coolant pump) or as a
result of maintenance and testing errors while in MODE 5. The reduced AOT
for a single channel in MODES 5 and 6 will help to emphasize the importance
of the Ltop system in mitigating overpressure transients and provide
additional assurance that plant operation is consistent with the design
basis transient analyses.
Based on the foregoing concerns, added assurance of Ltop availability is to
be provided by revising the current Technical Specification for Overpressure
Protection to reduce the AOT for a single channel from 7 days to 24 hours
when the plant is operating in MODES 5 or 6. Attachment B-1 is provided for
guidance for Westinghouse and CE plants. The guidance provided is also
applicable to plants that rely on both PORVs and RHR SRVs or that rely on
RHR SRVs only. Attachment B-2 provides the staff bases for the Overpressure
Protection Technical Specification.
In performing the studies for GI-94, the staff has assumed that the
administrative controls and procedures identified in Section 1 have been
implemented to ensure that the plant is being operated within the design
base. If it is determined that the design base was developed based on
restricted safety injection pump operability and/or differential temperature
restrictions for RCP restart and that these restrictions have not been
implemented as part of USI A-26, then these restrictions should be
implemented now. This is not a new requirement. Attachment B-3 is provided
for guidance.
.
B-6 Generic Issue 94
Enclosure B to Generic Letter 90-06
Attachment B-1
Modified Technical Specifications
for Combustion Engineering and Westinghouse Plants
REACTOR COOLANT SYSTEM
OVERPRESSURE PROTECTION SYSTEM
LIMITING CONDITION FOR OPERATION
3.4.9.3 Two power-operated relief valves (PORVs) shall be OPERABLE with a
lift setting of less than or equal to [450] psig.
APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than
or equal to [275] F, MODE 5, and MODE 6 when the head is on the reactor
vessel and the RCS is not vented through a square inch or larger vent.
ACTION:
a. With one PORV inoperable in MODE 4, restore the inoperable PORV to
OPERABLE status within 7 days or depressurize and vent the RCS
through at least a square inch vent within the next 8 hours.
b. With one PORV inoperable in MODES 5 or 6, either (1) restore the
inoperable PORV to OPERABLE status within 24 hours, or (2)
complete depressurization and venting of the RCS through at least
a square inch vent within a total of 32 hours.
c. With both PORVs inoperable, complete depressurization and venting
of the RCS through at least a square inch vent within 8
hours.
d. With the RCS vented per ACTIONS a, b, or c, verify the vent
pathway at least once per 31 days when the pathway is provided by
a valve(s) that is locked, sealed, or otherwise secured in the
open position; otherwise, verify the vent pathway every 12 hours.
e. In the event either the PORVs or the RCS vent(s) are used to
mitigate an RCS pressure transient, a Special Report shall be
prepared and submitted to the Commission pursuant to Specification
6.9.2 within 30 days. The report shall describe the circumstances
initiating the transient, the effect of the PORVs or RCS vent(s)
on the transient, and any corrective action necessary to prevent
recurrence.
f. The provisions of Specification 3.0.4 are not applicable.
.
B-7 Generic Issue 94
SURVEILLANCE REQUIREMENTS
4.4.9.3 Each PORV shall be demonstrated OPERABLE by:
a. Performance of an ANALOG CHANNEL OPERATIONAL TEST, but excluding
valve operation, at least once per 31 days; and
b. Performance of a CHANNEL CALIBRATION at least once per 18 months;
and
c. Verifying the PORV isolation valve is open at least once per 72
hours.
.
B-8 Generic Issue 94
Enclosure B to Generic Letter 90-06
Attachment B-2
3/4.4.9.3 OVERPRESSURE PROTECTION SYSTEM
Bases of the Limiting Condition for Operation and Surveillance Requirements:
The OPERABILITY of the PORVs is determined on the basis of their being
capable of performing the function to mitigate an overpressure event during
low-temperature operation.
The Modified Standard Technical Specification (STS) requirements include the
following changes from prior STS guidance:
1. The depressurizing and venting of the RCS is not classified as an
overpressure protection system. However, the APPLICABILITY of the LCO
excludes MODE 6 when the RCS is adequately vented. This avoids any
possible question on Specification 3.0.4 being applied to preclude
placement of the head on the vessel if any part of the LCO is not met
when the RCS is vented.
2. The APPLICABILITY for MODE 6 is clarified as "when the head is on
the reactor vessel" rather than as "MODE 6 with the reactor vessel head
on."
3. ACTION a. is revised to clarify that it is only applicable in MODE
4.
4. ACTION b. was added to reduce the allowed outage time for an
inoperable PORV to 24 hours in MODES 5 or 6. Because this LCO does not
apply under certain conditions specified under the APPLICABILITY for
this specification, the ACTION statements likewise do not apply under
those conditions. ACTIONS a. and b. do not repeat those qualifying
conditions that apply for these modes since the actions only apply when
the unit is under those conditions.
5. ACTION d. includes the requirements to verify that ACTIONS a., b.,
or c. continue to be met on an ongoing basis when the unit would be in
MODES 4, 5, or 6.
6. The Surveillance Requirements were simplified by removing
requirements that exist because of the general requirements applicable
to all surveillance requirements as specified in Section 4.0 of the TS.
7. Surveillance Requirement 4.4.9.3.2 was removed since it is
addressed by ACTION d.
For plants with existing TS for PORVs used for Ltop, the only required
change is that indicated to restrict the applicability of ACTION a. to MODE
4 and for incorporating ACTION b. Any other changes that are proposed
consistent with
.
B-9 Generic Issue 94
the above guidance are voluntary. For a plant without existing TS for PORVs
that are used for Ltop, a TS should be proposed that conforms to the above
guidance.
Because some plants use residual heat removal (RHR) safety relief valves for
Ltop, either in addition to or in lieu of PORVs, similar requirements are
included in TS as noted above for PORVs. The same changes in ACTION
requirements a. and b. are required, as noted above, for these plants.
Likewise, any plant without existing TS for RHR suction relief valves that
are used for Ltop should propose TS that are consistent with the above
guidance. When only RHR safety relief valves are used for Ltop, the
Surveillance Requirements would state: "No additional requirements other
than those required by Specification 4.0.5."
.
B-10 Generic Issue 94
Enclosure B to Generic Letter 90-06
Attachment B-3
Technical Specifications Guidance
for Combustion Engineering and Westinghouse Plants
Operational Limitations Consistent With the Design Basis Assumptions for the
Low-temperature Overpressure Protection (Ltop) System
The TS requirements for Ltop typically apply in MODE 4 when the temperature
of any cold leg is below 275 F, MODE 5, and MODE 6 when the head is on the
reactor vessel. During these conditions, one train (or channel) of the Ltop
system is capable of mitigating an Ltop event that is bounded by the largest
mass addition to the RCS or by the largest increase in RCS temperature that
can occur. The largest mass addition to the RCS is limited based upon the
assumption that no more than a fixed number of pumps are capable of
providing makeup or injection into the RCS. Hence, this is a matter
important to safety that pumps in excess of this design basis assumption for
Ltop not be capable of providing makeup or injection to the RCS.
The capability for makeup and injection to the RCS is also a safety concern
for normal makeup to the reactor coolant system for reactivity control as
well as for events that could result in a loss of coolant from the RCS. The
former are covered by Technical Specifications (TS) under Reactivity Control
Systems, Charging Pump - Shutdown (MODES 5 and 6); Charging Pumps -
Operating (MODES 1 through 4); and Flow Paths - Operating (MODES 1 through
4). The latter is covered by TS under Emergency Core Cooling Systems, ECCS
Subsystems -T(cold) Less Than 350 F (MODE 4).
The manner in which restrictions, consistent with the design basis
assumptions of the Ltop system, have been incorporated in TS that require
the operability of makeup or injection pumps has varied depending upon
plant-specific considerations for the Ltop design and plant-specific designs
for the use of pumps for makeup and ECCS functions. A common method has
been the use of footnotes to the pump operability requirements to note that:
A maximum of one Safety Injection [and/or] one charging pump shall be
OPERABLE when the temperature of one or more of the RCS cold legs is
less than 275 F.
This footnote is used for each specification that requires the operability
of a safety injection and/or charging pump in MODES 4 or 5.
The Surveillance Requirements typically include the following:
All Safety Injection [and/or] charging pumps, except the above required
OPERABLE pump[s], shall be demonstrated to be inoperable by verifying
that the motor circuit breakers are secured in the open position at
least
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B-11 Generic Issue 94
once per 12 hours whenever the temperature of one or more of the RCS
cold legs is less than or equal to 275 F.
Generally, it is preferable to include requirements for implementing the
intent of an LCO as part of an LCO rather than to only define requirements,
such as securing motor circuit breakers in the open position, in a
Surveillance Requirement. Furthermore, the requirements for operable pumps
could be stated in terms of requiring one pump to be operable rather in
terms of "at least one pump shall be operable" and then including a footnote
requiring that, in fact, no more than one pump shall be operable. The
preferred alternative would be an LCO which stated:
One Safety Injection [and/or] charging pump shall be operable and all
other Safety Injection [and/or] charging pumps shall be secured with
their motor circuit breakers in the open position.
The form of the above requirements for any given specification would be
dependent upon which pumps are addressed by that specification, e.g.,
charging or injection pumps or both.
The surveillance requirements would be similar to that noted above with the
following substitution:
. . .except the above required OPERABLE pump(s), shall be demonstrated
to be secured by verifying that the motor circuit breakers are in the
open position. . . .
Changes to plant TS should be proposed to incorporate one of the above
methods, to ensure that pumps are not capable of initiating a mass addition
to the RCS that exceeds the design basis assumptions for the Ltop system,
for plants that do not currently include such requirements.
The largest temperature increase in the RCS that could result in a challenge
to the Ltop system is dependent upon the differential temperature between
the RCS and the secondary system when starting a reactor coolant pump.
Hence, this is also a matter important to safety when reactor coolant pumps
are started and the resulting increase in RCS temperature is in excess of
the design basis assumption for the Ltop system to mitigate the resulting
increase in RCS pressure. The manner in which this design basis assumption
of the Ltop system is reflected in TS has been the use of a footnote to the
reactor coolant pump operability requirements to note that:
A reactor coolant pump shall not be started with one or more of the
RCS cold leg temperatures less than or equal to 275 F unless the
secondary water temperature of each steam generator is less than F
above each of the RCS cold leg temperatures.
The above footnote has been included in the TS for residual heat removal
under title of the Reactor Coolant System, Hot Shutdown.
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B-12 Generic Issue 94
Changes to plant TS should be proposed to incorporate the above method, to
ensure that the starting of RCS pumps are not capable of initiating a
pressure transient that exceeds the design basis assumptions for the Ltop
system, for plants that do not currently have this requirement.
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Page Last Reviewed/Updated Tuesday, March 09, 2021