United States Nuclear Regulatory Commission - Protecting People and the Environment

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. 

.

                                     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."

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                                     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.

.ENDEND
Page Last Reviewed/Updated Friday, June 28, 2013