ML20236L804
| ML20236L804 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/06/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20236L795 | List: |
| References | |
| NUDOCS 9807130185 | |
| Download: ML20236L804 (9) | |
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1 UNITED STATES g
P; NUCLEAR REGULATORY COMMISSION F
e WASHINeToN, D.C. 20008 4001 i
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO INSERVICE TESTING PROGRAM RELIEF REQUESTS RP-01 AND RV-05 COMMONWEALTH EDISON COMPANY LASALLE COUNTY STATION. UNITS 1 AND 2 DOCKET NOS. 50-373 AND 50-374
1.0 INTRODUCTION
The Code of Federal Regulations.10 CFR 50.55a, requires that inservice testing (IST) of certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)
Class 1,2, and 3 pumps and valves be performed in accordance with Section XI of the ASME
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Code and applicable addenda, except where altematives have been authorized or relief has been l
requested by the licensee and granted by the Commission pursuant to Sections (a)(3)(i),
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(a)(3)(ii), or (f)(6)(i) of 10 CFR 50.55a. In proposing altematives or requesting relief, the licensee j
l must demonstrate that: (1) the proposed altematives provide an acceptable level of quality and l
safety; (2) compliance would result in hardship or unusual difficulty without a compensating
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increase in the level of quality and safety; or (3) conformance is impractical for its facility.
Section 50.55a authorizes the Commission to approve altematives and to grant relief from ASME
- Code requirements upon making the necessary findings. NRC guidance contained in Generic Letter (GL) 89-04, " Guidance on Developing Acceptable Inservice Testing Programs," provides altematives to the Code requirements detarmined acceptable to the staff. Altematives that l-conform with the guidance in GL 89-04 may be implemented without additional NRC approval, but are subject to review during inspections. Further guidance was given in GL 89-04,
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Supplemont 1, and NUREG-1482,." Guidelines for Inservice Testing at Nuclear Power Plants."
In a letter dated March 4,1998, relief requests RP-01 and RV-05 were submitted by the
.Commonweaan Edison Company (Comed, the licensee) for the LaSalle County Station, Units 1 l
and 2, second 10-year interval program for inservice testing of pumps and valves. On May 15, l
1998, the NRC issued a request for additionalinformation. The licensee responded to this request in a letter dated June 12,1998. The evaluation of the relief requests and the supporting L
information is provided below.- The licensee's IST program covers the second 10-year IST I
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intervals from November 23,1994, to November 23,2004, for Unit 1 and from October 17,1994, L
to October 17,2004, for Unit 2. The LaSalle County Station, Units 1 and 2, IST program is based on the requirements of the 1989 Edition,Section XI, of the ASME Code, which, by reference, incorporates Part 6, " Inservice Testing of Pumps in Ught-Water Reactor Power Plants," and Part 10, " Inservice Testing of Valves in Light-Water Reactor Power Plants," of the i
ASME Operations and Maintenance Standard OMa-1988.
9907130185 900706 PDR ADOCK 050003~J G
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? l 2.0 RE'.lEF REQUEST RP-01 Relief is requested from the testing requirements of OM4, Section 5.2(b), for the Unit 1 and 2 High Pressure Core Spray (HPCS), Low Pressure Core Spray (LPCS), Residual Heat Removal (RHR), and Reactor Core Isolation Cooling (RCIC) water leg pumps listed below. The licensee L
has proposed to measure vibration in accordance with OM-6, trend differential pressure and vibration, and use the current monitoring scheme to verify that the HPCS, LPCS, RHR, and RCIC
. pump discharge lines are full of water. Flow measurement will not be performed.
1(2)E22 C003, High Pressure Core Spray Water Leg Pumps 1(2)E21-C002, Low Pressure Core Spray Water Leg Pumps 1(2)E12-C003, Residual Heat Removal Water Leg Pumps 1(2)E51-C003, Reactor Core Isolation Cooling Water Leg Pumps An earlier version of this request was approved in a safety evaluation (SE) dated December 6,
. 1996. Prior to December 6,1996, relief had been granted on an interim basis in a safety evaluation dated December 8,1995.. The March 4,1998, submittal revisM the request to include the RCIC water leg pumps 1(2)E51-C003.
2.1 Basis for Relief The 1:censee provided the following basis for the request:
Instrumentation is not installed for measuring flow rates. Pump flow varies with system operation and system leakage; therefore, establishing flow rates for testing purposes is not practical. The primary purpose of these pumps is to maintain the HPCS, LPCS, RHR, and RCIC pump discharge lines filled to limit the potential for a water hammer upon initiation. System modification to provide test measuring locations places undue burden on the utility without demonstrating any increase in the level of plant safety. These pumps are in continuous operation.
Pump p arformance is continuously monitored by a low pressure alarm on each HPCS, LPCS, RHR, and RCIC pump discharge header.
LaSalle Station monitors the pump for degradation by measuring and recording pump inlet pressure, discharge pressure, differential pressure, and vibration, with the differential pressure and vibration data trended. These measurements are taken quarterly and provide satisfactory indication of operational readiness as well as the ability to detect potential degradation. In addition, the main ECCS
[ Emergency Core Cooling System) pumps discharge headers each have a low pressure alarm which continuously monitors the operability of the respective water leg pump. Station Technical Specifications also verify operability of the water leg pumps by verifying flow through a high point vent on a monthly basis.
2.2 ALTERNATIVE TEST The licensee proposes:
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. Vibration measurement will be obtained under normal operating conditions and evaluated in accordance with OM-S. LaSalle verifies operability of these pumps i
by pressure maintenance of the HPCS, LPCS, RHR, and RCIC pump discharge l
' lines within allowable pressure limits. In addition, performance monitoring of the pumps' mechanical and hydraulic performance is trended.
2.3 EVALUATION
- The Code requires that each water leg pump be tested by establishing a fixed and repeatable hydraulic reference value of either differential pressure or pump flow, establishing the reference value during quarterly testing, and recording the measured hydraulic value and bearing vibration for comparison with Code acceptance criteria. The design of the HPCS, LPCS, RHR, and RCIC water leg pumps does not enable inservice testing to be readily performed in accordance with the Code. The necessary instrumentation is not installed in the systems and a modification would be required for measuring flow. The licensee proposes to monitor the pumps for mechanical degradation (i.e., vibration monitoring) and for hydraulic degradation by measuring and recording pump inlet pressure, discharge pressure, and differential pressure, without monitoring flow.
The NRC's evaluation of this relief request in the December 8,1995, SE stated that the licensee did not indicate whether portable flow instrumentation had been considered for flow measurement of the ECCS water leg pumps. At least one other plant (Perry) performs inservice testing in accordance with the Code requirements using altrasonic flow instrumentation.
' Although the licensee stated in its initial submittal that there is no flow criterion on these pumps that could be used to determine whether the pumps are satisfactorily performing their safety -
function, IST acceptance criteria do not require testing at the design condition.- Therefore, if ultrasonic flow instrumentation could be utilized, acceptance criteria in accordance with the Code requirements could be established. The NRC has stated that the installation of instrumentation to meet a later edition of the Code is not considered a backfit (see Minutes of the Public Meetings on GL 89-04, Question 105 and Response).
While the inservice testing would not be as complete as it would be if the Code requirements were imposed, Section 50.55a does include pmvisions for impracticalities due to design limitations, as the initialimposition of the Code requirements was subsequent to the design and construction of a number of nuclear plants. For the water leg pumps, which are continuously operating pumps, the~ safety function is to keep the ECCS pump discharge header piping in a l
filled condition to prevent a water hammer upon pump start. The actual output and hydraulic performance of the water leg pumps are not critical to the safety function, as long as the pumps are capaMe of maintaining the piping full. Alarms would promptly alert plant operators whenever the water leg pumps do not maintain the piping pressure to a set alarm level. In addition, vibration data will be indicative of levels trending toward unacceptable values and should allow time for the licensee to take corrective actions before the pumps fail. The proposed attemative provides a reasonable assurance of operational readiness because (1) differential pressure and l
bearing vibration are measured and trended; and (2) alarms are present which provide a L
continuous monitoring of degradation in the pressure of the ECCS discharge lines.
The licensee did not discuss whether the pumps are included in a preventive maintenance program because of the impracticalities of full compliance with the inservice testing requirements, if the pumps are not already included in such a program, it is recommended that t.
4 an assessment of the past operating history of the pumps be performed and a determination be made as to whether or not periodic maintenance is warranted. The granting of the relief is not, however, dependent on the licensee's prior performance of such an assessment. The monitoring during continuous operation via the low pressure alarms, pressure differential measurements, i
and vibration measurements will provide adequate assurance of the operational readiness for l
operation in an accident mitigation condition.
2.4 Conclusion Relief request RP-01 is granted pursuant to 10 CFR 50.55a(f)(6)(1) based on the impracticality of performing the required testing and in consideration of the burden on the licensee if the Code requirements were imposed. The proposed attemative testing will provide adequate assurance of the operational readiness of the HPCS, LPCS, RHR, and RCIC pumps to perform their safety function of maintaining the ECCS discharge piping full of water to prevent a water hammer in the event the ECCS is actuated.
3.0 RELIEF REQUEST RV-05 Relief is requested from OM-1, paragraph 1.3.4.3(a), which requires that "within every 6 months period operability tests shall be performed unless historical data indicates a requirement for more frequent testing." This request pertains to the suppression chamber-drywell vacuum breakers 1(2)-PC001 A, B, C, and D. The licensee proposed to setpoint test these valves in accordance with the 18 month setpoint test frequency in the Technical Specifications.
3.1 Basis for Relief.
i The licensee provided the following bas!s for the request:
The primary containment ensures that the release of radioactive materials will be restricted to those paths and associated leak rates assumed in the accident analyses.
This restriction in conjunction with the leakage !!mitation, will limit the site boundary radiation dose to within the limits of 10 CFR Part 100 during accident conditions. The primary contWnment walls have a steel liner, which acts as a low leakage barrier.
The primary containment structure consists of a drywell area and a suppression pool
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area. The prir,1ary function of the drywellis to contain the effects of a design-basis I
recirculation line break and direct the steam released from a pipe break into the l
suppression chamber pool. The drywell contains a floor that serves as a pressure barrier j
between the drywell and suppression chamber and as a support structure for the reactor pedestal.
The primary function of the suppression chamber is to provide a reservoir of water l
capable of condensing the steam flow from the drywell and collecting the non-condensable gases in the suppression chamber air space.
Vacuum relief valves are provided between the drywell and suppression chamber to prevent exceeding the drywell floor negative design pressure and backflooding of the suppression pool water into the drywell. The vacuum relief valves are designed to l
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J 5-equalize the pressure between the drywell and wetwell air space regions so that the reverse pressure differential across the drywell floor will not exceed the design value of five pounds per square inch. This is needed to maintain the structuralintegrity of the primary containment under the conditions of large differential pressures. Therefore, the subject relief valves are considered containment relief valves and are subject to the l
testing requirements outlined in sections 1.3.4.3(a) and 3.3.2.3 of OM 1 for Class 2 and 3 L
containment vacuum relief valves.
The vacuum relief solves (four assemblies) are outside the primary containment and form the extension of the primary containment boundary. The vacuum relief valves are mounted in special piping which connects the drywell and suppression chamber. In each vacuum breaker assembly, there are two local manual butterfly valves, one on each side of the vacuum breaker,.which are provided as system isolation valves should failure of the vacuum breaker occur and as isolation valves for testing. The vacuum relief valves are instrumented with redundant position indication in the main control room. The valves are provided with the capability for local manual testing.
In accordance with the requirements of LaSalle County Station Technical Specification Surveillance Requirement 4.6.4.1, each vacuum breaker is verified to be closed at least once per 7 days, full-stroke exercised at least once per 31 days and within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after discharge of steam to the suppression chamber from the safety-relief valves.
Additionally, both of the position indicators for each valve are verified operable at least once per 31. days. In addition to the above, the force required to open each vacuum breakar, from the closed position is verified to be less than or equal to 0.5 psid and the leakage rate of each vacuum breaker valve is verified at least once per 18 months.
The 6 month operability test identified in section 1.3.4.3(a) of OM-1 refers to the open and close capability (exercise) tests, set pressure tests, and performance tests of any pressure and position sensing accessories outlined in section 3.3.2.3 of OM-1 for Class 2 and 3 vacuum relief valves. Section 1.3.4.3(b) of OM-1 specifies that valve leakage tests be performed every 2 years unless historical data indicates a requirement for more frequent testing. Additionally, since these valves are check valves, section 4.3.2.1 of OM-10 requires that the subject valves be exercised every 3 months.
A comparison of the LaSalle County Station Technical Specification surveillance requirements, OM 1 and OM-10 test requirements for the subject valves indicated that the test frequencies identified in the Technical Specification are more limiting in all instances except for the performance of the valve setpoint tests.
There are two primary methods to verify opening setpoints for these types of vacuum breaker valves. Manual exercising while measuring breakaway torque or a valve setpoint test using air. As stated above the subject valves are provided with the capability for local manual testing. However, this method was determined to be impractical because of the inconsistencies in the test data identified during preoperational testing. This test method was identified as an open item by the NRC as documented in inspection report 50-373/81028. Resolution of this issue included a commitment to perform this testing 1
using pressurized air.
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. l It is impractical to verify vacwm breaker setpoints using air every 6 months during plant operation. Verifying vacuum breaker setpoints requires the closure of the two manually operated butterfly valves upatream and downstream of the subject valves, the hookup of air supply hoses, the pressurization of large volume piping and blowdown of air inventory into the nitrogen inerted drywell. Since the two manually operated butterfly valves must be closed to perform this test, a Technical Specification Action Statement must be entered for the sole purpose of performing this test.-
l A review of the maintenance history of the subject valves tested at the 18-month Technical Specification frequency indicates that there have not been frequent failures of these velves to relieve pressure as designed. Additionally, the 18-month Technical Specification setpoint test frequoney is consistent with the two-year test frequency outlined in section i 1.3.7 of mandatory Appendix I of the 1995 Edition of the ASME OM Code.
The licensee provided the following additional information by letter dated June 12,1998, in response to the staff's questions listed below:
NRC Reouest 1:
Address whether the check valves in question (1[2]PC001A, B, C, and D) are capacity certified. If a check valve is not capacity certified, it can be classified as a check valve and tested in accordance with OM-10. If a check valve is a capacity certified valve, then it can be classified as a pressure or vacuum relief device and tested in accordance with OM 1. The valves in question are not required to be tested in accordance with both i
OM-10 (as a check valve) and OM-1 (as a vacuum relief device). This clarification is provided on page A3-31 of" Summary of Public Workshops Held in NRC Regions on Inspection Procedure 73756, ' Inservice Testing of Pumps and Valves,' and Answer to Panel Questions on Inservice Testing issues," dated July 18,1997.
Comed Response:
A subsequent review of the function of the subject valves (1(2)PC001A, B, C, and D, Drywell-to-Suppression Pool Vacuum Bre: der Valves) have been completed. As indicated [in] Reference 1 [ letter dated March 4,1998), the subject valves are capacity certified, and therefore subject to the requirements of ASME/ ANSI OM, Part 1 ((OM-1), as j
a vacuum relief device only). As stated in your [NRC's) letter, since the valves in i
question are capacity certified, they are not required to be tested in accordance with l
OM-10 (as a check valve). This is also supported by the clarification provided on page 1
l A3 31 of" Summary of Public Workshops H6td in NRC Regions on Inspection Procedure i
73756, ' inservice Testing of Pumps and Valves,' and Answer to Panel Questions on
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Inservice Testing Issues," dated July 18,1997, j
NRC Reauest 2:
For any valves that are capacity certified, provide the following information:
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7 2.1 In GL 87-09, the NRC stated its position that the structure of the technical specification accounts for entry into an LCO [ Limiting Condition for Operation) to perform surveillance testing. If the time allowed for equipment to be out of service is not sufficient to perform a surveillance test, a technical specification change requesting additional out-of service time to allow for surveillance would be required provided safety is not compromised by the increased out-of-service time.
Discuss your ressons for not requesting a techr.ical specification change in this case.
j 2.2 Address whether operators can manually manipulate one or more valves to l
restore a system to an operable status in the event the system function is required during inservice testing.
1 2.3 Expand the discussion on the impracticality of verifying the vacuum breaker y
setpoints every 6 months using the guidelines in Sections 2.4.5 and 3.1 of NUREG-1482.
Comed Responses:
2.1 As stated above, there are four vacuum breaker valves which provide redundancy for each Unit. Each of the subject valves is isolable by means of butterfly valves installed upstream and downstream of each valve and pressure taps have been designed in the piping on both sides of each vacuum breaker. Thus, with only three out of four vacuum breaker valves uvailable, system availability is not affected up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. With these provisions, the LCO time allotted for this equipment to be out-of-service in Section 4.6.4.1 of the LaSalle County Nuclear Station Technical Specification is sufficient to perform the required surveillance i
testing. Therefore, no request for additional out-of service time la necessary as outlined in Generic Letter 87-09.
2.2 in the noimal operating configuration, the drywell-to suppression pool vacuum breaker system, four valves (1(2)PC001 A, B, C, and D) are available to perform the required function. Thus, during the event that the system function is required, and with one vacuum breaker removed from service for inservice testing, testing will be halted and appropriate operator action can be taken to manipulate the applicable valves to restore the system to an operable status, to ensure system availability is not compromised.
2.3 As stated in Reference 1 (letter dated March 4,1998], we, Comed, identified that Valve Relief Request RV-05 (previously evaluated by the NRC, as identified by letter dated December 8,1995) was revised. The previous revision of RV-05 requested relief from the exercise test requirements of ASME/ ANSI OM, Part 10 (OM-10) section 4.3.2. It was determined by the NRC that the proposed testing was not a deviation frorn the Code requirements.
1 During Revision 2 of the LCNS second 10-Year interval Pump and Valve Inservice l
Testing Program Plan, the subject valves were determined to be capacity certified l
I and therefore subject to the requirements of ASME/ ANSI OM-1. As a result,
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l l RV-05 was revised to request relief from the test requirement of Section 1.3.4.3(a) l of OM 1.
l The following reinforces why it is impractical to verify vacuum breaker setpoints I
using pressurized air every 6 months during plant operation. As identified in RV-05, installation of air supply hoses and the pressurization of a large volume of -
piping, and then blowdown of this air inventory into the nitrogen inerted drywell could require a power reduction due to oxygen concentrations exceeding the required limits as specified in LCNS Technical Specification 3.6.6.2. During refueling outages, this technical specification is not required. Although these valves are not physically located in the inerted containment, the potential exists to de-inert the drywell during the performance of setpoint testing these valves at reactor power.
L Additionally, the maintenance history of the subject valves tested at the 18-month
. Technical Specification frequency indicated no frequent failures of these valves to relieve pressures as designed. Thus, the performance history of these valves to function as a vacuum relief, provides assurance and supports testing these at extended intervals without compromising safety. Setpoint testing of these valves more frequently causes unnecessary cycling of these valves and provides no additionalincrease to safety.
Lastly, the 1995 Edition of ASME/ ANSI Code, Appendix 1, has been revised to increase the frequency of setpoint testing pressure or vacuum relief devices from 6 months to 2 years. Therefore, setpoint testing at a 18-month interval is consistent with, if not more frequent, than the 2-year test frequency outlined in Section i 1.3.7 of mandatory Appendix 1 of the 1995 Edition of the ASME/ ANSI Code.
3.2 Proposed Attemative Testina The licensee proposes:
LaSalle will setpoint test these valves in accordance with the 18-month set point test frequency identified in section 4.6.4.1 of the LCNS Technical Specifications.
3.3 Evaluation The licensee proposes to verify the vacuum breaker setpoints for drywell-to-suppression pool vacuum breaker valves 1(2)PC001 A, B, C, and D evary 18 months instead of every 6 months as required by the Code. The basis provided states that the performance history of these valves to
. function as a vacuum relief provides assurance and supports testing at the extended interval without compromising rafety. Additionally,'setpoint testing every 6 months would result in a blowdown of air into the nitrogen inerted drywell during power and could require a power reduction resulting from oxygen concentrations exceeding the limits specified in LCNS Technical Specification 3.6.6.2. With regard to similar circumstances involving de-inerting of containment to allow testing, Section 3.1.1.3 of NUREG-1482 states that " valves rnay be tested during refueling outages if they would otherwise be tested during cold shutdown outages that require
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the containment to be de inerted for performance of this testing." Therefore, the proposed L
deferral of testing to refueling outages to prevant unwanted de-inerting of containment during U
power operation and cold shutdowns is consistent with the NRC's position in Section 3.1.1.3 of l
Further, the required interval of setpoint testing for Class 2 and 3 containment vacuum relief valves has been increased from 6 months to 2 years in the 1995 Edition of the ASME Code,
' Appendix 1.. In previous refety evaluations for Vogtle Electric Generating Plant, dated November 27,1996, and for Hatch Nuclear Plant, dated August 29,1995, the NRC found that implementing the portion of the 1995 Edition of the OM Code that add asses the testing of pressure relief devices is acceptable because the testing will be performed in accordance with requirements that are essentially the same as those referenced in the current regulations.
Setpoint testing at an 18-month interval meets the test frequency outlined in Section i 1.3.7, Appendix l of the 1995 Edition of the ASME/ ANSI Code that has been found to be acceptable in previous NRC safety evaluations.
The proposed altemative, therefore, provides an acceptable level of quality and safety for the following reasons: (1) the basis provided by the licensee states that the performance history of these valves to function as a vacuum relief provides assurance and supports testing at the 18-month interval without compromising safety; (2) the proposed deferral of testing to refueling
' outages to prsvant unwanted de-inerting of containment during power operation and cold shutdowns is consistent with the NRC's position in Section 3.1.1.3 of NUREG-1482, regard:ng de-inerting of containment to allow testing; and (3) setpoint testing at a 18-month interval meets the test frequency outlined in Section i 1.3.7, Appendix 1, of the 1995 Edition of the ASME/ ANSI Code that has been found to be acceptable in previous NRC safety evaluations.
3.4 Conclusion The attemative is authorized pursuant to 10 CFR 50.55a(s)(3)(i) based on the attemative providing an acceptable level of quality and safety.
4.0 CONCLUSION
Relief request RP-01 is granted pursuant to 10 CFR 50.55a(f)(6)(i) based on the impracticality of performing the required testing and in consideration of the burden on the licensee if the Code requirements were imposed. The proposed altemative testing will provide adequate assurance of the operational readiness of these pumps for performing their safety function of maintaining the ECCS piping full of water to prevent a water hammer in the event the ECCS is actuated. The altemative proposed in relief request RV-05 is authorized pursuant to 10 CFR 50.55a(a)(3)(i) r.
based on the attemative providing an acceptable level of quality and safety.
l The staff has determined that granting relief pursuant to 10 CFR 50.55a(f)(6)(i) and authorizing attematives pursuant to 10 CFR 50.55a(a)(3)(i) will not endanger life or property, or the common defense and security and are otherwise in the public interest. In making this determination, the staff has considered the impracticality of performing the required testing and the burden on the licensee if the requirements were imposed.
- Principal Contributor: K. Dempsey, EMEB Dated: July 6,1998 l
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