HL-5276, Application for Amends to Licenses DPR-57 & NPF-5,which Provides an Alternate Method of Testing S/Rvs During Shutdown Conditions Rather than During Unit Startup as Is Currently Done

Application for Amends to Licenses DPR-57 & NPF-5,which Provides an Alternate Method of Testing S/Rvs During Shutdown Conditions Rather than During Unit Startup as Is Currently Done

ML20133F291
Person / Time
Site:	Hatch
Issue date:	01/07/1997
From:	Woodard J GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20133F295	List: HL-5276, Application for Amends to Licenses DPR-57 & NPF-5,which Provides an Alternate Method of Testing S/Rvs During Shutdown Conditions Rather than During Unit Startup as Is Currently Done ML20133F364
References
HL-5276, NUDOCS 9701140138
Download: ML20133F291 (11)

Text

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Georgia Power Company 333 Piedmont Avenue Atlanta. Georgia 30308

. Telephone 404 526-3195 4 Mailing Address 40 inverness Center Parkway l

[ Post Office Box 1295 3  : Birmingham, Alabama 35201 Telephone 205 868-5086 f J. D. Woodard the southem ewtrc system I' j- Senior Vice President s-

{ January 7, 1997  ;

' Docket Nos. 50-321 HL-5276

50-366 ,

' U. S. Nuclear Regulatory Commission

i ATTN. Document Control Desk ,

Washington, D. C. 20555 Edwin I. Hatch Nuclear Plant Request for Relief From ASME Code  ;

Valve Testing Requirements ,

and l Request to Revise Technical Specifications: ,

Safetv/ Relief Valve Surveillance Testing i Gentlemen: +

In accordance with the provisions of 10 CFR 50.90, as required by 10 CFR 50.59(c)(1),

• Georgia Power Company (GPC) hereby proposes changes to the Technical Specifications  :

for Plant Hatch Units I and 2, Appendix A to Operating Licenses DPR-57 and NPF-5,  ?

respectively. The proposed changes are associated with surveillance testing that requires l manually actuating every safety / relief valve (S/RV) during each unit startup from a refueling outage. The proposed changes provide an alternate method of testing the S/RVs during shutdown conditions rather than during unit startup as is currently done. This approach will reduce valve leakage, thereby reducing the possibility ofinadvenent valve actuation and resultant plant transients. Additionally, deletion of testing for the safety t mode of the S/RVs is proposed since other testing provides operability verification. 1 i Furthermore, we seek relief from the applicable requirements of the ASME OM Code \

which also requires manual actuating of S/RVs during unit startup. The relief request is included as an attachment to this letter. i I

Enclosure 1 provides a description of the proposed changes and a justification for each change. Enclosure 2 details the basis for GPC's determination that the proposed changes '\ j do not involve a significant hazards consideration. Enclosure 3 provides page change ,

instructions for incorporating the proposed Technical Specifications changes. Following

]

Enclosure 3 are the revised Technical Specifications pages and the corresponding i marked-up pages. Enclosure 4 provides the associated Bases pages for your information. )

l 9701140138 970107 PDR ADOCK 05000321 P PDR

i l

~ Georgia Power 1 l

?

U. S. Nuclear Regulatory Commission Page 2 January 7, 1997 Since these changes affect testing performed during unit startup and require plant procedure revisions, GPC requests approval of the proposed amendments as soon as possible. GPC also requests that the amendments be made efTective within 30 days from -

the date ofimplementation In accordance with the requirements of 10 CFR 50.91, the designated State official will be sent a copy of this letter and all applicable enclosures.

Mr. J. D. Woodard, states he is Senior Vice President of Georgia Power Company and is  !

authorized to execute this oath on behalf of Georgia Power Company, and to the best of his knowledge and belief, the facts set forth in this letter are true.

Sincerely, ,

- ., ~ . ,

s J. 8 00 ard N3 i Sv andsubscribedbefore me this f dayof si;ad 1997. .

,( svtf~  !

flic~ ~

My ommission Expires: 9-/f-78 OCV/eb

Enclosures:

1. Description of and Justification for Proposed Changes

2. No Significant Hazards Determination

3. Page Change Instmetions and Revised Technical Specifications Pages

4. Revised Bases Pages

Attachment:

Valve Relief Request RR-V-11 cc: (See next page.)

HL-5276

2

GeorgiaPower A i

U. S. Nuclear Regulatory Commission Page 3 January 7, 1997 1

l cc: Georgia Power Company Mr. H. L. Sumner, Nuclear Plant General Manager NORMS U. S. Nuclear Regulatory Commission. Washington. D. C. ,

4 Mr. K. Jabbour, Licensing Project Manager - Hatch

• U. S. Nuclear Reentatory Commission. Reeton H Mr. S. D. Ebneter, Regional Administrator Mr. B. L. Holbrook, Senior Resident Inspector - Hatch State of Georgia

, Mr. J. D. Tanner, Commissioner - Department of Natural Resources i

d s

i 6

5 ,

HL-5276

Enclosure 1 Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications:

Safety / Relief Valve Surveillance Testing Description of and Justification for Proposed Changes

Background

Georgia Power Company, as well as other utilities through BWR Owners Group activities, has studied the problem of safety / relief valve (S/RV) leakage. The consensus reached is that several aspects of S/RV design and operation; i.e., simmer margin, reseat margin, testing pressure, pilot valve disc and rod configuration, and system and valve cleanliness, contribute to valve leakage.

The following actions are completed or underway to decrease SRV pilot leakage:

• New Target RockAVyle procedures to check and maintain tighter disc / seat tolerances.

• Revised Wyle procedures to perform final leakage tests at higher pressure.

. Increased focus on cleanliness during all phases of SRV removal, rework, testing, transportation, and reinstallation.

• A Technical Specifications change is currently being reviewed by the NRC which will raise mechanical setpoints to increase simmer margin.

This submittal, which will eliminate the requirement to manually actuate the valves at pressure, is expected to have a positive impact on the above listed contributors, resulting in reduction of S/RV leakage.

Description of Proposed Chanees Current Unit I and Unit 2 Technical Specifications Surveillance Requirements (SRs) 3.5.1.12 and 3.6.1.6.1 require that each S/RV be manually actuated at pressure conditions. GPC proposes to revise SRs 3.5.1.12 and 3.6.1.6.1 to require the S/RVs to be manually actuated in the relief mode during the plant outage before steam is generated. The solenoid valve will be energized, the actuator will stroke, and the pilot rod lifl will be measured. This in-situ test will verify that, given a signal to the solenoid, the pilot disc rod will lifl. If steam were present, the pilot disc would open and initiate opening of the main stage.

GPC also proposes to delete current Unit I and Unit 2 SR 3.4.3.2, which also requires that each S/RV be manually actuated. Ho Never, as discussed further in thejustification portion of this HL-5276 El-1

' j Enclosure 1 Discussion of Change and Justificat;on submittal, this test is not necessary to assure S/RV operability in the safety mode since it does not confirm operability of the safety mode and since other tests, taken together, confirm the entire S/RV assembly functions adequately.

Currently, these tests are performed on an 18-month Frequency. This verification is performed for the S/RV safety mode (SR 3.4.3.2), the Automatic Depressurization System (ADS)

(SR 3.5.1.12), and the Low-Low Set (LLS) system (SR 3.6.1.6.1). The tests are performed during reactor startup from refueling when reactor pressure and flow are adequate.

Additionally, GPC seeks relief from ASME OM Code (1995), Appendix I, paragraph I 3.4.l(d),

which also requires that the S/RVs be manually actuated at pressure conditions. The relief j request is provided as an attachment to this letter.

i Justification For Proposed Changes 1

Before elaborating on thejustification for the changes, a brief description of the S/RV system and l the current testing performed on the valves is desirable:

System Description

The Unit I and Unit 2 S/RVs are the Target Rock Two-Stage, Model 7567F design. The S/RVs I are dual-function valves capable of being independently opened in either the safety or the relief mode of operation. A total of 11 S/RVs are installed on each unit.

In the safety mode of operation, each S/RV opens when system pressure exceeds the valve's preset setpoint pressure, which is controlled by precompression of the setpoint spring acting down on the pilot disc. Venting the volume behind the pilot disc creates a differential pressure across the main piston, thereby providing a force to open the main disc and relieve system overpressure.

Hence, reactor vessel steam is allowed to flow directly through the main disc to seat opening and to the suppression pool via the discharge piping. All 11 S/RVs operate in the safety mode, which provides the safety function of overpressurization protection. The requirements for this mode are listed in Technical Specification 3.4.3.

In the relief mode of operation, each S/RV is opened by an electro-pneumatic actuator, which consists of a three-way solenoid valve, an attachment manifold, and a pneumatic operator. When the solenoid valve is energized, pneumatic pressure is routed into the operator to lift the pilot rod against the force of the compressed setpoint spring. This allows system pressure to lift the pilot disc, venting the volume behind the disc, and opening the valve as in the safety mode discussed above. This mode of operation is used for ADS, LLS, and remote manual operation. Technical Specifications 3.5.1 and 3.6.1.6 provide requirements for the ADS and the LLS System. Manual operation is not safety-related and is not addressed by Technical Specifications. In each unit, seven S/RVs are part of ADS, while the remaining four constitute LLS.

HL-5276 El-2

Enclosure 1 l Discussion of Change and Justification l

Current Testing at Plant Hatch Testing of Plant Hatch S/RVs is performed to satisfy Technical Specifications Surveillance Requirements and the ASME OM Code (1995)," Code for Operation and hiaintenance of Nuclear Power Plants." Certain tests are performed with the S/RVs installed (in-situ), while others are l performed as " bench tests" after the valve is removed and transported to a maintenance and testing facility. Current requirements are as follows:

1. Plant Hatch Units 1 and 2 Technical Specifications SR 3.4.3.2 requires that each S/RV be opened by manual actuation to demonstrate that the S/RV safety mode is operable. This is accomplished by showing that mechanically the valve is functioning properly and no blockage exists in the discharge piping. SR 3.4.3.2 is performed on an 18-month Frequency during reactor startup from refueling at a reactor pressure of at least 920 psig.

2. SRs 3.5.1.12 and 3.6.1.6.1 provide similar S/RV manual actuaticn testing for the ADS and LLS Functions, respectively, to demonstrate operability of the S/RV relief mode.

3. Remote manual actuation is also required by the ash 1E OM Code, Appendix I, paragraph 3.4. l(d), to verify open and close capability of the valve before resumption of electric power generation. This applies to valves that were either maintained or refurbished in place, or removed for maintenance and testing and reinstalled. The remote manual actuation is performed at reduced or normal system pressure.

Plant Hatch currently meets these testing requirements by opening and closing each S/RV using control room switches. Verification of valve opening and closing is made by monitoring the valve discharge piping temperature and/or pressure.

4. Plant Hatch Units I and 2 Technical Specifications SRs 3.5.1.11 and 3.6.1.6.2 require that the ADS and LLS S/RVs be opened on an actual er simulated automatic initiation signal to demonstrate that the solenoids operate when initiated by a signal. Actual valve actuation is excluded from these tests which are performed on an 18-month Frequency.

Plant Hatch currently meets these testing requirements by performing the test in conjunction with Logic System Functional Tests for the initiating instrument logic, which are also required by Technical Specifications.

5. ASME OM Code (1995) 13.3.1 (d) and (e) require that S/RV auxiliary components be tested in place as follows: solenoid valve and pneumatic actuator integrity is verified by performance ofleak rate tests, and solenoid valve electrical function is verified. These tests are performed following maintenance on the valves and together demonstrate operability of the valve pneumatic actuation system.

HL-5276 El-3

Enclosure 1 Discussion of Change and Justification Current Testing at Outside Facilities During each refueling outage, which occurs on an 18-month Frequency, all 11 S/RV pilot assemblies and approximately 25% of the main stages are removed and shipped to Wyle Laboratories for "as-received" testing, which includes visual inspection, leakage testing, pilot disc-to-seat sticking testing, and set pressure testing, all of which are performed prior to any maintenance on the valve. The leakage test and set pressure test are performed at a steam pressure of approximately 1010 psig. Both tests meet the requirements of ASME OM Code (1995) I 3.3. l(a), (b), and (c).

Following the "as-received" testing, the S/RVs are given a dimensional inspection followed by refurbishment if required. This work is performed by the valve supplier, Target Rock Corporation.

Following valve refurbishment, post maintenance testing is performed at a steam pressure of approximately 1010 psig. This includes initial valve leakage testing; safety mode valve actuation to satisfy requirements for set pressure, reseat pressure, and main seat stroke time; and final leakage testing. Seat leakage tests are performed at approximately 1045 psig. Upon successful test completion, the valve receives written certification from the lab and is returned to Plant Hatch for reinstallation. To receive certification, the valve must have zero seat leakage and meet the acceptance criteria for set and reseat pressure. These tests help meet the requirements of ASME OM Code (1995) I 3.3.1 and Technical Specifications SR 3.4.3.1 (for lifl setpoint pressure verification). l l

General Change Justification j Leaking S/RVs result in the following challenges to Plant Hatch components and operation:

1. Extreme leakage during operation may cause the valve to inadvertently actuate, likely resulting in an unplanned plant shutdown, with its attendant challenges to plant safety systems and components. This has occurred recently at a domestic BWR plant.

2. Leaking S/RVs create operational problems with the suppression pool. S/RV leakage increases both pool temperature and level, requiring more frequent use of the suppression pool cooling mode of the Residual Heat Removal (RHR) system.

3. Plant efliciency is impacted because the transfer of heat to the suppression pool is a source of thermal heat loss from the power generatian steam cycle, thereby reducing electrical generating capacity. S/RV leakage results in radiological challenges since steam is an additional source of radioactive nuclides that become a potential source for personnel contamination.

As described previously, each S/RV pilot assembly and approximately 25% of the main stage valves are bench tested at Wyle Laboratories during each refueling outage. The valves are HL-5276 E l-4

Enclosure 1

Discussion of Change and Justification refurbished as necessary to meet the acceptance criteria of zero leakage, and are certified in writing as being leak free. The valves are reinstalled in the plant and remotely opened at a system i pressure of at least 920 psig. Following this surveillance test, Plant Hatch typically experiences  !

severalleaking valves from what was originally a leak-free population. For example, in order to

{

prevent the possibility of an inadvertent actuation, Plant Hatch was shutdown in March of 1996 as a result of a leaking S/RV. During this forced outage, two S/RVs were replaced with leak-tight valves, which were then manually actuated, during startup, at or above 920 psig. One of the S/RVs immediately began leaking following startup.

Also, the requirement for manual actuation during plant startup creates the need to progress in startup to a point where the S/RVs are required to be operable, but are not proven operable. The design of the Plant Hatch S/RVs dictates that a manual actuation of the valves be performed only when sufficient steam pressure is present to " cushion" the valve when it reseats. Otherwise, ,

significant valve damage may result. Therefore, the valve actuation may only be performed after the plant has progressed in startup to the point where steam pressure is sufficient. However, in order to achieve this, the unit has to move into a mode of operation requiring operable S/RVs before the test demonstrating operability can be performed. This is an exception to normal Technical Specifications operability requirements for mode changes, but is recognized as necessary by a Note of exception given in the Technical Specifications.

Several aspects of S/RV design and operation contribute to valve leakage. As mentioned earlier, I these include simmer margin, reseat margin, testing pressure, pilot valve disc and rod configuration, and system and valve cleanliness. The manual actuation of the S/RVs at pressure )

allows these contributors to impact the ability of the valve to re-close completely. As stated previously, efTorts not requiring prior approval are being made in parallel with this submittal to minimize the effects of these contributors. Also, a Technical Specifications revision request is presently being reviewed by the NRC to increase the simmer margin. However, elimination of valve testing at pressure conditions is expected to have the most positive impact in reducing the S/RV leakage. j Additionally, reducing challenges to the S/RVs is a recommendation of NUREG 0737, "TMI j Action Plan Requirements" item II.K.3.(16). The recommendation is based on a stuck open S/RV  ;

being a possible cause of a Loss of Coolant Accident. This submittal is consistent with that j recommendation.  :

Specific Change Justification As an alternate for SRs 3.5.1.12 and 3.6.1.6.1, GPC proposes to actuate the S/RVs in the relief mode before steam is generated. The solenoid valve will be energized, the actuator will stroke, and the pilot rod lif1 will be measured. This in-situ test will verify that, given a signal to the solenoid, the pilot disc rod willlifl. If steam were present, the pilot disc would open and initiate opening of the main stage.

HL-5276 El-5

Enclosure 1 Discussion of Change and Justification i Alternate testing isjustified since the remaining segments of the S/RV mode of operation are ,

proven by other tests. The ability of the pilot disc to open is shown in the safety mode actuation l

bench test. The integrity of the solenoid and pneumatic system for the S/RV is verified by 4

perfoimance ofleak rate tests post maintenance, and the electrica! function of the solenoid is verified. Automatic valve actuation is proven operable by logic system functional tests which l 4

include verification that the solenoid actuates from the automatic signal.

Each refueling outage all 11 pilot assemblies and three or four main disc assemblies are sent to Wyle Labs and tested with system pressure. As a result, even though actual valve movement is

not performed after the S/RV is re-installed in the plant, all pilot assemblies are tested with system pressure once per cycle and all the main discs are tested with system pressure approximately once every three cycles. This fact, together with the current and proposed S/RV testing, completely demonstrates operability of the valves.

GPC also proposes to delete SR 3.4.3.2, which requires that the S/RVs be manually actuated to

, demonstrate operability of the safety mode. However, manual actuation can only be performed ,

using the pneumatic actuation system: i.e., the relief mode. The pneumatic system provides I suflicient force to overcome the safety mode setpoint spring downward force and lift the pilot i disc, allowing the disc to lift and open the valve. Therefore, the ability of system pressure alone i

to overcome the spring force is not demonstrated. It follows that this surveillance is not verifying safety mode operability.

The ASME OM Code (1995) requires set-pressure determination, and Plant Hatch Technical Specifications SR 3.4.3.1 requires verification of the safety function lift setpoints. While the S/RVs are being bench tested at Wyle Labs, the pilots are opened using system pressure. These i activities are suflicient to verify safety mode operability.

The Technical Specifications Bases for the remote actuation of each S/RV state that one purpose of the test is to verify that no blockage exists in the valve discharge line. Plant Hatch has foreign l material exclusion controls in place on all openings when S/RVs are removed and reinstalled.  !

These controls, as well as the horizontal orientation of the S/RV discharge line mating surfaces, provide reasoncble assurance that no obstruction will be admitted into the S/RV discharge tailpipe. Additionally, in the history of replacing S/RVs during each refueling outage, Plant Hatch has not experienced a single surveillance failure related to line blockage.

HL-5276 El-6

Enclosure 2 1

Edwin I. Hatch Nuclear Plant l Request to Revise Technical Specifications:

Safety / Relief Valve Surveillance Testing No Sigpificant Hazards Determination 1

In 10 CFR 50.92(c), the NRC provides the following standards to be used in determining the existence of a significant hazards consideration:

.a proposed amend:nent to an operating license for a facility licensed under 50.21(b) or 50.22 for a testing facility involves no significant hazards consideration ifoperation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of an accident of a new or different kind from any previouly evaluated; or (3)

Involve a significant reduction in the margin of safety.

Georgia Power Company has reviewed the proposed license amendment request and determined its adoption does not involve a significant hazards consideration. In support of this determination, l an evaluation of each of the three 10 CFR 50.92 standards follows.

1. The proposed changes do not im'oh e a significant increase in the probability or consequences ofan accident previously erahiated.

Since the proposed Technical Specifications changes and ASME Code relief do not impose any physical changes to the S/RVs, their design function is unaffected. The submittal only proposes changes to the manner in which the S/RVs are tested. As discussed in Enclosure 1, the combination of current S/RV testing and the proposed alternate testing will continue to adequately demonstrate the operability of the S/RVs for both the safety and relief modes.

Under the proposed testing requirements, it is expected that S/RV leakage will decrease; thus, the probability of occurrence of an inadvertent S/RV actuation is actually reduced.

FSAR analyzed events, such as MSIV closure, generator load reject, turbine trip with failure of switchyard breakers to open, and pressure regulator failure, take credit for the S/RVs mitigating the consequences of these events. These proposed changes will not increase the consequences of these events, since a series of S/RV tests (on the bench and installed) will ensure all S/RV components necessary to ensure valve opening will function. The S/RVs will therefore be capable of performing their design functions.

Furthermore, reducing the number of manual actuations of the S/RVs decreases the likelihood of a stuck open S/RV, which is an analyzed event in the Hatch FSAR.

HL-5276 E2-1

Enclosure 2 No Significant Hazards Considerations Therefore, the probability of occurrence and the consequences of previously analyzed events are not increased.

2. The proposed changes do not create the possibility ofan accident ofa new or different kind from anypreviously evahiated.

The proposed changes affect the manner in which S/RV operability is verified in that one Technical Specifications SR is being deleted and two are being revised; however, they do not affect the way the S/RVs are operated. The S/RVs will not be operated or tested in a manner contrary to their design. As a result, no new mode of operation is introduced. That is, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. The proposed changes do not involve a sigmficant reduction in the margin ofsafety.

The present method of S/RV testing unnecessarily challenges the valves, and is linked to S/RV degradation through pilot valve and/or main valve leakage. This Technical Specifications change should decrease S/RV leakage and improve S/RV reliability by reducing the potential for spurious valve actuation at full power. In this sense, the margin of safety is actually increased; e.g., the likelihood for spurious S/RV actuation is reduced.

Deleting the test ofinstalled S/RVs at rated temperature and pressure will not significantly i reduce the margin of safety for events in which'S/RV actuation is assumed, since each S/RV l receives a series of tests which insure each component necessary for successful opening of the S/RV functions properly. Thus, the S/RV is assured of opening in either the safety or the I relief mode. For example, at Wyle Labs, the valves undergo testing at operating steam pressure. This test ensures operability of the pilot and main discs and also verifies set pressure, reseat pressure, and main steam stroke time. As noted previously, upon successful completion of these tests, including verification of zero seat leakage, the valves receive a written certification from the lab and are returned to Plant Hatch for installation.

GPC further proposes that, upon installation, but before steam is generated, the valves receive a test requiring the solenoid to be energized. This test provides additional verification that the pilot disc opens. The remaining segments of the S/RV tests verify the ability of ADS and LLS logic to energize the solenoid.

In summary, this amendment does not involve a significant reduction in the margin of safety, because of the reduction in S/RV degradation, and because remaining tests confirm the valves will function properly when required.

HL-5276 E2-2