ML20100J831

Application for Amend to License NPF-29,modifying Safety Setpoint to Tolerance Requirements for Safety/Relief Valves

ML20100J831
Person / Time
Site:	Grand Gulf
Issue date:	02/22/1996
From:	Meisner M ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20100J816	List: ML20100J813 ML20100J831 ML20100J839 ML20100J862
References
NUDOCS 9602290269
Download: ML20100J831 (14)

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GNRO 96/00008 Att: chm:nt 1 Page 1 of 1 BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION LICENSE NO. NPF-29 DOCKET NO. 50-416 IN THE MATTER OF MISSISSIPPI POWER & LIGHT COMPANY and SYSTEM ENERGY RESOURCES, INC, and SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION and ENTERGY OPERATIONS, INC.

AFFIRMATION I, M. J. Meisner, being duly sworn, state that I am Director, Nuclear Safety and Regulatory Affairs, GGNS of Entergy Operations, Inc.; that on behalf of Entergy Operations, Inc., System Energy Resources, Inc., and South Mississippi Electric Power Association I am authorized by Entergy Operations, Inc. to sign and file with the Nuclear Regulatory Commission, this application for amendment of the Operating License of the Grand Gulf Nuclear Station; that I

- signed this response as Direc r, Nuclear Safety and Regulatory Affairs, GGNS of Entergy Operations, Inc.; and that t stateme ts made and the matters set forth therein are true and correct to the best of my k le formatiogand belief.

- 1 41, ner STATE OF MISSISSIPPI COUNTY OF WARREN SUBSCRIBED AND SWORN TO before me, a tjotary Public, in and for the County and State above named, this 22

• day of Feeruarr ,1996.

l f y ~/o

~\ Notary Public MsINdNENN ks NT COMMISSION EXPIRES JUNE 5. I BONDED TliRU 8TEGAll NOTARY SERVICE 9602290269 960222 PDR ADOCK 05000416 P PDR

i GNRO 96/00008 Attachment 2 Page 1 of 13 Detailed Description of the Proposed Chanae, Justification, and the No Sianificant Hazards Considerations DISCUSSION History Licensing Topical Report (LTR), NEDC-31753P, was submitted by the BWROG and the General Electric Company (GE) to request generic approval of an increased tolerance for the safety relief valve (S/RV) safety function setpoint. A Safety Evaluation Report (NEDC-31753-P-A SER) has been issued by the NRC on this LTR, and indicates that a generic change of S/RV setpoint tolerance to 13%'is acceptable, provided certain plant specific analyses are performed.

System Description

Each Grand Gulf Nuclear Station (GGNS) S/RV is a Dikkers,8 X 10, direct-acting, spring loaded, safety valve with attached pneumatic cylinder for relief mode i operation. Each S/RV can perform its intended function through two modes of operation:

The safety mode of operation which consists of direct action of the steam pressure against a single spring-loaded disk that will open when the valve inlet pressure force exceeds the spring force. The safety function set pressure of each S/RV is determined by changing the value of the compressed spring force.

The relief mode of operation which consists of using an auxiliary actuating device consisting of a pneumatic piston / cylinder and mechanicallinkage assembly which opens the valve by overcoming the spring force.

GGNS has a total of twenty S/RVs installed on the four main steam lines. All valves are of the same design, and each valve can be operated in either the safety or the relief mode, j

The GGNS Inservice Testing (IST) program and associated plant procedures performs testing on each S/RV in accordance with ASME Boiler and Pressure Vessel Code,Section XI,1980 Edition, with Addenda through Winter 1980 (Subsections IWA and IWV). Each S/RV that is removed during a refueling outage is tested on a test fixture, and is certified for relief mode operation and seat leakage. Also, the S/RV is tested for safety mode set pressure, to ensure that the safety function lift setpoint is within 11% of the set pressure, as required by the GGNS Operating License. Presently, GGNS removes all twenty of the S/RVs each

GNRO 96/00008 Attachment 2 Page 2 of 13 outage to prevent a potential refueling outage extension as a result of setpoint drift outside of the 11% tolerance.

PROPOSED TS CHANGES GGNS proposes increasing the present 11% tolerance on the safety function lift setpoint for the S/RVs to 13%. This change would affect Technical Specification Surveillance Requirement 3.4.4.1. The GGNS Inservice Testing (IST) program controls the frequency of S/RV testing as required by the GGNS Operating License; thereforo, this proposal will also incorporate changes to applicable IST procedures. GGNS willincorporate the recommendations of the NEDC 31753-P-A SER, by resetting the safety function lift setpoints for all tested valves to within 11% of the design lift setpoint and increasing the test sample size by two valves for each valve found outside of the 13% safety function lift setpoint. GGNS will determine the number of S/RVs to be tested each outage using the currently licensed ASME Boiler and Pressure Vessel Code,Section XI,1980 Edition, with Addenda through Winter 1980 (Subsections IWA and IWV), rather than the prescribed 50% of allinstalled valves as recommer.ded by the NEDC-31753-P-A ,

SER. This proposed change does not affect the relief mode of S/RV operation, nor its setpoints.

JUSTIFICATION FOR CHANGES NEDC 31753-P-A SER Reauested Analyses Licensing Topical Report (LTR), NEDC-31753P, was submitted by the BWROG to request generic approval of an increased tolerance for the S/RV safety function setpoint. A Safety Evaluation Report (NEDC 31753-P-A SER) has been issued by the NRC on this LTR, and indicates that a generic change of S/RV setpoint tolerance to 13% is acceptable, provided certain plant specific analyses are performed. Each of the requested analyses, along with a summary of the GGNS specific evaluaticn, follows. Specific details of each item are provided in Attachment 4,

1. Transient analysis of all abnormal operational occurrences as described in NEDC-31753P, should be performed utilizing a 13% setpoint tolerance for the safety mode of spring safety valves (SSVs) and SRVs. In addition, the standard reload methodology (or other method approved by the staff) should be used for this analysis.

NEDC-31753P, Section 4.3, states that group 3 plants (BWR 5 & 6 design) are not affected by the evaluation of abnormal operational occurrences since this program only proposes changes to the safety mode of actuation. However, GGNS abnormal operating occurrences

GNRO 96/00008 Attachment 2 Page 3 of 13 that result in S/RV actuation credit operation of seven safety mode S/RVs, the minimum number required by the GGNS Operating License.

Therefore, it was necessary for GGNS to determine if the proposed setpoint tolerance request would affect any of the previously analyzed abnormal operating occurrences. Each of these abnormal operating occurrences were analyzed using safety function lift setpoints above the r

' quested 13%. None of the analyses result in steam dome pressures exceeding the ASME pressure vessel code limit of 1375 psig. These analyses were performed using NRC approved methodology.

2. Analysis of the design basis overpressurization event using the 3%

tolerance limit for the SRV setpoint is required to confirm that the vessel pressure does not exceed the ASME pressure vessel code upset limit.

The GGNS design basis (worst case) overpressurization event is a i closure of all main steam isolation valves while the reactor is operating ;

at 100% rated power and 105% rated core flow. Reactor scram on i MSIV position is assumed to fail, so the scram is assumed to occur on high neutron flux. The BWR 6 design meets the ASME Section til i Article NB 7542 allowance that up to half of the installed S/RVs may I take credit for the auxiliary actuating device (relief mode); however, it should be noted that the GGNS analysis only credits six of the installed twenty S/RVs for actuation in the relief mode.

The design basis overpressurization analysis for the upcoming fuel cycle (Cycle 9) has not been performed. This analysis will be performed as part of the normal reload analysis process. For cycle 9, this analysis will be performed by GE using the NRC approved ODYN methodology described in GESTAR-II, NEDE-24011-P-A.

Overpressurization analyses performed for previous fuel cycles using SRV opening pressures in excess of the13% showed considerable margin (-100 psi) to the ASME Boiler and Pressure Vessel Code limit of 1375 psig. Based on these results and the relative insensitivity of the results to the fuel design parameters, future analyses including that planned for Cycle 9, are expected to yield peak pressures with similar margin to ASME Boiler and Pressure Vessel Code limit. As required by the NEDC 31753-P-A SER, future reload safety analyses will bound the proposed 13% tolerance.

3. The plant specific analysis described in items 1 and 2 should assure that the number of SSVs, SRVs, and relief valves (RVs) included in the analysis correspond to the number of valves required to be operable in the technical specification.

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I GNRO 96/00008 Attachment 2 Page 4 of 13 i

The number of S/RVs assumed in the analyses required in items 1 and 2 above is consistent with the Operating License, Limiting Condition of Operation 3.4.4, by crediting operation of only seven safety mode S/RVs and six relief mode S/RVs.

l l 4. Re-evaluation of the performance of high pressure systems (pump capacity, i discharge pressure, etc.), motor-operated valves, and vessel l instrumentation and associated piping must be completed, considering the 3% tolerance limit.

Hiah Pressure Systems l' GGNS has three high pressure vessel injection / spray systems: 1) high pressure core spray (HPCS),2) reactor core isolation cooling (RCIC),

and 3) standby liquid control (SLC).

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1) HPCS is an emergency core cooling system designed to deliver j sufficient coolant to the reactor core in conjunction with other l ECCS systems. HPCS is designed to deliver a minimum water )

l flow of 550 gpm to the reactor vessel with reactor vessel pressure l l 1177 psi above the pressure source of pump suction. A review of )

l the HPCS preoperational test data indicates that the system has

! sufficient margin to deliver flow in excess of 550 gpm at the l anticipated elevated vessel pressure. The design discharge pressure of the HPCS pump is 1575 psig, and the discharge ,

piping from the pump to the reactor vesselis designed for 1575 l psig. A relief valve with a setpoint of 1560 psig is provided for protection against system overpressure. These pressures are well above the pressures that may result from the S/RV setpoint tolerance relaxation; therefore, adequate margin is maintained.

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2) The RCIC system is not credited in the GGNS vessel pressurization safety analyses; therefore, the present system operating parameters do not require to be changed based on this

j. proposal. The evaluation reviewed the effects of the proposed change on the system flow, leak detection isolation flow setpoint, l l overspeed trip setpoint, and initiation time to rated flow. The

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pump design discharge pressure is 1525 psig, and the discharge piping has a design pressure of 1500 psig. The inlet piping to the RCIC turbine is rated at the same pressure as the reactor vessel.

These pressures are above the pressures that result from the S/RV setpoint tolerance relaxation; therefore, adequate margin is retained. Although the RCIC system was determined to provide a flow rate slightly lower than the original capacity of 800 gpm while operating at the elevated pump discharge pressure associated with the proposed safety setpoint tolerance, the system will

GNRO 96/00008 Attachment 2 Page 5 of 13 continue to perform its intended design functions as described in Attachment 4.

3) The SLC system operation is not affected by the S/RV safety setpoint tolerance increase. The pressure used for system performance is based on the S/RV relief settings of the system, not the S/RV safety settings. This proposed change does not affect the relief settings of the S/RVs.

Motor-operated Valves (MOVs)

As described in NEDC-31753-P-A and the Technical Evaluation Report prepared as part of the NRC's evaluation, consideration should be given l to testing MOVs exposed to reactor pressure at higher differential J pressures. Because MOV dynamic testing is done at the highest differential pressure achievable under normal operational l configurations, dynamic testing parameters are unaffected by the S/RV l safety setting tolerance increase. However, MOV operator settings for static testing are also based on the calculated maximum expected l differential pressure (MEDP) values (as one of the input parameters for I determining required settings).

l Adequacy of MOV settings was assessed by evaluating the adequacy of the MEDP calculation assumptions and resulting MEDP values  ;

established for those MOVs potentially affected by the increase in S/RV safety setpoint tolerance. MEDP calculations for several MOVs at i GGNS were identified that use a generically derived reactor pressure based solely on the safety function lift setpoint plus 1% tolerance.

GGNS is a group 3 (i.e., BWi 5/G) plant as defined in NEDC-31753-P-A, has dual mode S/RVs (relief and safety), and credits the safety-grade relief mode of operation in the analyses of abnormal operational occurrences and the ASME overpressurization analysis. An evaluation that included reliance on both the relief mode and safety mode S/RV functions was performed to assess the impact of the 3% tolerance on the MEDP calculations. This evaluation determined that the use of reactor pressure based solely on the S/RV safety mode setpoint plus the 1% tolerance as the basis for the calculated MEDP remains a l bounding assumption.

l Vessel Instrumentation The design pressure of process piping is adequate to provide margin above the pressure resulting from an increase in S/RV setpoint tolerance. Each instrument potentially affected by the proposed change l

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GNRO 96/00008 Attachment 2 Page 6 of 13 was evaluated for the effects on pressure boundary integrity, instrument calibration, instrument scaling calculations and instrument setpoint/ uncertainty calculations. The evaluation determined that there is no impact on plant instrumentation as a result of the proposed safety setpoint tolerance change.

5. Evaluation of the 13% tolerance on any plant specific alternate operating modes (e.g., increased core flow, extended operating domain, etc.) should be completed.

Maximum Extended Operatina Domain (MEOD)

All transient and accident analyses performed in support of the Cycle 8 reload incorporated the extended load line and increased core flow

, available under the GGNS MEOD. Rated analyses have been

, performed at 104.2% rated power and 108% rated flow, which bounds the 100% power,105% flow combination possible under MEOD. This analysis was performed using safety function setpoints that exceeded the requested 13%.

Sinale Loop Operation (SLO)

GGNS operation with a single recirculation loop out of service for a period of time has been analyzed by GE. This analysis included reviews and re-analysis of the applicable transients. These analyses concluded that since the plant cannot reach rated bundle powers in SLO due to the reduced core flow capabilities of a single loop, the three most limiting transients are (per UFSAR Section 15C.3.1):

feedwater controller failure (maximum demar.d),

load reject no bypass, and pump seizure.

As discussed in UFSAR Section 15C.3.1 for cycle 6, as a result of GE's i analyses, Siemens Power Corporation has determined that the pump k seizure accident is a limiting event at GGNS during SLO. The results of l this analysis indicate that no plant parameters will be exceeded as a

! result of the proposed change to the safety function setpoint tolerance.

Feedwater Heaters Out of Service (FWHOS) i GGNS is licensed to operate with FWHOS with feedwater temperatures as low as 370' F. As discussed in UFSAR Section 15.1.2.3.3, the l

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GNRO 96/00008 Attachment 2 Page 7 of 13 feedwater controller failure (FWCF) with FWHOS was found to be bounded by the FWCF event without bypass. This conclusion was re-stated in the Cycle 8 fuel reload safety analysis. The FWHOS operational mode is therefore bounded by previously analyzed abnormal operating occurrences as described in item 1.

6. Evaluation of the effect of the 3% tolerance limit on the cor'tainment response during loss of coolant accidents and the hydrodynamic loads on the SRV discharge lines and containment should be completed.

Containment Response Durina Loss of Coolant Accidents (LOCA)

The most limiting event in terms of peak containment pressure and temperature and peak suppression pool temperature is the design basis accident (DBA) LOCA, a double ended guillotine break of the steam line. Relaxation of the S/RV setpoint tolerance has no effect on this event because the vessel depressurizes without any S/RV actuations.

Therefore, there is no impact on the DBA-LOCA peak containment pressure and temperature and on the peak DBA LOCA suppression pool temperature.

S/RV Discharae Line Loads The S/RVs were analyzed in three sections: 1) S/RV to the first anchor point,2) discharge line downstream of the S/RV anchors, and 3) quenchers.

1) An increase in the S/RV safety mode open setpoint tolerance to 3% from the current value of 1% will result in an increase in the S/RV opening discharge flow rate into the S/RV discharge line.

This in turn results in an increase in the loads associated with S/RV openings. An evaluation showed that the GGNS-specific S/RV discharge line loads were lower than the BWR generic S/RV discharge line loads with 1% setpoint tolerance, due to existing I conservatisms in the generic S/RV discharge line loads. Based on ,

the results of the evaluation, a relaxation of the safety setpoint i pressure to 13% will not result in any allowable stresses being l exceeded in the SRV discharge line piping and supports between the S/RV and the first anchor point.

l 2) The existing analysis of the discharge line downstream of the S/RV anchors was reviewed for margin and original assumptions.

It was determined that increasing the S/RV setpoint tolerance from 1% to 3% does not affect the existing analysis of the S/RV discharge loads for eight of the twenty lines. The remaining twelve lines will see less than or equal to 0.7% increase in relief valve

GNRO 96/00008 Attachment 2 Page 8 of 13 thrust loads. The net increase in pipe stress, quencher nozzle loads and support loads due to the 3% setpoint tolerance will be negligible.

3) GGNS uses the standard GE X-quencher design. The generic loads defined for the quencher were found to be significantly greater than the GGNS-specific quencher loads, due to the conservatism of the GE design. Therefore, the loads were determined to be unaffected by the relaxation of the S/RV safety mode tolerance to 13%.

Containment Hydrodynamic Loads Loads on the submerged boundary and on submerged structures are based on the peak bubble pressure determined with the GE X-quencher methodology. Based upon the results of the evaluation, there is no affect on the GGNS S/RV pool boundary load definition, and loads currently defined for the GGNS submerged structures are also not affected by the relaxed setpoint tolerance.

NEDC 31753-P-A SER Conclusions and Limitations The SER for NEDC-31753P provided four conclusions and limitations. The GGNS response to each conclusion and limitation follows:

a. A generic change of valve setpoint tolerance to13% is acceptable.  !

This proposal agrees with this proposed tolerance.

b. The staff concludes that the philosophy of an upper limit is not acceptable as a means to further reduce the number of LERs, and that an evaluation to determine the necessity for filing an LER must be made for setpoints when drift outside 13% is found.

This proposal agrees with this approach. GGNS will continue to determine the necessity of filing an LER consistent with past practices.

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c. The recommendation to modify the TS to classify a valve as operable with the setpoint of a valve (s) outside of the 3% tolerance is not technically justified by the LTR or by the supplemental data, and therefore, not acceptable.

This proposal agrees with this statement.

GNRO 96/00008 Attachment 2 Page 9 of 13

d. Modification of the current requirement for spring safety valves (SSV) and SRV testing to require testing of one-half of the valves at least once per 18 months and all within 40 months is acceptable. Plant specific TS changes to raise the allowable drift tolerance to 13% must include the requirement that additional testing be conducted if failures are experienced; i.e., two additional valves for each valve found with a setpoint outside the 3%

tolerance. In all cases the valve setpoint will be restored to within 11% prior to plant startup. Such testing requirements are consistent with the existing testing requirements.

The GGNS Inservice Testing (IST) program controls the frequency of S/RV testing as required by the GGNS Operating License (TS);

therefore, this proposal will also incorporate changes to applicable IST plant procedures. GGNS willincorporate the recommendations of the NEDC 31753-P-A SER, by resetting the safety function lift setpoints for all tested valves to within 11% of the design lift setpoint and increasing the test sample size by two valves for each valve found outside of the 13% safety function lift setpoint. GGNS will determine the number of S/RVs to be tested each outage using the currently licensed ASME Boiler and Pressure Vessel Code,Section XI,1980 Edition, with Addenda through Winter 1980 (Subsections IWA and IWV), rather than the prescribed 50% of allinstalled valves as recommended by the NEDC 31753-P-A SER and the GGNS Updated Final Safety Analysis Report.

This method of selection agrees with the NEDC 31753-P-A SER Technical Evaluation Report Section 4.4; *However, the adoption of the industry practices proffered by ASME-OM-1981 would provide assurances that an adequate number of operable SRVs exists to prevent the reactor pressure from exceeding design pressure". The currently licensed version of the GGNS ASME Boiler and Pressure Vessel Code will require a sample test population of six S/RVs during the next refueling outage; whereas, ASME-OM-1981 will require the testing of five S/RVs. Therefore, GGNS will be testing an additional valve during the next refueling outage beyond the requirements of ASME-OM-1981.

One of the major factors in the requirement of additional testing population beyond ASME Boiler and Pressure Vessel Code is that many plants were experiencing failures with multiple stage pilot operated S/RVs. The safety function of this type of S/RV requires operation of a pilot valve that is susceptible to excessive leakage and corrosive bonding to cylinder walls; thereby preventing proper safety function operation. The GGNS Dikkers S/RVs are direct acting, and do not require the operation of a pilot valve for the safety function. It f should also be noted that this method of sample selection has been

GNRO 96/00008 Attachment 2 Page 10 of 13 granted to a similar BWR facility requesting the 13% increase in safety function lift setpoint.

NO SIGNIFICANT HAZARDS CONSIDERATIONS Entergy Operations, Inc. is proposing that the Operating License for Grand Gulf Nuclear Station (GGNS) be amended to increase the tolerance of the safety function lift setpoints to 13%. The GGNS Inservice Testing (IST) program controls the frequency of safety relief valve (S/RV) testing as required by the GGNS Operating License; therefore, this proposal will also incorporate changes to applicable IST procedures. GGNS willincorporate the recommendations of the NEDC-31753-P A SER, by resetting the safety function lift setpoints for all tested valves to within 11% of the design lift setpoint anc increasing the test sample size by two valves for each valve found outside of the 13% safety function lift setpoint.

S/RV test sample population will be determined based upon the currently licensed ASME Boiler and Pressure Vessel Code.

The commission has provided standards for detercnining whether a no significant hazards consideration exists as stated in 10CFR50.92(c). A proposed amendment to an operating license involves no significant hazards if the operation 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 possibihty of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant recuction in a margin of safety.

Entergy Operations, Inc. has evaluated the no significant hazards considerations in its request for a license amendment. In accordance with 10CFR50.91(a), Entergy Operations, Inc. is providing the following analysis of the proposed amendment against the three standards in 10CFR50.92(c):

a. No sianificant increase in the probability or consecuences of an accident previously evaluated results from this chance.

The GGNS safety design bases for the S/RVs are:

1) Prevent overpressurization of the nuclear system that could lead to failure of the reactor coolant pressure boundary,

2) Provide automatic depressurization for small breaks in the nuclear system,

3) Permit verification of operability,

4) Withstand adverse combinations of loadings and forces during abnormal, accident, or special event conditions.

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GNRO 96/00008 Attachment 2 Page 11 of 13 ,

l The most limiting vessel overpressurization event is a closure of all main steam isolatiori valves with a high flux scram. This event was analyzed for GGNS using the minimum number of S/RVs required by the GGNS Operating License. The safety function lift setpoint  !

tolerance used in the analysis bounds the proposed 13% setpoint tolerance. The analysis indicates that de S/RVs are capable of maintaining adequate margin below the Operating License Reactor Coolant System Pressure of 1325 psig.

Anticipated operational transients can also challenge the operation of the S/RVs, for instance, Generator Load Reject without Bypass.

Analyses have been performed ori the limiting events that bound other pressure transient events using safety function limit setpoint  ;

tolerances that bound the proposed 13% tolerance request. Fuel ,

operating limits arc based on the results of these analyses; therefore, <

adequate fuel thermal margin is maintained.

Plant transients and events that require the use of automatic depressurization and the low-low set feature utilize the relief mode of S/RV operation.- This proposed change does not affect the relief mode of S/RV operation.

The verification of valve operability will still be performed in accordance with the GGNS Inservice Testing Program, and S/RV safety mode operability will be verified prior to reinstallation. Analysis of the loads placed on each S/RV sub-systa (discharge piping, spargers and associated components) verifies that adequate margin exists to ensure that the overpressurization system can perform its designed function.

The negative tolerance of the safety function lift setpoint remains above the highest setpoint of the S/RV relief mode, and therefore normal vesse! pressure. This margin provides reasonable assurance that inadvertent opening of an S/RV will not occur during power operations.

GGNS will replace each S/RV removed for IST program testing with an S/RV that has been reset to within 11% of the designed safety function lift setpoint During each refueling outage, at least six of the installed S/RVs will be tested for safety lift setpoint in accordance with the current IST program plant procedures. This sample population is in agreement with the current ASME Boiler and Pressure Vessel Code reouirements for the GGNS IST program, and is more restrictive than the ANSI /ASME OM-1 1981 requirement upon which tne setpoint tolerance was based. For S/RV setpoint testing (as-found), additional valves will be tested if the as-found setpoint is outside 13% of its

7NRO 96/00008 Attachment 2 Page 12 of 13 designed safety function lift setpoint. Sample expansion will be consistent with the NEDC 31753-P-A SER requirement of two additional valves per valve failure.

The GGNS UFSAR currently requires at least fifty percent of the installed valves to be removed and tested during each refueling outage. GGNS FSAR Questions & Responses # 211.49 discusses the bases for this requirement. The concern regarded the performance of S/RVs installed in operating plants at the time of GGNS construction and licensing, and that new plants should have significantly better performing S/RVs. The fifty percent requirement provides a very conservative margin of testing to demonstrate that no common cause of S/RV failure occurs within any one operating cycle.

The minimum testing of six valves proposed for each outage, with additional testing for each failure from the initial test population, provides reasonable assurance that no common cause failure is occurring without early detection.

One of the major factors in the requirenent of additional testing population beyond ASME Boiler and Pressure Vessel Code is many of the older plants were experiencing failures with multiple stage pilot operated S/RVs. The safety function of this type of S/RV requires operation of a pilot valve that is susceptible to excessive leakage and corrosive bonding to cylinder walls; thereby preventing proper safety function operation. The GGNS Dikkers S/RVs are direct acting, and do not require the operation of a pilot valve for the safety function. j The Dikkers S/RV Instruction Manual recommends "to replace part of j the installed valves each maintenance stop (refueling outage)", and i does not prescribe any particular amount.

Therefore, no significant increase in the probability or consequences l of an accident previously evaluated results from this proposed i change.

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b. This chance would not create the possibility of a new or different kind of accident from any previously analyzed. l The plant specific analyses verify that each S/RV will still perform the j intended function of preventing overpressurization of the nuclear  ;

system. The vessel will have adequate margin below the Operating l License Reactor Coolant System Pressure of 1325 psig, and p, ant .

system response will not deviate from the e,.pected sequence of  !

events. Each system, structure, and component that communicat :s with the reactor vessel has been verified to be within its design and operational margin, and no unanticipated plant transients will occur as j a result of the safety hit function setpoint tolerance change. q

I GNRO 96/00008 Attachment 2 Page 13 of 13 -

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l The negative tolerance of the safety fu'nction lift setpoint remains 1 above the highest setpoint of the S/RV relief mode, and therefore  !

normal vessel pressure. This margin provides reasonable assurance {

that inadvertent opening of an S/RV will not occur during power J operations.

This proposed change does not add any new systems, structures or supports, nor does it introduce new S/RV operating modes. ,

Therefore, this change would not create the possibility of a new or

- different kind of accident from any previously analyzed.

c. This chanae would not involve a sianificant reduction in the marain of safety.

L The increase in the S/RV safety function lift tolerance has been

analyzed for bounding limiting events and accident conditions. No l condition exists that reduces the margin of safety on the reactor  ;

l coolant pressure boundary or any system, structure or component that l is required to operate during vessel overpressurization events. Fuel l operating limits are based on the results of these analyses; therefore, i j adequate fuel thermal margin is maintained.

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l Therefore, this change would not involve a significant reduction in the margin of safety. ,

i Based on the above evaluation, Entergy Operations, Inc. has concluded that operation in accordance with the proposed amendment involves no significant i hazards considerations. '

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