ML20045F646
| ML20045F646 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/29/1993 |
| From: | Olshan L Office of Nuclear Reactor Regulation |
| To: | Donnelly P CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| GL-89-10, TAC-M84941, NUDOCS 9307080182 | |
| Download: ML20045F646 (14) | |
Text
,
Docket No. 50-155 J e 29, 1993 Mr. Patrick M. Donnelly, Plant Manager Big Rock Point Plant Consumers Power Company 10269 U.S. 31 North Charlevoix, Michigan 49720
Dear Mr. Donnelly:
SUBJECT:
EMERGENCY CONDENSER AND MAIN STEf,M ISOLATION VALVES (TAC NO. M84941)
From April 27 through May 8, and on May 12, 1992, Region III conducted an inspection of your program developed in response to Generic Letter (GL) 89-10, i
" Safety-Related Motor-Operated Valve Testing and Surveillance," at the Big Rock Point Nuclear Plant. The inspection raised concerns involving the emergency condenser inlet isolation valves and the main steam isolation valve.
Enclosed is our assessment of these valves.
We are requesting that you take certain actions that will provide additional assurance that these valves will close as required.
Please provide your response within 45 days of the date of this letter.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten correspondents, therefore, OMB clearance is not required under P.L.96-511.
Sincerely, Original signed by Leonard N. Olshan, Project Manager Project Directorate III-l Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation Request for Additional Information cc w/ enclosure:
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Mr. Patrick M. Donnelly Big Rock Point Nuclear Plant Big Rock Point Plant j
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Mr. Thomas A. McNish, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 j
Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Jane E. Brannon, County Clerk i
County Building Charlevoix, Michigan 49720 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road i
Glen Ellyn, Illinois 60137
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Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N. Logan Street P. O. Box 30195 Lansing, Michigan 48909 l
I U.S. Nuclear Regulatory Commission l
Resident Inspector Office Big Rock Point Plant i
10253 U.S. 31 North Charlevoix, Michigan 49720 l
j Mr. David P. Hoffman, Vice President l
Nuclear Operations l
' Big Rock Point. Plant 27780 Blue Star Memorial Hwy.
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Covert, Michigan 49043 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, N. W.
Washington, DC 20037 i
Enclosure REQUEST FOR ADDITIONAL INFORMATION EMERGENCY CONDENSER INLET ISOLATION VALVES M07052 AND N07062 NRC Insoection 50-155/92010 On April 27 through May 8, and May 12, 1992, Region III conducted an inspection of the program being developed in response to Generic Letter (GL) 89-10, " Safety-Related Motor-Operated Valve Testing and Surveillance," at the Big Rock Point Nuclear Plant.
In its letter on June 5, 1992, forwarding NRC Inspection Report 50-155/92010, Region III stated that the inspection results indicated that the motor-operated valve (MOV) program developed in response to GL 89-10 at Big Rock Point was weak.
The region stated that the licensee should take several actions to address the concerns identified in the inspection report.
For example, the inspectors found that the licensee did not consider the emergency condenser (EC) inlet isolation valves MO-7052 and MO-7062 to be safety-related.
However, the inspectors stated that the MOVs appeared to have a safety function.
The inspectors referred to NUREG-0737 as amended by an NRC letter dated December 15, 1981, that appeared to require the valves to be operational from the control room for the purpose of isolating an Et tube break.
The inspectors stated that these valves provide the only form of isolation between the primary system and the environment in the event of a tube break.
The report indicates that the inspectors performed calculations to evaluate the capability of the MOVs.
The inspectors found that the MOVs might not be capable of isolating a tube break under worst-case conditions.
Assuming a 0.3 valve factor, 0.2 stem friction coefficient, 88% of rated voltage as the minimum voltage at the motor,1545 psid differential pressure, the inspectors determined that the MOVs would be capable of providing 11900 pounds of thrust to close the valves, but predicted that 12800 pounds would be required.
The inspectors also stated that the torque switches for these MOVs were currently set to trip at approximately 9500 pounds of thrust during valve closure.
In the report, the region considered this matter to be unresolved and requested the licensee to provide the basis for excluding MOVs MO-7052 and MO-7062 from its GL 89-10 program.
Licensee Resoonse to NRC Insoection Reoort In its July 20, 1992, response to the NRC Inspection Report 50-155/92-010, the licensee states that it has never considered the operability function of the Et inlet isolation valves as safety-related. The licensee states that operability of the valves is not required to perform the safety function (establish emergency cooling) because the valves are maintained in the normally open position. The licensee states that the design-basis specification in procuring the valves was the capability to isolate a leaking Et tube.
However, the licensee asserts that this function was not considered safety-related for the following reasons:
(1) The Final Hazards Summary Report (FHSR) and the Big Rock Point Technical Specifications 3.4.2(e) and 3.7(b) established that the emergency condenser does not utilize containment isolation valves.
(2) Systematic Evaluation Program (SEP) in Topic Ill.5.A evaluated the
addition of containment isolation valves to the emergency condenser but concluded that the modification was not warranted.
(3) A postulated tube failure in the emergency condenser, even if left un-isolated, results in consequences less than 10 CFR 100 guidelines.
(4) The Big Rock Point design did not provide a safety-related uninterruptible power supply for the EC inlet valves thus concluding that the function of these valves is not required for design bases events.
The licensee notes that the staff did evaluate the need to install automatic emergency condenser isolation on high radiation in Topic II.K.3.14 of NUREG-0737 and concluded that manual trip on high radiation was sufficient. The licensee asserts that the staff's review did not conclude that this function had to be considered safety-related.
The licensee states that in the development of its program in response to GL 89-10, it had included the EC inlet isolation valves in the program as
" position-changeable" H0Vs. The concern postulated by the licensee was that the valves might need to be opened to re-establish the cooling function of the emergency condenser from an assumed inadvertent closure of an inlet valve.
The licensee states that it removed these MOVs from the GL 89-10 program following the issuance of Supplement 4 to GL 89-10 which eliminated the recommendation for boiling water reactor (BWR) plants to address potential valve mispositioning.
In Attachment 2 to its July 20 response, the licensee provides a summary of its estimate of the offsite dose from the release of primary coolant.
The licensee assumed that all of the primary coolant volume of 2689 cubic feet is released in a two-hour time frame at the Technical Specification limit for lodine-131. The licensee states that it assumed the worst-case meteorological conditions for a ground level release. The licensee predicted a thyroid dose rate of 110 rems per hour and a two-hour cumulative dose of 220 rems which the licensee indicates is less than the 10 CFR 100 criterion of 300 rems to the thyroid.
In response to NRC questions, the licensee comparea its method to the staff's methodology in SEP Topic XV-18, dated September 15, 1982, and considers its method to be conservative.
P in Attachment 3 to its July 20 response, the licensee provides its evaluation of the capability of the EC inlet isolation valves to close under worst-case conditions.
The licensee calculates motor actuator output torque and thrust capability as follows:
Output torque capability - MT x OAR x Eff x AF x DVF where the licensee assumes the following parameters MT - nominal motor torque - 15 ft lbs OAR = overall gear ratio - 56.4 Eff - pullout efficiency - 0.4 AF - application factor - 0.90 DVF = degraded voltage factor based on 85% of rated motor voltage Output torque capability - 15 ft lbs x 56.4 x 0.4 x 0.90 x (0.85)^2
- 220 ft lbs 2
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Output thrust capability = Output torque / stem factor where the licensee based stem factor on a 0.2 stem friction coefficient Output thrust capability - 220 ft lbs/0.0198
- 11111 lbs The licensee used the above information to predict the maximum differential pressure under which the MOVs could close:
Required thrust - DP. load + Stem Rejection Load + Packing Load where Required thrust - 11111 lbs DP Load - valve factor x orifice area x maximum dp
- 0.4 x 17.95 in*2 x max dp Stem Rejection load - stem area x max dp - 2.41 in^2 x max dp Packing Load - 1750 lbs 11111 lbs - (0.4 x 17.95 in*2 x max dp) + (2.41 in^2 x max dp) + 1750 lbs solving the equation yields max dp - 976 psid Therefore, the licensee predicts that the MOVs have sufficient capability to close under 976 psid.
In addition, the licensee states that the torque switches are set to trip at approximately 9000 lbs of thrust which predicts a maximum differential pressure of 756 psid.
Reaion III Recuest for Technical Assistance In a memorandum on November 6, 1992, to J. A. Zwolinski, NRR, from H. J.
Miller, Region III, under a Task Interface Agreement (TIA), Region III requests technical assistance in determining:
(1) the design basis requirements in light of the resolution of the SEP topic and (2) the acceptability of the licensee's position that the closing function of the valves is not safety-related and the valves are not required to be in the GL 89-10 program.
Specific Region III questions in Attachment I to the TIA are:
(1) Did SEP topic III-5.A conclude that it was not necessary to require the addition of containment isolation valves because it assumed that the existing isolation valves could perform properly?
(2) In addition, did NUREG-0737, as amended by letter dated December 15, 1981, backfit a safety-related function on these valves?
(3) If not, what is the significance of requiring the MOVs to be operational from the control room for the purpose of isolating a tube rupture?
(4) Is it possible that during a large break loss-of-coolant accident (LOCA) or rapid depressurization, that gap releases could occur, which would increase the dose rate discussed in Attachment 2 of the July 20, 1992, response from the licensee?
EMEB/NRR Evaluation In Section 4.10, " Topic III-5.A, Effects of Pipe Break on Structures, Systems, 3
t and Components Inside Containment," of the Big Rock Point SEP, the staff stated the safety objective for the topic review was to ensure that if a pipe should break inside containment, the plant could safely shut down without a loss of containment integrity and the break should pose no more severe conditions than those analyzed by the design-basis accidents.
In the SEP, the staff states that it had reviewed the licensee's cost-benefit evaluation and concluded that plant modifications to mitigate the consequences of pipe breaks inside containment or to provide protection against cascade failures would not be cost effective.
It is not apparent whether the staff considered the availability of the EC inlet isolation valves in this evaluation.
In Section 4.20.6, " Closed Systems," of the SEP, the staff agreed with the licensee that the cost of adding isolation valves to the emergency condenser vent (which is a different line than the EC inlet) was not warranted, provided the system integrity is periodically verified to qualify the system as an extension of the containment.
In the SEP, the staff agreed that the licensee's roving patrols inside the containment and its leak detection systems were sufficient in lieu of installing vent isolation valves. With respect to the question from Region III on the 5EP, EMEB NRR believes that the staff's statements in f
the SEP focus on the emergency condenser vent and do not resolve Region III's questions.
EMEB/NRR does not recommend that the issue of the capability of the Big Rock Point EC inlet isolation valves be resolved based on the SEP.
I Under Item II.K.3.14, " Isolation of Isolation Condensers on High Radiation,"
in NUREG-0737, the staff stated that the design of the isolation condensers should be modified such that the isolation condensers are automatically isolated upon receipt of a high-radiation signal at the vent rather than at the steam line.
In a letter dated December 15, 1981, the staff enclosed Safety Evaluation Report (SER), " Application of NUREG-0737 Item II.K.3.14 -
' Isolation of Isolation Condensers' to Operating BWRs."
In that SER, the staff stated that Big Rock Point was designed with a radiation monitor and alarm at the isolation vent, but did not have an automatic isolation feature on high radiation at the vent or steam line.
The staff concluded that the manual trip on high radiation levels at the vent is sufficient to provide the amount of flexibility and system availability as a heat sink intended by the NUREG-0737 requirement.
EMEB/NRR considers the staff's SER to assume that the EC isolation valves could be closed by the control room operator. With this assumed capability of the MOVs, the staff concluded in the SER that the licensee need not modify the system to install an automatic isolation feature. The licensee has removed these MOVs from its GL 89-10 program based on Supplement 4 to GL 89-10 on valve mispositioning.
Consequently, EMEB/NRR considers the staff's assumption that the MOVs could be operated from the control room to no longer be a valid assumption, and the licensee not to be meeting the intent of NUREG-0737.
Although the design-basis specification in the procurement of the EC inlet isolation valves was to isolate an EC tube break, EMEB/NRR agrees with the licensee that the Big Rock Point licensing basis did not include these MOVs as safety-related.
However, in the wake of the Three Mile Island accident, the staff required licensees to upgrade various systems to provide additional flexibility in responding to unanticipated events. With respect to the question from Region III, EMEB/NRR does not believe that NUREG-0737 required 4
l licensees to upgrade EC isolation MOVs to full safety-related status (such as safety grade power supplies). However, it is apparent that, in NUREG-0737, the staff determined that these MOVs may be called upon to isolate upon a high radiation signal (such as resulting from an EC tube break) and, therefore, may need to be upgraded to increase their likelihood of performing that function.
Other licensees consider the isolation condenser valves to have a safety function. For example, the licensees of Dresden 2 and 3, Millstone 1, Nine Mile Point 1, and Oyster Creek include the isolation condenser valves in their GL 89-10 programs.
EMEB/NRR performed the following calculations to evaluate the thrust capability of the EC inlet isolation valves:
Output thrust capability = [MT x OAR x Eff x AF x DVF]/SF where EMEB/NRR assumes the following parameters MT - nominal motor torque - 15 ft lbs OAR = overall gear ratio - 56.4 Eff - run efficiency = 0.5 AF = application factor - 1 DVF = degraded voltage factor based on 85% of rated motor voltage SF = stem factor based on 0.2 stem friction coefficient (and is assumed to include load sensitive behavior)
Output thrust capability = [15 ft lbs x 56.4 x 0.5 x 1 x (0.85)^2]/0.0198
- 15435 lbs EMEB/NRR used the above information to predict the maximum differential pressure under which the MOVs could close as follows:
Predicted thrust requirement - DP load x Stem Rejection Load x Packing Load where Predicted thrust requirement - 15435 lbs l
DP Load - valve factor x orifice area x maximum dp l
- 0.4 x 17.95 in^2 x max dp Stem Rejection Load - stem area x max dp = 2.41 in'2 x max dp Packing Load - 1000 lbs as measured by licensee 15435 lbs - (0.4 x 17.95 in'2 x max dp) + (2.41 in'2 x max dp) + 1000 lbs solving the equation yields max dp - 1505 psid The guidelines of the BWR nuclear steam system supplier recommend that MOVs in lines connected to the reactor vessel be capable of closing against the lower safety relief valve setpoint (1545 psid at Big Rock Point) if there is a potential for the main steam isolation valves to close during the postulated line break. Assuming the licensee validates the assumptions in the above calculations, EMEB/NRR considers the capability of the MOVs to close against 1505 psid to provide adequate confidence that the valves could isolate an EC tube break and to satisfy the intent of NUREG-0737.
The licensee provided a calculation that predicts a dose rate of 110 rems per hour and a two-hour cumulative dose of 220 rems at the site boundary if all of the primary coolant with a Technical Specification limit of Iodine-131 is 5
released in a two-hour time frame. Although more recent than the original licensing of Big Rock Point, Section 15.6.4, " Radiological Consequences of Main Steam Line Failure Outside Containment (BWR)," (Rev. 2, July 1984) of the NRC Standard Review Plan.(SRP) states that the distances to the site boundaries and the operation of the safety systems must be sufficient to provide reasonable assurance that the calculated radiological consequences of a postulated main steam line failure outside containment do not exceed ten percent of the exposure guidelines set forth in 10 CFR Part.100 for the case that the failure occurs with a primary coolant activity corresponding to the maximum equilibrium concentration for continued full power operation as stated.
in the standard technical specifications. Similarly, Section 15.6.2,.
" Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment," (Rev. 2, July 1981) of the SRP states that the plant site and safety features are acceptable with respect to radiological consequences of a postulated failure outside containment of a small line carrying reactor coolant if the calculated whole-body and thyroid dose at the exclusion area and the low population zone outer boundaries do not exceed a small fraction of the exposure guidelines values of 10 CFR Part 100 (where a small fraction is defined in the section as 10 percent).
Further, as noted by Region III, gap releases could increase the dose rate predicted by the licensee. Therefore, EMEB/NRR considers the predicted dose rate-to be sufficiently high to suggest that the. licensee take action to ensure that the MOVs are capable of closing in the event of an EC tube break.
Conclusion EMEB/NRR believes that neither the licensing basis nor NUREG-0737 requires the Big Rock Point licensee to upgrade the EC inlet isolation valves to full safety-related status. However, EMEB/NRR considers that staff findings regarding Item II.K.3.14 of NUREG-0737 to have been based on EC inlet isolation valves being capable of isolating an EC tube break.
EMEB/NRR believes that, if tested and maintained properly, the EC inlet isolation valves are sized sufficiently to isolate flow in the event of a line break, although the torque switch might trip prematurely and these MOVs might need to be re-signaled to close.
The staff should request the licensee to perform the following actions to provide assurance that the EC inlet isolation valves are maintained capable of isolating an EC tube break:
(1) raise the torque switches to their maximum allowable setting (without exceeding structural and motor capability limits), (2) maintain and test these MOVs to ensure their continued capability (such as through the program established in response to Generic Letter 89-10), and (3) modify the emergency procedures and position indication, if necessary, to ensure that the MOVs closed fully following a break.
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MAIN STEAM ISOLATION VALVE (MSIV) M07050 NRC Inspection Report 50-155/92010 In NRC Inspection Report 50-155/92010, Region III stated that the main steam isolation valve (MSIV) design basis review submitted in response to Supplement 3 to GL 89-10 was deficient. This analysis determined that the differential pressure across the MSIV would be 235 psid assuming an automatic signal to close approximately 46 seconds after the break. However, MSIV closure woulo actually start 10 seconds after the break based on SEP Topic XV-18,
" Radiological Consequences of a Main Steam Line Failure Outside Containment."
The effect of using a 46 second signal to close in lieu of the 10 second signal would be to decrease the differential pressure across the valve because the reactor coolant system (RCS) would have had more time to blowdown.
The licensee agreed to reanalyze the design basis for the MSIV.
In the inspection report, Region III stated that the licensee had not analyzed smaller breaks in the main steam line because the licentee did not consider these breaks to be part of the licensing basis for the plant.
Small and medium sized breaks may require the MSIV to close against much higher differential pressures due to the limited blowdown of the RCS.
- Further, operator or automatic action may close the MSIV early in the event and may require the valve to close against such high differential pressure.
The NRC inspectors reported that the MSIV design basis document (DBD) referenced three FHSR Chapter 15 accidents (Inadvertent Closure of the MSIV, Increase in Feedwater Flow, and Spectrum of Rod Drop Accidents), but did not address the differential pressure requirements for the MSIV during two of the accidents (feedwater and rod drop).
The licensee considered the thrust requirements for an inadvertent closure of the MSIV to be bounded by the inadvertent turbine trip accident in the DBD.
However, in the turbine trip analysis, the turbine trip valve was assumed to shut before the MSIV closed and the differential pressure across the MSIV was assumed to be zero.
The analysis neglected the automatic opening of the turbine bypass valve which would allow flow and differential pressure across the MSIV.
The NRC inspectors were concerned that the non-safety related turbine bypass valve appeared to be undersized to isolate flow in the event that the HSIV did not close.
Further, the inspectors noted that the turbine bypass valve had a history of sticking in the open position.
The NRC inspectors found several FHSR Chapter 15 accidents that were not referenced in the MSIV DBD where the MSIV may be required to close manually or automatically.
For example, during an Anticipated Transient Without Scram (ATWS) accident, the HSIV may receive an automatic signal to close (low reactor water level).
If the water level continued to decrease, the reactor depressurization system (RDS) would initiate to depressurize the RCS and fuel failure is assumed to occur.
If the HSIV were to stall in a partially open position due to the limited capabilities of the valve, the motor may be damaged beyond recovery because the motor for the MSIV does not have thermal overload protection. Other Chapter 15 accidents that were not included in the MSIV DBD included:
(I) Decrease in Feedwater Temperature, (2) Increase in Steam Flow, and (3) Inadvertent Openir.g of a Safety Relief Valve.
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E The NRC inspectors performed calculations which indicated that the MSIV did not appear to be capable of closing against a differential pressure of 1545 psid (first relief valve setpoint). Using standard industry values for valve and stem friction factors (0.3 and 0.2 respectively), 56% of the nominal voltage available at the motor (based on the licensee's degraded. voltage calculation), and a differential pressure equal to 1545 psid, the inspectors-estimated 46000 pounds of thrust may be required to perform the operation but that the motor actuator would be capable of providing only 32400 pounds of thrust to close the valve.
The NRC inspectors considered the design basis analysis for the MSIV to be deficient because it did not consider credible worst-case differential pressure conditions. The NRC inspectors requested the licensee to evaluate the MSIV and identify the design basis, valve capability under worst-case conditions, corrective actions planned to rectify any deficiencies, and the schedule for completion of those activities. The-licensee was also requested to submit a corrected response to Supplement 3 of GL 89-10 that addresses the appropriate design basis requirements for the MSIV. Region III considered this matter to be unresolved pending further review of the licensee's
- response, k
Licensee Response to NRC Insoection Report In its July 20, 1992, response to the NRC inspection report, the licensee stated that it had performed a detailed review of the FHSR in establishing the functional requirements for the MSIV. The licensee's review of the licensing basis for the MSIV concluded that two Chapter 15 accident' analyses required closure of the MSIV to mitigate the consequences of the event: (1) Containment Isolation following a LOCA, and (2) Steam Line Break Outside Containment.
i For LOCA analysis, the licensee stated that Chapter 15 of the FHSR considered a complete spectrum of pipe breaks up to and including a complete severance of the largest pipe in the system. The licensee stated that the worst or limiting break was determined to be a 0.375 ft^2 split break.
For the LOCA events, the HSIV will receive a close signal on low reactor water level or high containment pressure except during assumed loss of offsite power events when closure occurs approximately 10 seconds after the loss of power. The licensee asserted that, based on reactor pressure blowdown data in Chapter 15.6.4 of the updated FHSR, reactor pressure is less than 200 psia at the end of the MSIV closure stroke.
In the Steam Line Break Outside Containment, the licensee stated that the Chapter 15 analysis assumed a steam line break size of 0.63 ft^2 and MSIV closure to terminate the release. Using the reactor pressure blowdown data for the 0.63 ft*2 break, the reactor pressure at the start of MSIV closure is approximately 1000 psia ar.d at full closure approximately 250 psia for the closure time of 42 seconds. Based on the data, the licensee chose 500 psid as the worst-case differential pressure for the MSIV.
The licensee states that it did not analyze smaller breaks in the main steam line because GL 89-10 did not require an evaluation beyond the existing approved design basis (which has been referred to as the licensing basis).
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i Although the original design-basis specification for the MSIV required the valve to close against 1500 psid, the licensee states that the licensing basis for the MSIV is a double-ended rupture of the Main Steam Line outside containment, as explained in SEP Topic XV-18:
Radiological Consequences of a Main Steam Line Failure Outside Containment (BWR).. Based on this analysis, i
the licensee states that the maximum pressure the MSIV has to close against is less than 500 psia.
In Attachment 4 of its response, the licensee provides a calculation of the capability of the MSIV. The information and assumptions included in the licensee's calculations were (1) stem diameter of 2.75 inches, (2) packing load of stem diameter times 1000 pounds per inch,. (3) 0.4 valve factor based -
on analysis performed by Edwards Valve Incorporated, (4) orifice diameter of 9 and 9/32 inches, (5) 0.2 stem friction coefficient.which is intended to account for degradation over preventive maintenance period and load sensitive behavior, (6) degraded voltage factor of 53.25 percent, (7) piston effect in open direction of 0 pounds, (8) 42 second closure time for MSIV although licensing basis analysis assumes 60 second closure time, and (9) maximum differential pressure analysis of 500 psid which is said to be the' reactor pressure (500 psia) approximately 15 seconds after MSIV receives close signal.
The licensee calculates output torque capability for this de-powered MOV as:
Output torque - MT x OAR x Eff x.AF x DVF where MT - nominal motor torque - 60 ft lbs OAR = overall actuator ratio - 90.7 Eff - pullout efficiency - 0.35 AF = application factor = 0.90 DVF - degraded voltage factor
.5325 f
Output torque - 912.8 ft lbs Output thrust - Output torque / stem factor where the stem factor is based on a 0.2 stem friction coefficient i
Output thrust - 912.8 ft lbs/ 0.0296 lbs - 30837 lbs Required Thrust - DP load + Stem Rejection Load + Packing Load l
where Required thrust = 30837 pounds Packing load = 2750 pounds i
Stem Rejection Load - stem area x max dp - 5.94 in^2 x max dp DP load - valve factor x orifice area x max dp
- 0.4 x 67.65 in^2 x max dp 5
30837 - 2750 + [5.94 x max dp] + [0.4 x 67.65 x max dp).
)
yields max dp - 851 psid
)
The licensee considers the calculation of 851 psid to be acceptable because it bounds the licensing-basis differential pressure of 500 psid.
In Attachment I to its July 20, 1992, response, the licensee provides an additional response to Supplement 3 to GL 89-10. The licensee states that its assumptions are (1) a differential pressure of 500 psid -(rather than the previous 235 psid) based on SEP Topic XV-18: Radiological Consequences of a Main Steam Line Failure Outside Containment, (2) a packing load of 2750 pounds 9
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(rather than the previous 4000 pounds) based on diagnostic measurements, (3) orifice diameter of 9 and 9/32 inches (rather than 12 inches) based on dimensions f;om Wm. Powell Company, (4) valve factor of 0.4 (rather than 0.6) based on analysis by Edwards Valve Incorporated, and (5) degraded voltage value of 53.25% (rather than 80%) based on degraded voltage calculations.
Using these assumptions, the licensee calculates the required thrust to be 19250 pounds and the required torque to be 569.8 ft lbs (based on a 0.2 stem friction coefficient). Assuming nominal motor torque (60 ft lbs), pullout efficiency (0.35), application factor (0.90), the licensee calculates the output thrust and torque capability to be 30837 pounds and 912.8 ft lbs, respectively. Therefore, the licensee considers the MOV to be adequately sized.
Reaion Ill Reauest for Technical Assistance In a memorandum on November 6,1992, to J. A. Zwolinski from H. J. Miller, Region III requests technical assistance in determining:
(1) the design basis requirements for the MSIV, and (2) the acceptability of the licensee's position in that the MSIV is not required to close at full reactor pressure and flow.
Specific Region III questions in Attachment 2 to the TIA are:
(1) Does NRR agree that small breaks outside containment are outside the licensing basis for the plant, even though small breaks may present a greater challenge to the MSIV?
(2) Would the ability of the MSIV to close under the automatic isolation signals as specified in the plant Technical Specifications (TS) be considered to be part of the licensing basis?
(3) If the calculations show that at degraded voltage the TS stroke time would be exceeded, would that be considered unacceptable, such that the licensee would be encouraged to take action to improve the MOV?
EMEB/NRR Evaluation In Chapter 15 of its FHSR, the licensee provides the results of a 0.375 ft*2 pipe break for " Containment Isolation Following a LOCA" and a 0.63 ft^2 pipe break for " Steam Line Break Outside Containment."
In Section 15.6.4 of the FHSR, the licensee states that break sizes up to and including the complete severance of the largest pipe in the system were analyzed to determine that the worst break for peak clad temperature was the 0.375 ft*2 split break.
EMEB/NRR believes that these analyses focused on the high fuel cladding temperature in determining the worst-case break.
The licensee stated in its July 20 response that the original design-basis specification for the MSIV was that the valve be capable of closing against 1500 pounds.
Therefore, it is assumed that the staff reviewers of the FHSR did not consider it necessary to evaluate the pipe break size that resulted in the greatest differential pressure across the valve because the MSIV was purchased to close against high differential pressure.
In Attachment 5 of its July 20, 1992, response, the licensee presents the results of its analysis of a 4 inch, 2 inch, and 1 inch steam line break. The licensee's assumptions included (1) the reactor operator manually scrams the 10
reactor at 20 seconds after the pipe break, and (2) the operator manually closes the MSIV at 60 seconds after the plant was scrammed. However, the Human Factors Branch of NRR has provided guidance that, without interlocks or specific operating procedures, licensees should generally not assume that a control room operator will take no action in determining the worst-case differential pressure in closing valves during an accident because operators might recognize the event promptly and take action when the reactor pressure is greater than the MOV capability.
- Further, the licensee might not have bounded the worst-case differential pressure with its selection of the three break sizes because Figure 1 in Attachment 5 of its July 20 response indicates that the differential pressure is increasing for break sizes beyond the 4 inch break. As noted by Region III in its November 6 memorandum, the MSIV receives an automatic signal to close upon low reactor water level or high containment pressure when the differential pressure across the valve could be higher than assumed by the licensee. Therefore, EMEB/NRR considers that the MSIV might be required to close against a differential pressure significantly greater than 500 psid assumed by the licensee.
EMEB/NRR evaluated the capability of the MSIV as follows:
Output thrust capability = [MT x OAR x Eff x AF x DVF]/SF where EMEB/NRR assumes the following parameters MT - nominal motor torque - 60 ft lbs OAR - overall gear ratio - 90.7 Eff = run efficiency = 0.4 AF = application factor = 1 DVF = degraded voltage factor based on 53.25% of rated motor voltage SF - stem factor based on 0.2 stem friction coefficient (and is assumed to include load sensitive behavior)
Output thrust capability = [60 ft lbs x 90.7 x 0.4 x 1 x 0.5325]/0.0296
- 39160 lbs EMEB/NRR used the above information to predict the maximum differential pressure under which the MOV could close as follows:
Predicted thrust requirement - DP Load x Stem Rejection Load x Packing Load where Predicted thrust requirement - 39160 lbs DP Load - valve factor x orifice area x maximum dp
- 0.4 x 67.65 in^2 x max dp Stem Rejection Load - stem area x max dp - 5.94 in^2 x max dp Packing load - 2750 lbs as measured by licensee 39160 lbs - (0.4 x 67.65 in^2 x max dp) + (5.94 in*2 x max dp) + 2750 lbs solving the equation yields max dp - 1103 psid EMEB/NRR predicts that the MSIV would be capable of closing against a differential pressure of 1103 psid which is less than its original design-basis specification of 1500 psid. The GE Owners Group has recommended that BWR licensees assume the first relief valve setpoint (1545 psid for Big Rock Point) for typical lines connected directly to the RCS.
Further, as indicated by Region III, the HSIV might receive signals to close either automatically or 11
manually during transients with high differential pressure across the valve.
Therefore, EMEB/NRR determines that the licensee has not demonstrated that the MSIV will be capable of closing against its credible worst-case differential pressure condition.
Region III asks whether the ability of the MSIV to close under the automatic isolation signals as specified in the Big Rock Point TS is considered to be part of the licensing basis of the valve.
In this evaluation, EMEB/NRR has considered the licensing basis for the MSIV to-focus on its FHSR requirements.
The Big Rock Point TS should not require the MSIV to close either automatically or manually under conditions which the MSIV is not capable of operating.
With respect to the Region III question on stroke time, EMEB/NRR does not consider the issue of stroke time to be significant unless the stroke time exceeds the time assumed for the valve to close in the safety analyses.
The licensee should ensure that the MSIV can perform its function within an acceptable time period of its safety analyses.
The Big Rock Point licensee argues that its decision not to take action to upgrade the capability of these MOVs is supported by consequent offsite releases being slightly lower than the Part 100 limits.
As discussed with respect to the EC inlet isolation valves, EMEB/NRR considers the predicted dose rate to be sufficiently high to suggest that the licensee teke action to improve the capability of the MSIV.
Conclusion EMEB/NRR concludes that the accident scenarios analyzed in the Big Rock Point FHSR require as part of the plant's licensing basis that the MSIV be able to close against low differential pressure. The FHSR does not include small steam-line break scenarios which might result in the need for the MSIV to close against high differential pressure.
EMEB/NRR considers the MSIV to not be capable of closing against its original design-basis differential pressure specification (1500 psid) that might credibly occur during a small steam-line break or plant transient when the valve might receive an automatic or manual signal to close.
The staff would need to justify any modification to upgrade the MSIV as a backfit because the licensing basis for the MSIV does not include high differential pressure conditions.
In lieu of processing a backfit analysis, EMEB/NRR recommends that the licensee be requested (1) to verify the calculational assumptions, (2) to set the torque switch to the maximum allowable setting (without exceeding structural or motor capability limits), (3) to establish operator procedures to prevent premature attempts to close the valve prior to differential pressure being within the capability of the MSIV, and (4) to ensure that the plant TS do not require the MSIV to close either automatically or manually under conditions which the MSIV is not capable of operating.
EMEB/NRR considers that these minimal-cost actions would provide an acceptable level of confidence that the MSIV could close against high differential pressure and reduce the risk associated with a small steam-line break or other plant transients.
If the licensee took these actions, EMEB/NRR believes that a modification to upgrade the MSIV would not be necessary if evaluated as part of a backfit cost-benefit analysis.
12 Date:
June 29, 1993
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