ML20078E694
| ML20078E694 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 09/23/1983 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20078E683 | List: |
| References | |
| NUDOCS 8310060050 | |
| Download: ML20078E694 (11) | |
Text
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[,g NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 E
SAFETY EVALUATION AMENDMENT NO. 10 TO NPF-13 GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 Introduction The licensee proposed changes to the Technical Specifications and changes to the operating license for Grand Gulf Unit I which are as follows:
(a) Changes to the following Technical Specifications (MP&L letters dated June 23, 1983, and August 1, 1983):
1.
Tables 3.3.3-1 and 4.3.3.1-1, Bases Figure 3/4 3-1: Redefines OPERABILITY range for High Pressure Core Spray (HPCS) until first refueling outage due to water level instrumentation inaccuracies at low pressure (August 1, 1983).
2.
Table 3.6.4.1: Design change to prevent automatic tripping of RHR jockey pumps, needed to prevent potential damage from waterhammer (June 23,1983).
(b) One time exceptions to the folicwing Technical Specifications (MP&L letters dated June 14, 1983, and August 1, 1983):
3.
4.4.2.1.2.b. 4.4.2.2.1.b and Table 3.3.3-1: Provisions of Specification 4.0.4 suspended to allow plant to attain operating conditions necessary for ADS Trip System surveillance testing (June 14, 1983).
4.
4.1.3.1.4.a:
Provisions of Specification 4.0.4 suspended to allow plant to attain operating cor.ditions necessary for Scram Discharge Volume surveillance testing (August 1, 1983).
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. Evalcation (a) Technical Specification Changes 1.
Tables 3.3.3-1 and 4.3.3.1-1, Bases Figure 3/4 3-1:
HPCS OPERABILTY In a letter dated August 1,1983, MP&L requested changes to the Technical Specifications to modify the Specificrtions on the high pressure core spray (HPCS) system so that the injection function of Drywell Pressure - High and Manual Initiation are not required to be OPERABLE when the indicated water level on the wide range instrument is greater than Level 8 coincident with the reactor pressure being less than 1025 psig.
Informal discussions between the staff and MP&L on August 25, 1983, in regard to the pressure value revised this value to 600 psig. Written confirmation of the 600 psig value was submitted by MP&L in a letter dated September 13, 1983. These changes would be in effect until modifications are made to the instrumentation but no later than startup following the first refueling outage.
The reactor water level instrumentation at Grand Gulf is the condensate chamber reference leg type. These instruments are strictly differential pressure devices which are reactor coolant density sensitive and are calibrated to be most accurate at the specific vessel conditions appropriate for the associated system functions actuated by the instrumentation. The shutdown water level range and fuel zone water level range instruments are calibrated to read accurately at atmospheric pressure; the upset, narrow and wide range water level instruments are calibrated for normal operating conditions (saturated steam at 1025 psig). At low coolant temperatures and pressures, those instruments calibrated for normal operating conditions will read higher than actual level.
For example, with an actual level of 21.5" above instrument zero at 120*F and atmospheric pressure, MP&L has observed that the narrow range instrumentation would indicate a level of 32" and the wide range instruments would attempt to indicate a level of 82.5" (in actuality the upper limit of the range of the instrument is 60").
The HPCS discharge valve is interlocked closed at the vessel Level 8 setpoint (less than or equal to 55.7" above instrument zero or 220" above the active fuel). An artificially high level indication at low pressure may result in HPCS isolation when the actual vessel level is below the Level 8 setpoint.
The isolation logic may be manually reset once the indicated I
vessel level drops below this setpoint or it will reset automatically when the indicated vessel level reaches the Level 2 HPCS initiation setpoint (a level of - 41.6" below instrument zero or 125" above the fuel).
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. Technical Specification Table 3.3.3-1. requires HPCS Drywell Pressure-High and Manual Initiation Actuation Instrumentation to be OPERABLE in various Operational Conditions. Actuation of these devices will result in vessel injection unless reactor vessel level is above the Level 8 setpoint or the Level 8 isolation has not been reset. The changes proposed by MP&L would add a note to Technical Specification Tables 3.3.3-1 and 4.3.3.1-1 to indicate that the injection function of Drywell Pressure-High and Manual Initiation are not required to be OPERABLE at times when a false Level 8 isolation signal is present.
Only one accident analysis presented in the Grand Gulf FSAR assumed HPCS initiation via a high drywell pressure signal.
This event was a steamline break inside the drywell. MP&L presented the results of a re-analysis of this event assuming that the high drywell initiation feature was defeated and that HPCS initiated on low reactor vessel level (Level 2) only.
The model accounts for density changes in the coolant by utilizing the mass of the coolant as the parameter which actually initiates HPCS injection. Thus, the mass of coolant in the vessel is the same at the time of system initiation regardless of reactor pressure. The worst single failure for the revised case was determined to be the Division I diesel generator failure. The calculated peak cladding temperature (PCT) is 1322*F versus 900 F as previously calculated without the inaccuracy in the instrumentation.
MP&L also indicated that at pressures greater than or equal to 600 psig, the indicated level error is small enough that manual reset of the HPCS isolation logic is possible for actual vessel water levels up to the high end of the normal operating range. The proposed Technical Specification, therefore, requires that the injection function be OPERABLE whenever indicated level is less than Level 8 or reactor pressure is greater than or equal to 600 psig.
Comments on these changes were received from the State of Mississippi in a letter dated August 11, 1983, and from Mr. Kenneth Lawrence by telephone on September 6,1983.
The State of Mississippi was concerned with the effect of the water level instrumentation inaccuracies on other safety analysis and with the 422 F increase in peak cladding temperature (PCT) for the reanalyzed event involving steamline break inside containment including the effect of the increased temperature upon fuel failure rate. On August 30, 1983, we discussed these concerns with Mr. Eddy Fuente, Director, Division of Radiological Health. We discussed in detail our view regarding why a predicted increase in the worst-case peak cladding temperature, an increase of 422 F to 1322*F, was not considered to be a significant safety concern. We indicated that based on data provided from tests performed at y
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. the Power Burst Facility (PBF) no significant change ir. the number of fuel rod failures would be expected. We also discussed that on a best-estimate basis, little or no change in the peak cladding temperature would be expected for this transient with or without HPCS initiation.
With regard to the performance of the water level instrumentation for the HPCS, we discussed with Mr. Fuente that a number of other systems would be more responsive to mitigation of accidents at low pressure and low temperature and would be available in the event of a LOCA during these infrequent rodes of operation; that is during start up and shut down. The models used in the safety analysis adequately model more realistic plant situations. For licensing evaluations, analysis models compensate for density variations in the calculation of actual vessel level and PCT under LOCA conditions. The original safety analyses which initiated HPCS on high drywell pressure contained an erroneous input assumption, i.e., HPCS was always available for initiation from a high drywell pressure signal. This is not always true because HPCS lock out is possible from an indicated level signal. The indicated level signal is not density compensated at low pressure, and a false high water level signal can lock out the high drywell and manual initiation signals. The model for safety analysis itself does not use indicated levels in the calculation of PCT, but rather calculates it as a function of density. The revised analysis corrects the erroneous assumption of HPCS initiation on high drywell pressure. A mass setpoint which is independent of pressure is used to initiate HPCS. The setpoint is determined as that mass which would have a level equal to Level 2 ff reactor pressure was 1000 psi. The actual level, when the liquid mass is equal to the setpoint mass, is a function of pressure. However, the actual level does not affect HPCS initiation either in the real plant or in the safety analysis model. The actual level is calculated in the model for purposes of heat transfer calculations and PCT determination. Mr. Fuente generally agreed with our conclusions and was satisfied that we had appropriately considered their comments provided in the August 11, 1983, letter, Mr. Lawrence was also concerned about the 422 F increase in peak cladding temperature.
In our discussion on September 6, 1983, we pointed out to him that the 422'F calculated increase in peak cladding temperature would result in a PCT of 1322 F which was significantly below the long established regulatory limit of 2200*F. As noted in the discussion on the subject above, we further conclude that if such a 422 F increase were to occur, it would not result in any significant change in the number of fuel failures.
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. Only one accident analysis was involved that assumed HPCS initiation during a time affected by a false water level reading. As to the one affected reanalysis, the delay in HPCS initiation resulted in a PCT of 1322*F. The applicable acceptance criterion for PCT, as given in 10 CFR 50.46, reads 4
as follows:
"The calculated maximum fuel element cladding temperature shall not exceed 2200*F." Since 1322*F is well below the established safety limit, these changes in the Technical Specifications have been determined not to result in exceeding regulatcry limits.
The Commission has provided guidance for the application of the standards for making a "no significant hazards con-sideration" determination by providing examples of amendments that are considered not likely to involve significant hazards consideration (48 FR 14870). One of the examples of an amendment which will likely be found to involve no significant hazards consideration is listed as follows:
(vi) A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or componcnt specified in the Standard Review Plan.
While the changes do result in a reduction in the safety margin in that the predicted fuel cladding temperature is increased, the peak temperature is well within the limits of 10 CFR 50.46. For this accident, a margin of over 800*F exists between the revised accident PCT of 1322*F and the 2200*F PCT limit permitted by the acceptance criteria for Emergency Core l
Cooling Systems for Light Water Reactors. Thus, these changes are similar to those considered in the Commission's example (vi) of changes not likely to involve significant hazards considerations.
We have reviewed the Technical Specification changes proposed by MP&L and the results of the supporting analyses. We find the changes acceptable on the basis of technical and safety considerations. We further conclude that the proposed action does not:
(1)
Involve a significant increase in the probability or consequences of an accident previously evaluated; or e
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. (2) Create the possibility of an accident of a type different from any evaluated previously; or l
(3)
Involve a significant reduction in a margin of safety.
The proposed Technical Specification changes are acceptable t
and since the action conforms to example (vi) given above, i
do not involve a significant hazards consideration.
(2) Table 3.6.4.1: Jockey Pump Isolation By letter dated June 23, 1983, MP&L proposed the implementation of a design change package (DCP) and an accompanying change to Technical Specification Table 3.6.4.1 to correct a design deficiency reported by MP&L letter dateJ April 15, 1983. The design deficiency concerns the inability due to loss of the RHR jockey pumps to maintain a continuous pressurized water supply to keep the main RHR pump discharge piping full. The proposed design change will prevent the automatic tripping of the jockey pumps, partial draining of the RHR pump discharge piping and, subsequent waterhammer in the RHR system due to minimum flow valve closure.
The NRC staff has reviewed the design change which revises the logic for both the jockey pumps and their associated minimum flow valves and concludes that it is both appropriate and necessary to minimize potential waterhammer in the RHR system.
The design change will assure that a pressurized source of water is provided to the RHR pump and discharge piping continuously by removing the containment isolation signal from the minimum flow valves. Effective containment isolation for the minimum flow lines will be maintained with the jockey pump running because the pump discharge pressure is higher than containment pressure, assuring in-leakage, not out-leakage. On the other hand, when the pump is not running, the revised logic will sense this and, only in this mode, close the minimum flow valve.
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. Waterhammer events in nuclear power reactors have been reported and several incidents have resulted in piping and valve damage. Thus, since the changes clearly correct a deficiency in the RHR system, this action is necessary to prevent the possibility of damage to the RHR system by waterhammer, and we, therefore, conclude that the proposed action does not involve a significant increase in the probability or consequences of an accident previously evaluated or involve a significant reduction in a margin or safety. Furthermore, since the revised logic has no real effect on containment isolation, this action does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Thus, we find that the Technical Spec ~ification changes are acceptable on the basis of technical and safety considerations and that the proposed action does not involve a significant hazards consideration.
(b) One Time Technical Specification Exceptions (3) 4.4.2.1.2.b, 4.4.2.2.1.b and Table 3.3.3-1: ADS Trip System Testing In a letter dated June 14, 1983, MP&L requested that the provisions of Specification 4.0.4 be suspended to allow surveillance testing of safety relief valves in the ADS System after reactor steam pressure is adequate to perform the test. Specification 4.0.4 ensures that surveillance activities associated with a Limiting Condition for Operation have been performed within the specified time interval prior to entry into an applicable OPERATIONAL CONDITION or other specified applicability condition. The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation. Suspension of the provisions of this specification will allow the plant to restart prior to performing the specified surveillance
. testing for the ADS Trip System. For performing this surveillance test, a reactor pressure of at least 100 psig is required.
In this exception, the surveillance test must be completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after attaining a pressure sufficient to perform the test. Due to the low reactor pressure and temperature required to pert'orm this test during the initial heatup with essentially unirradiated fuel, we conclude tha.t the proposed action is acceptable on the basis of technical and safety considerations and will not:
(1)
Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create 'the possibility of a new or different kind of accident from any accident previously evaluated; or (3)
Involve a significant reduction in a margin or safety.
Thus, we find that this action is acceptable and does not involve a significant hazards consideration.
(4) 4.1.3.1.4.1:
Scram Discharge Volume Testing In a letter dated August 1,1983, MP&L requested that the provisions of Specification 4.0.4 be suspended to allow surveillance testing of the Scram Discharge Volume after achieving the control rod configuration required to perform the test. Specification 4.0.4 ensures that surveillance activities associated with a Limiting Condition for Operation have been performed within the specified time interval prior to entry into an applicable OPERATIONAL CONDITION or other specified applicability condition. The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation. Suspension of the provisions of this specification will allow the plant to restart prior to performing the specified surveillance tisting for the Scram Discharge Volume.
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. For performing this surveillance test, a normal control rod configuration of less than or equal to 50% R0D DENSITY is required.
In this exception, the surveillance test must he completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after attaining a R00 DENSITY sufficient to perform the test. Due to the low power (5-10%)
required to achieve the necessary R0D DENSITY and the unirradiated condition of the initial core, we conclude that the proposed action is acceptable on the basis of technical and safety considerations and will not:
(1)
Involve a significant increase in th'e probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)
Involve a significant reduction in a margin of safety.
Thus, we find that this action is acceptable.and does not involve a significant hazards consideration.
Environmental Consideration We have determined that this amendment does not authorize a change in effluent types or total amount nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that this amendment involves action which is in-significant from the standpoint of environmental impact, and pursuant to CFR Section 51.5(d)(4), that an environmental impact statement or negative declaration ana environmental impact appraisal need not be prepared in connection with the issuance of this statement.
i Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered, does not create the possibility of a new or different kind of accident from any previously evaluated, and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety
-of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the. issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
- See previous concurrences DL:LB#2/PM
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Involve a significant reduction in a margin of safety.
Thus, we find that this action is acceptable and does not involve a significant hazards consideration.
Environmental Consideration i
We have determined that this amendment does not authorize a change in effluent types or total amount nor an increase in pover level and will not result in any significant environmental impact. Having made this determination, we have further concluded that this anendment involves action which is in-significant from the standpoint of environmental impact, and pursuant to CFR Section 51.5(d)(4), that an environmental impact statement or negative declartion and environmental impact appraisal need not be prepared in connection with the issuance of this statement.
Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered, does not create the possibility of a new or different kind of accident from any i
previously evaluated, and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the
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common defense and security or to the health and safety of the public.
Dated:
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s 4 For performing this surveillance test, a normal control rod configuration of less than or equal to 50% R0D DENSITY is required.
In this exception, the surveillance test must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after attaining a R0D DENSITY sufficient to perform the test. Due to the low power (5-10%)
required to achieve the necessary R0D DENSITY and the unirradiated condition of the initial core, we conclude that l
the proposed action will not:
(1)
Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)
Involve a significant reduction in a margin of
- safety, i
Thus, we find that this action does not involve a significant hazards consideration.
Environmental Consideration The Commission has determined that the issuance of this amendment will not result in any environmental impacts other than those evaluated in the Final Environmental Statement since the activity authorized by the amendment is encompassed by the overall action evaluated in the Final Environmental Statement dated September 1981.
Conclusion 4
We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: September
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