ML20133A761
| ML20133A761 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 09/25/1985 |
| From: | MISSISSIPPI POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20133A758 | List: |
| References | |
| TAC-59440, TAC-59536, TAC-59644, NUDOCS 8510020304 | |
| Download: ML20133A761 (11) | |
Text
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I ADMINISTRATIVE CONTROLS INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) (Continued)
RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification
- that these activities are performed correctly and that human errors are reduced as much as practical.
AUTHORITY 6.2.3.4 The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities or other means of improving unit safety to the Senior Vice President, Nuclear.
6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide technical support to the Shift Superintendent in the areas of thermal hydraulics, reactor engineering and plant analysis with regard to safe operation of the unit.
6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 4
NRC letter #to all licensees, except for the Chemistry / Radiation Control Super-l intendent who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975; the Shift Technical Advisor who shall meet or exceed the qualifications referred to in Section 2.2.1.b of Enclosure I of the October 30, 1979 NRC letter to all operating nuclear power plants; and those members of the Independent Safety Engineering Group used for meeting the minimum complement m
specified in Section 6.2.3.2, each of whom shall have a Bachelor of Science d-degree or be registered as a Professional Engineer and shall have at least two
- Q years experience in their field, at least one year of which experience shall
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be in the nuclear field.
dQ 6.4 TRAINING g
- L 6.4.1 A retraining and replacement training program for the unit staff shall
- j be maintained under the direction of the Training Superintendent, shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1 2971 d
and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Sections A cnd C of Enclosure 1 of the March 28, 1980 NRC letter #to all licensees, and shall include familiarization with relevant industry operational experience.
6.5 REVIEW AND AUDIT 6.5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC)
FUNCTION 6.5.1.1 The PSRC shall function to advise the GGNS General Manager on all matters related to nuclear safety.
"Not responsible for sign-off function.
Except that the experience and other training information provided in a letter to the NRC dated July 29, 1985 are acceptabic for the individuals listed in GItA00 GU N *
- 1 6-6 Amendment No.
0510020304 050925y 46 PDR ADOCK O
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s 2.
NPE-85/14 and OLCR-NLS-85/04 (PCOL-85/11 additional information)
SUBJECT:
Technical Specification Bases 3/4.8.1, 3/4.8.2 and 3/4.8.3, page B 3/4 8-1 DISCUSSION:
It is proposed to revise the subject technical specification bases to provide a description of the method MP&L proposes to meet the intent of Regulatory Guide 1.9 requirements. The description states that the SSW pump load values are not used as the largest single load rejected from the diesel generators when testing to meet the requirements of Regulatory Guide 1.9.
The change to the bases continues by stating that since the SSW system supplies cooling water to its associated diesel generator, tripping the SSW pump to perform this test would result in loss of the diesel generator.
JUSTIFICATION: MP&L submitted changes to the technical specifications and to Facility Operating License NPF-29 in a letter dated August 23, 1985. These proposed changes resulted from a design change to increase the pumping capacity of Standby Services Water (SSW) loop B.
Receat discussions with NRC staff resulted in the additional information provided in this submittal.
Upon completion of the proposed modifications to the B train of SSW, the new SSW pump will be the largest single load on diesel generator 12 (997 kw versus the 550 kw for RHR B/C pumps). This larger SSW pump load should not be used in Surveillance Requirement 4.8.1.1.2.d.2 to test the diesel generator's capability to reject a large load. Surveillance Requirement 4.8.1.1.2.d.2 demonstrates the ability of the diesel generator to continue to function after rejecting a large load. Since the SSW system supplies cooling water to its associated diesel generator, tripping the SSW pump to perform this test would result in loss of the diesel generator. The present load reject testing requirements using the RHR B/C pump load of 550 kw is the largest single load on the diesel generator that can be rejected and still maintain continued operation of the diesel generator. MP&L believes that this position meets the intent of Regulatory Guide 1.9 requirements and does not propose a change to Technical Specification 4.8.1.1.2.d.2 for this concern. However, MP&L will submit an FSAR revision with the 1986 FSAR update that will address our interpretation of compliance with Regulatory Guide 1.9 for this application.
This change to the technical specification bases was requested by the NRC as a result of discussions concerning the MP&L August 23, 1985 submittal on SSW. This change is made to ensure that users of the technical specifications understand the bases for not using the SSW pumps for load reject testing.
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.SIGNIFICANT HAZARDS CONSIDERATIONS:
This proposed change to the technical specification bases is made to reflect MP&L's position with respect to the largest single load rejected during diesel generator. testing. The intent of the test is to show that the diesel generators can continue to function properly after dropping their largest single load. Since tripping of an SSW pump will cause loss of cooling water to its associated diesel generator, SSW pumps should not be used to perform a largest single load reject test.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the largest single loads that can be rejected and maintain the diesel generators operating are the LPCS pump on diesel generator 11 and RHR B/C pump on diesel generator 12.
This is consistent with present technical specification testing requirements.
The proposed change does not create.the possibility of a new or different kind of accident from any accident previously evaluated because the present single largest load reject testing requirements in the technical specifications reflect the largest loads that can be successfully disconnected without tripping the diesel generators.
The proposed change does not involve a significant reduction in a margin of safety because present technical specification testing' requirements for the diesel generators are not changed and reflect the intent of Regulatory Guide 1.9 requirements.
Therefore, the proposed change involves no significant hazards considerations.
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3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2 r.nd 3/4.8.3 A.C. SOURCES, D.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident con-i ditions within the facility.
The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix "A" to 10 CFR 50.
The volume of fuel specified for each fuel storage system represents usable fuel.
The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commen-surate with the level of degradation.
The OPERABILITY of the power sources are consistent with the initial condition assumptions of the accident analyses and are based upon maintaining at least Division 1 or 2 of the onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source.
Division 3 supplies the high pressure core spray (HPCS) system only.
The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources," December 1974. When diesel generator 11 or 12 is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel genera-tor 11 or 12 as a source of emergency power, are also OPERABLE.
This require-ment is intended to provide assurance that a loss of offsite power event will not result in a cumplete loss of safety function of critical systems during the period diesel generator 11 or 12 is inoperable.
The term verify as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for main-tenance or other reasons.
It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.
The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that (1) the facility can be maintained in the shutdown or refueling condition for extended time periods and (2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status.
The surveillance requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Sup-plies," March 10, 1971, Regulatory Guide 1.108, " Periodic Testing of Diesel j
Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants,"
Revision 1, August 1977; and Regulatory Guide 1.137 " Fuel-Dil Systems for Standby Diesel Generators," January 1978, as addressed in the FSAR, except that Division 1 and 2 diesel generator load testing requirements that were restricted as noted in a letter from E. G. Adensam to L. F. Dale, dated 17 July 1984. A pg eg7-GRAND GULF-UNIT 1 B 3/4 8-1 MeSed No.
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Insert to Technical Specification Bases Page B 3/4 8-1 The SSW pump load values are not used as the largest single load rejected from the diesel generators when testing to meet the requirements of Regulatory Guide 1.9.
Since the SSW system supplies cooling water to its associated diesel generator, tripping the SSW pump to perform this test would result in loss of the diesel generator.
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3.
NPE-85/15 (PCOL-85/11 additional information)
SUBJECT:
Technical Specification 4.5.1, page 3/4 5-5 DISCUSSION:
This technical specification change results from a design change to add high/ low pressure interlocks to the injection valves on the low pressure ECCS systems and is planned for implementation during an outage scheduled to begin October 12, 1985. This change is in. response to an NRC request to provide valve opening times in the technical specifications and is supplemental to the technical specification change request submitted on August 12, 1985, as item 13 of MP&L serial AECM-85/0228.
It is proposed to add a requirement to the subject technical specification that the valves be verified, at least every 18 months, to open in.less than or equal to 29 seconds when tested pursuant to Specification 4.0.5.
JUSTIFICATION: The design change for which this technical specification change is required is described in detail in item 13 of the memorandum dated August 12, 1985, from O. D. Kingsley, Jr., to Mr. Harold R. Denton, MP&L serial AECM-85/0228. As stated therein, the design change is responsive to NRC concerns about postulated
.intersystem LOCA, and will fulfill the requirement of Operating License NPF-29 Condition 2.C.(18) that such a change be implemented prior to startup following the first refueling outage. The supporting technical specification change requested will ensure adequate operability and surveillance requirements exist for the modified system and will make the technical specifications consistent with the as-built plant upon completion of the design change.
The aforementioned memorandum requested deletion of the ECCS system response times for LPCI and LPCS as they will no longer be applicable.
It further stated that the valve opening times would be assured to remain within the limits of the safety analysis by testing in accordance with the requirements of the GGNS Inservice Testing Program. NRC reviewers have since advised MP&L, however, after preliminary review of the change request, that a requirement for verification of the valve opening times must be included in the technical-specifications, and this change request addresses that concern.
The proposed maximum opening time of 29 seconds is conservative with respect to the safety analysis, which assumed valves fully i
open 30 seconds after reactor pressure had decreased to the Reactor Vessel Pressure-Low (Injection Permissive) lower analytical limit during the Design Basis Accident.
Its inclusion in the technical specifications will ensure that the J13 MISC 85091702 - 1
valves are verified to operate within the limits of the analysis every 18 months, which is consistent with the existing interval for verification of the ECCS system response times that will no longer be applicable.
SIGNIFICANT HAZARDS CONSIDERATIONS:
Evaluation of the proposed design change by MP&L in accordance with the provisions of 10 CFR 50.59 has revealed that the change will introduce an additional failure mode for the LPCI and LPCS
- systems, i.e., should an interlock fail in such a manner as to prevent valve opening ECCS injection would not occur from that associated system when required. The required design change has, however, been prepared by MP&L in accordance with applicable regulatory and industry codes and standards. As discussed in Section 6.3.4 of Supplement 2 to the Grand Gulf Safety Evaluation Report, it is the NRC staff's position that redundant protection against overpressurization of the low-pressure ECCSs be provided. The proposed change to add the requirement for verification of valve opening time does not involve a significant increase in the probability or consequences of an accident previously evaluated because the valve opening time is currently surveilled by the requirements of Technical Specification 4.3.3.3 (ECCS system response times).
The ECCS system response times include valve opening times for LPCI and LPCS, and this change simply requires continued surveillance of valve opening times only, since ECCS system response times for' LPCI and LPCS will no longer be applicable once the design change is complete. This change does not create the possibility of a new or different kind of accident from any accident previously evaluated because, as stated, it merely ensures continued surveillance of the valve opening times once the ECCS system response times are no longer applicable.
Neither does this change involve a significant reduction in a margin of safety as no margin of safety is affected by ensuring continued surveillance of ECCS injection valve opening times.
Therefore, the proposed change requiring surveillance of LPCI and LPCS injection valve opening times involves no significant hazards considerations.
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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2)
Low pressure setpoint of the:
(a) LPCI A and B subsystem loop to be 1 38 psig.
(b) LPCI C subsystem loop and LPCS system to be > 22 psig.
(c) HPCS system to be 1 18 psig.
b)
Header delta P instrumentation and verifying the setpoint of the HPCS system and LPCS system and LPCI subsystems to be 1.2 1 0.1 psid change from the normal indicated AP.
3.
Verifying that the suction for the HPCS system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank low water level signal and on a suppression pool high water level signal.
d.
For the ADS at least once per 18 months by:
1.
Performing a system functional test which includes simulated
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autonatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
2.
Manually opening each ADS valve when the reactor steam dome pressure is greater than or equal to 100 psig* and observing that either:
a)
The control valve or bypass valve position responds 1
accordingly, oc b)
There is a corresponding change in the measured steam flow.
g, (INSE RT) l i
AThe provisions of Specification ~4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.
GRAND GULF-UNIT 1
~3/4 5-5 Amendr** d N -
}
INSERT to Technical Specification Page 3/4 5-5 e.
For the LPCI and LPCS injection valves, at least once per 18 months, by verifying that the time required for each valve to travel from fully closed to fully open is s 29 seconds when tested pursuant to Specification 4.0.5.
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4.
OLCR-NLS-85/04 (PCOL-85/11 additional information)
SUBJECT:
Facility Operating License NPF-29, License Condition 2.C.(20), Page 8.
DISCUSSION:
The proposed change to the subject license condition clarifies the surveillance requirement applicable to the valves isolating the spent fuel pool coolers from the SSW system. The proposed change states that the surveillance to be performed is to verify the valves are locked closed and to verify that any SSW loop with valves which are not locked closed is declared inoperable.
JUSTIFICATION:
MP&L submitted a proposed change to the subject license condition in a letter from Mr. O. D. Kingsley, Jr. to Mr.
Harold R. Denton on August 23, 1985 (MP&L serial AECM-85/0258). Recent discussions with NRC staff personnel have resulted in this proposed change to clarify the original submittal. The original change request could have been interpreted to require no surveillance on valves isolating the spent fuel coolers from SSW for both SSW loops if only one SSW loop is declared inoperable. The proposed change will ensure that the surveillance to be performed on the valves isolating the spent fuel coolers from SSW is to verify that the valves are locked closed and to verify that any SSW loop with valves which are not locked closed is declared inoperable. This proposed change ensures compatibility between the requirements of license condition 2.C.(20) and the provisions of Surveillance Requirement 4.7.1.1.a.2 regarding valve surveillance and SSW operability.
Further details regarding the reasons for the original submittal are provided in the August 23, 1985 letter mentioned above.
SIGNIFICANT HAZARDS CONSIDERATION:
This change is proposed to Facility Operating License NPF-29 in' order to clarify a previous submittal. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the. change provides compatibility with Surveillance Requirement 4.7.1.1.a.2 which is in the present licensing basis. The proposed change.does not create the possibility of a new or different kind of accident from any j
accident previously evaluated because this proposed change i
does not alter the present adequate surveillance requirement on the valves isolating service to the spent fuel pool coolers from SSW. The proposed change does not involve a significant reduction in a margin of safety because present surveillance requirements are not reduced.
Therefore, the proposed changes involve no significant hazards considerations.
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l The survedlanc~e 4o be perbmoet is to ver4 +Ae vaives are lockect c)csed and to ve/% ht an3 55W loop itk vo.lves uku a.re not iceked elosed is de clo. red. m opero.ble.
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8 or de assocded SSW subsystern sWI be oleelarol 1j
%jE inoperable.
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(SSW) system and verification that the design flow can be achieved to all SSW system components. However, should a core sT offloading be necessary prior to completion of these modifications
$U,,
(scheduled for the.first scheduled refueling outage), irradiated 2,
jy, fuel may be placed in the spent fuel pool when the RHR system operating in the spent fuel pool cooling mode is available. Until
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the SSW system is modified, the spent-fuel pool cooler shall be isolated from the SSW system by locked closed valvest The post-l!
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tion of these valves shall be verified every 31 days until the design flowrate for SSW system is demonstrated.,
1 (21) Spent Fuel Pool Ventilation System (Section 9.4.2, SER, SSER d2)
If spent irradiated fuel is placed in the spent fuel pool prior to installation and operability of the safety related backup fuel pool cooling pump room coolers, the plant shall be placed in shutdown condition and remain shutdown with the RHR system dedicated to the fuel pool cooling mode.
(22) Remote Shutdown Panel (Section 9.5.4.1 SER, SSER #2)
Prior to startup following the first refueling outage, MP&L shall
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install electrical isolation switches between the control room and the Division I remote shutdown panel.
i (23) Fire Protection Program (Section 9.5.9, SER)
MP&L shall maintain in effect and fully implement all provisions of the approved Fire Protection Plan.
In addition, MP&L shall 1
maintain the fire protection program to meet the intent of
^
Appendix R to 10 CFR Part 50, except that an oil collection system for the reactor coolant pump is not required.
i (24)
Interplant Consnunication Systems (Section 9.6.1.2, SER, SSER #2, 55ER #4, 55ER #5)
Tests of the comunication systems used to mitigate the con-sequences of an event and attain a safe plant shutdown shall be completed during preoperational and startup tests. An evaluation
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of the test results shall be provided for NRC review within 90 days after test completion. Any system modifications found necessary as a result of NRC review shall be completed prior to startup following the first refueling outage.
J (25) Reliability of Diesel-Generators (Sections 8.3.1, 9.6.3 through
)
9.6.7,SER,55ER#2,55ER#4,55ER#6) e
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(a)
Prior to startup following the first refueling outage, a heavy duty turbocharger gear drive assembly shall be installed on all EMD diesel-generators.
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