ML20155H977

Summary of 880831 Meeting W/Util in Rockville,Md Re Status of Leasing Activities & Schedule for Completion of Activities.Related Info,Including Meeting Agenda & List of Attendees Encl

ML20155H977
Person / Time
Site:	Grand Gulf
Issue date:	10/18/1988
From:	Kintner L Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
GL-83-28, GL-87-09, GL-87-9, GL-88-12, IEIN-86-081, IEIN-86-81, IEIN-87-043, IEIN-87-43, NUDOCS 8810250029
Download: ML20155H977 (36)

Text

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October 18, 1988 Docket flo. 50-416 DISTRIBUTION See next page LICEllSEE: System Energy Resources, Inc. (SERI)

FACILITY: Grand Gulf Ntclear Station, Unit 1 (GGNS-1)

SUBJECT:

SulWARY OF AUGUST 31, 1988 fiEETING REGARDING LICENSING ACTIVITIES A mecting was held with System tnergy Resources, Inc., on August 31, 1988 in Rockville, Maryland. The purpose of the nieeting was to discuss the status of licensing activities and the schedule for completing them. Enclosure 1 is a list of attendees at the meeting. Enclosure 2 is the meeting agenda prepared t'y the NRC staff. Enclosure 3 is the licensee's proposed markup of changes to Technical Specifications (TS) in Section 6.0. Enclosure 4 is a handout prepared by the licensee for use in describing the design and operation of the proposed new alternate decay heat removal system to be installed in the third refueling outage (RF03) scheduled to begin March 1,1989.

The staff provided the status of the licensee's submittals requesting TS changes. To achieve greater operational flexibility, the licensee has proposed .

to separate the diesel generator 24-hour surveillance test from the test of a simulated loss-of-coolant accident (LOCA) and shnulated loss of offsite power (LOP). This is a generic TS change that is being reviewed by the Electrical Systems Branch. The present TS require running the LOCA and LOP test itraediately af ter the 24-hour test when diesel generator thermal conditions have stabilized. A footnote to the TS allows a separate LOCA and LOP test if there is a failure of this test after a successful 24-hour test. The footnote requires a pt i or run of the diesel generator for one hour or until thernal conditions have stabilized. The proposed separate LOCA and LOP test would have the same preraquisite as the present footnote. The staff indicated that this is priority 4 review and the target date for issuance of the atendment is January 15, 1989.

The proposed change to the snubber sonpling plans in the TS is a priority 4 review. The proposal is similar to those reviewed and accepted on other plants. The amendment is scheduled for issuance on January 30, 1989.

The proposal to lower the doc travel power cutoff setpoint for the fuel hoist is a priority 2 review and is scheduled for coripletion on December 15, 1966.

The basis for the cutoff point is to allow the fuel grapple to travel low enough to engage fuel assembly and control blade guide handles and to prevent the grapple from traveling below the reactor vessel top grid where it may be damaged. The present TS uses fuel assembly handles as the reference for making the setpoint, but fuel assembly len th increases with irradiation, and the iicensee experienced difficulty dur ng RF02 engaging control blade guides which

, do not grow with irradiation. The new TS would use the reactor vessel top cuide as a reference for r.aking the setpoint.

8s10:50029 ss1018 DR ADOCK 0500 6

) The proposed changes to TS 3.0 and 4.0 (Generic Letter 87-09) were discussed.

Based on its review to date, the staff noted that the proposal identifies each i TS that would be changed by changing TS 3.0.4 to allow operational condition

changes with inoperable equipment, provided the applicable action statements do l not require a plant shutdown. However, safety analyses and determinations that i there were no significant hazards considerations (NSHC) were not provided for

each TS chtnge. The staff stated that a safety analysis and NSHC determination should be made for each TS that was changed. Those TS which presently have an i i exception to TS 3.0.4 would not need to be analysed individually, provided l i deletion of the exception does not change the TS. In addition to safety t

analyses of plant specific changes to TS, assurance should be provided that the l maintenance priorities will be such that plant startup will normally be '

i initiated only when all equipment required to meet LCOs is operable. The staff l scheduled October 21, 1988 as the target date for submitting additional  !

information and scheduled an amendment issuance date of December 16, 1988.

The schedule dates for completion of other licensing actions is as follows:

October 30, 1988 - Extension of schedule for installation of RG 1.97 '

neutron flux monitors f om RF03 to RF04

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January 15, 1989 - Increase reactor protection system surveillance  !

intervals, l December 31, 1988 - Change fire protection license condition. This date  !

is based on receipt of a response to Generic Letter 88-12 on October 30, l 1988.  !

September 26, 1988 - Change TS to implement 10 CFR Part 55 regarding  !

training of licensed personnel. This date was subsequently changed to l October 28, 1988 per telephone call with H. Crawford (SERI) on September j 30, 1988.

October 6, 1988 - Change fire protection zore because of room modifica-tions. This amendnent was issued September 23, 1988.  :

i September 30, 1988 - Change position of PSRC member because of l organization change. This amendment was issued September 21, 1988.  :

6 October 13, 1988 - Deletion of daily functional test in Rod Pattern j Control System  ;

October 21, 1988 - Add action statements for inoperable pressure and  !

leakage instruments for the control rod scram accumulators.  ;

The licensee stated a TS change would be submitted in October to replace  !

position titles in Section 6.0 to preclude unnecessary TS changes when j organization changes Gre made. A draft markup was provided and discussed ,

(Enclosure 3). The licensee questioned whether the specific title of top i management should remain. In an October 6, 1988 telephone call, the staff said  !

that the Vice President, Nuclear Operations and the GGNS General Manager should I remain as specific titles. Further, in response to the licensee's query, the i I

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staff said on October 6, 1988 that the ANSI /ANS Standard 3.1-1981 would not need to be referenced because the SRC member qualifications were stated in the TS, except for the qualifications of the SRC Chairman. Qualifications for the Chairman as stated in ANS 3.1-1981 Section 4.7.lb, should be included in that paragraph (See marked up TS pages 6-9, Enclosure 3).

The iicensee stated that it will send a letter to withdraw its request of July 6,1987 regarding isolation valve operability during shutdown.  ;

Regarding followup information, the following coments were made:

The documentation for completed license conditions will be provided by the licensee prior to RF03. i The procedures for use of the MSA-GMR-1 canisters to enter areas with a high airborne radioactivity were issued January 18, 1988 ,

(015-08-04 Revision 12). Training lesson plans were revised January i 19, 1988. Training will be a part of the regular training program; but if needed for use before training is completed, personnel will be -

trained prior to use.

• The licensee issued Material Noncomformance Report (MNCR) No. 0163-88 as followup to the NRC Information Notice (IN) 87-43 regarding gaps in  ;

the poison material found in Joseph Oats high dentity fuel racks at  ;

other facilities. Blackness tests were completed in the sumer of 1988 and no gaps were found. Samples of the racks, both irradiated and l untrradiated, indicated that some tearing or gaps nay be present.  :

Examinations and corrective actions, if needed, will be completed i before new fuel arrives on site in January 1989. '

l The licensee visually inspected springs in the main steam isolation -

valves (MS!V) in RF01 as a followup to IN 86-81, which described i broken springs found in Atwood Morrill MSIVs at Fermi. Inspections L recomended by General Electric Company in SIL 422 were performed.

The licensee talked to personnel at Detroit Edison Company who said the failures of the springs were plant specific. The NpRDS file showed i that of 2704 total valves in 26 plants, the only failures of this type l were the four valves at fermi. SERI engineering department recomended j tests of all replacements springs at 105% of the design load. Followup 1 of IN 86-81, Supplement 1, will be done during a Project Manager site visit.

The licensee said that the responses to Generic Letter 83-28 regarding the Salem ATWS event are applicable to both Units 1 and 2. For those an responses which were submitted only on Unit 1senttomakeresponsesapplicableto

• The licensee will submit its analysis of the heat removal capability of the spent fuel pool cooling and cleanup system prior to RF03. The present TS limit the nurber of ftel assemblies that can be put into the high density spent fuel racks until adequate cooling capacity is demonstrated. This limiting number of fuel assemblies in the present T3 will not be reached until RF05.

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4-L The Project Manager advised the licensee of the San Onofre Unit I water hammer event of November 21, 1985, which was caused by failure of check [

valves. The valves failed because of repeated harmering of the disc l stud and stud nut against the backstop during prolonged operation with low flow in the pipe. The licensee said it would study its systems for ,

susceptibility to this type of failure. The Project Manager will  !

obtain results of the study on a site visit. j I

The licensee stated that new TS submittals would be made as follows- t

• September 1, 1988 - Deletion of leak rate tests for containment isolation valves in small diameter (1. inch) test, vent and drain lines.  !

September 1, 1988 . Isolation of fuei transfer canal during refueling i to work on transfer cechanism, if neces u ry.

7 September 9, 1988 . T5 changes related to the alternate decay heat removal system (ADURS). The staff said that the design and safety analysis of the ADHRS should be submitted for the staff review and not  ;

just the changes to TS (new radiation monitors for plant service water t and additional thermal overload protection devices for isolation  !

valves).  !

• December 9, 1988 ~ Reload for RF03 l f

The licensee's engineering personnel described the ADHRS cesign and operation f (Enclosure 4). The system is designed to remove decay heat in the reactor >

starting 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown. It would be used only during refueling  ;

outages, and then only a short time when both RHR trains are out of service. i Design pressure is 250 psig, which is less than the design pressure of the  !

LPCI.C piping to which it would discharge. The two pumps and two heat exchangers would be located in the RHR C Pump Room. Plant service water would be used in the heat exchangers.

W\ l Lester L. Kintner, Senior Project Manager i Project Directorate !!.1 f Divison of Reactor Projects I/II  :

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Enclosure:

l As stated i I

cc w/ enclosure, i See next page l h  !

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DATE :10//i/88 :10/]/88 :10//A/88  :  : .*  :

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. . . . - - - , . - - ___ _- - - .,._n - - _ --. _ ,_ -

l Mr. T. H. Cloninger System Energy Resources, Inc. GrandGulfNuclearStation(GGNS)

CC:

Robert B. McGehee, Esquire Mr. C. R. Hutchinson Wise, Carter, Child, Steen and Caraway GGNS General Manager ,

P. O. Box 651 System Energy Resources, Inc.  !

Jackson, Mississippt 39205 Post Office Box 756 '

Port Gibson, Mississippi 39150 i Nicholas S. Reynolds Esquire The Honorable William J. Guste, Jr. l Bishop, Liberman, Cook, Purcell Attorney General ,

and Reynolds Department of Justice 1 1400 L Street, N.W. State of Louisiana 1 Washington, D.C. 20005-3502 Baton Rouge, Lou ntana 70804 Mr. Ralph T. Lally Office of the Governor ,

j Manager of Quality Assurance State of Mississippi  !

Middle South Utilities System Jackson, Mississippi 39201 l '

Services, Inc.

630 Loyola Avenue, 3rd Floor Attorney General New Orleans, Louisiana 70113 Gartin Building -

' Jackson, Mississiopi 39205 i Mr. John G. Cesare Director, Nuclear Licensing Mr. Jack McMillan, Director  !

System Energy Resources, Inc. Division of Solid Waste Management .

Mississippi Department of Natural P. O. Box 23054  :

Jackson, Mississippi 39205 Resources Post Office Box 1C385 '

Mr. C. B. Hogg, Project Manager Jackson, Mississippi 39209  !

Bechtel Power Corporation  !

P. O. Box 2166 Alton B. Cobb, M.D.

, houston. Texas 77252-2166 State Health Officer J State Board of Health Mr. Ross C. Butcher P.O. Box 1700 i 3

Senior Resident Inspector Jackson, Mississippi 39205  !

i U. S. Nuclear Regulatory Commission l l Route 2. Box 399 President ,

i Port Gibson, Mississippt 39150 Claiborne County Board of Supervisors j 1 Port Gibson, Mississippi 39150 -

4 Nr. William T. Cottle  !

GGNS Site Director Regional Administrator, Region II I

System Energy Resources, Inc. L'. S. Nuclear Regulatory Commission !

P. O. Box 756 101 Marietta Street l Port Gibson, Mississippi 39150 Suite 2900 Atlanta, Georgia 30323  ;

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ENCLOSURE 1 ATTENDEES SERI-NRC 14EETING AUGUST 31. 1988 NAME AFFILIATION M. L. Crawford SERI J. O. Fowler SERI J. K. Fortenberry SERI L. L. Kintner NRC/P02-1 J. G. Cesare SERI F. W. Titus SERI W. K. Hughey SERI R. J. Wright SERI A. Chu NRC/PSB M. Hartzman NRC/ NEB H. W. Hodges NRC/RXSB i

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,..' . ha n:g jo,, UNITEDSTATES

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5 NUCLEAR REGULATORY COMMISSION wash WGTON, D. C. 20664

% ENCLOSURE 2 AGENDA FOR 8/31/86 MEETING hRC - SERI GRAND GULF NUCLEAR STATION. UNIT 1 REGARDING LICENSING ACTIVITIES ROOM 14 8 9 8:30 a.m. Review of TS changes Submittals, (NRC)

DG 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test, Snubber samples, l Oown travel cutoff of fuel hoist.

TS 3.0 and 4.0, RG 1.97 flux monitor  ;

RPS Surveillance intervals., Deletion of Fire '

Protection (GL 88-12), 10 CFR 55 Training TS, (

Fire Detection TS, Deletion of caily functional Test in RPCS, new action statement for SCRAM '

accumulator TS, revised PSRC composition. l 1

f 10:00 a.m. New TS Submittals (SERI) ,

- Remove position titles from TS 6.0

- Disposition of July 6, 1987 proposed change regarding isolation valve operability during shutdown 12:00 noon LUNCH 1:00 p.m. FollowupInformation(SERI) ,

Completion of OL Conditions - Status

- Procedures for using MSA-GMR-1 canisters

- IN 87-43 Gaps in high density spent fuel racks

- IN 86-81 Broken springs in MSIVS '

- Applicability of GL 83-28 responses to Unit 2 i

- Additional heat removal capability for Spent fuel '

- Check valve reliability i

2:00 p.m. Alternate Decay Heat Removal System (SERI)

Design Criteria Function. Safety Analysis l

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ENCLOSURE 3 i' NL 88/10 1 w ____

J.0 AMINISTRATIVE CONTROLS f

6.1 RESPONSIBILITY ,

6.1.1 The ^^"' "- - # anager M shall be responsible for overall unit operation and shall delegate in writing the succession to this respnnsibility during his absence.

6.1.2 The Shif t Superintendent or, during his obsence from the Control Rcom, a designated individual shall be responsible for the Control Room command function. A management directive te this effect sigr.ed by the Vie; freeident, N.ci... Opere iengshall be reissued to all station personnel on an annual

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(Senior 'Co'rporate' Nuclear OfI'ij 6.2 ORGANIZATION 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS i Onsite and offsite orgsnizations shall be established for unit organization ,

and corporate management, respectively. The onsite and offsite organizations' shall include the positions for activities affecting the safety of the nuclear

- power plant.

a. Lines of authority, responsibility, and communication shall be establishM and defined for the highest management levels through j

intermediate levels to and including all operating organization .

positions. These relationships shall be documented and upda?td, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms

of documentation. These requirements shall be documented in the ,

UFSAR and updated at least annually,

b. The 00N; C.necei % anager shall be responsible for overall unit safa 1 operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c. The#Vice Tresidea; " clear Opece;tene shall have corporate responsibility for overall plant nuclear safety and shall take any 1 Senior Corporate

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measures needed to ensure acceptable performance of the staff in Nuclear Officer operating, maintaining, and providing technical support to the plant to ensure nuclear safety, i d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite managar; however, they shall have sufficiant organizational f reecom to ensure their independence from operating i pressures.

6.2.2 UNIT STAFF  !

The unit organization shall be subject to the following; [

j a. Each on duty shift shall be composed of at lesst the stafaum shift crew composition shown in Table 6.2.2-1.

b. At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor. In addition, while the reactor is in OPERATIONAL CON 0! TION 1, 2 or 3, at least one licensed Senior Iteactor Operator shall be in the Control Room. .

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! GRAND GULF-UNIT 1 6-1 Amendment No. 45j

NL 88/10 ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)

c. A health physics technician

• shall be onsite when fuel is in the reactor,

d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator

< Limited to Fuel Handling who has no other concurrent responsibilities during tnis operacion.

e. A site Fire Brigade of at least 5 members shall be maintained onsite at all times". The Fi:e Brigade shall not include the Shift Superintendent, the STA, the two other memoers of the minimum shift crew necessary for safe shutdown of the unit, and any personnel  ;

required for other sssential functions during a fire emergency. At least one A0 shall be available to respond to non-fire-fighting commands from the Control Room.

f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions; e.g. , senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

Adequate shift coverage shall be maintained without routine heavy use of overtime. However, in the event that unforeseen problems require substantial amounts of overtime to be used, the following guidelines shall be followed:

1. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> j straight, excluding shift turnover time.

2. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven-day period, all excluding shift turnover time.

3. A break of at least eight hours should be allowed between work I periods, including shift turnover time.

4. Except during extenced shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authori:ed by the

, Manager or his designet, or higher levels of management, Plan.t/ in accorcance with established procedures and witn documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly The number of health physics technicians and Fire Brigade personnel say be l

1ess than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence provided immediate action is ta ,

to fill the required positions.

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! GRAND GULF-UNIT 1 6-2 Amendment No. 45)

,>' NL 88/10 ADMINISTRATIVE CONTROL 5 ,

UNIT STAFF (Continued) by the LJ ......f'rianager or his designes to assure that excessive hours have not been assigned. Routine deviation from tne above guide-lines is not authorized. ,

g. The coerations Suoerintendent, Shift Superintendents. Operations ,

Assistants, and Shift Supervisors shall each hold a Senior Aeactor  !

Operator License.

h. The Manager Plant Operations must have been a Senior Reactor Operator or have been certified on a plant of this type. l 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG1 FUNCTION  ;

6.2.3.1 The ISEG shal1 function to examine unit operating characteristics, l NRC issuances, industry advisories, Licensee Event Reports, and other sources l of plant design and operating experience information, including plants of 'similar l design, which may indicate areas for improving plant safety.

. . l COMPOSITION 6.2.3.2 The ISEG shall be composed of a multi-disciplined dedicated, onsite, group with a minimum assigned complement of five engineers or appropriate specialists.

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Amendment No. 45m

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NL 88/10 ,

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 prececures, equipment modifications, maintenance activities, operations activities or other means of improving unit safety to the Vis. ri .iceui, Nwcieec Ope,e;iens.

6.2.4 SHIFT TECHNICAL ADVISOR SeniorCorpor'ajehl'ea 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 OUALIFICATIONS de[aktach'ed'inse'rN_

p 6.3.1 Each member of the unit staff shall meet or exceed 34;1centd the minimum qualifica-requir:-

tions of ANSI N18.1-1971 forcomparablepositions[endth Tcnt; :pecified in R tica ' :nd C er Caclesoce ; ef ui. Maidi 20, 1303 NRC .i.i g, jd

M to d i 'iccn:::;, ex::st for the Chcaistc,/I,edietien Owoiivi 'o. g . ini.uw nt =mf n; shd i ::: cr cx::: the auelificetiona ef Ise;wie 6 .7 Owid. 1.0, 5.wi.... m .9#

bN!"1. . s'wM m h.o ss N..ib"1's,Y!30I_"N o ....w v wowiv m' } i'h ".'j vs

. . wi on N.i.www,3b' 7*5 ma ='.515'3b"5 ivww.r 4 i is t; di cperating n.cicac ;;-c. piente, and those members of the Independent Safety Engineering Group used for meeting the minimum complement specified in Sec-tion 6.2.3.2, each of whom shall have a Bachelo of Science degree or be regis-tered as a Professional Engineer and shall have at least two years experience in their field, at least one year of which experience shall be in the nuclear field.

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6.4 TRAINING k0 [c)redit[ation critelria]

6.4.1 A retraining andreplacementtrainingprograatforhtheunitstaffshallbe maintainedluncer the direction of the Trainingh.,ioteodent, shall meet or ex-ceed ' . 'r w i ANSI N13.1 1971 and m A;;;y n; -cauirc ;nt; :nd re ccendations of Cectieuncia " A" of

.g 10 CFR s;fi.d :n Part 55 Cenicas : end :

a : n e c 33 v r e ; e t n e r ei n x , = N r,: : e u.. a u ..:n3'u and shall include familiarization with relevant industry operational experience. j3 j

6.5 Review AND AUDIT 4 6.5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC)

FUNCTION i -

6.5.1.1 The PSRC shall function to advise the OGNO Conecel'.'iianager on all matters related to nuclear safety.

"Not resconsible for sign-off function. El

1. ucewi inet tM amerience and other training inf ormtion m : :a in wae sM licensee's letter to the w n ee=Jui, c., W ice acceotable for the iad'- idais iisted in that letter. (Deleie d) --

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H bssp; t*" the incividual icentified in MP&L's letter to tha C"O descu .pe,d December 11, 1985 is sun id: P a"ilifiad t: nwiu sne position of Chemistry / .a

,/ }e  ;

Radiation Control 4%ionTent baseo on sna og: ianna. education, and ctN. inrormation grovided or referenced in that letter.

GRAND GULF-UNIT 1 6-6 Amends t No.10cp b

N1. 88/10

,,,v,.,,,,, ..=q yRadiation Protection Manager Insert 1: " except for the'th;;;;;rj/";di;;i; . Oc.t. ;l L;;cic.;;c.J,c,; and Shift Technical Advisor, who shall meet or exceed the education and experience requirements of ANSI /ANS 3.1-1981 as endorsed by Regulatory Guide 1.8 Revision 2,1987, and licensed personnel who shall meet or exceed the criteria of the accredited license training program;"

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NL 88/10 ADMINISTRATIVE CONTROLS PLANT SAFETY REVIEV COMMITTEE (PSRC) ( ntj nued) m_ w _ - -

(75ipot e ~oT78 Wembers who manage in the onsite COMPOSITION \o,rganization at the Super,intendent level or 4bove,.

6.5.1.2 The PSRC shall beM : ::::d 9 t %

man:* Manager, Plant Support Vice / Member: Manager, Plant Operations Member: Manager, Plant Main  :

Member: tions 5 endent Member: Tech

• ort Superintendent ,

Member: ager, Qual i*c e s Member: Chemistry / Radiation Co uoerintendent ,

I&C Superintencent Member-r: (belehd Plant Licensing Suoerintendent ALTERNATES l 6.5.1.3 All alternate memoers shall be appointed in writing by the """' "---- '

Plant Manager to serve on a temporary basis; however, no more than two alternates  !

shall participate as voting members in PSRC setivities at any one time. l MEETING FREQUENCY 6.5.1.4 The PSRC shall meet at least once per calendar month and as convened by the PSRC Chairman or Vice Chairman.  !

L QUORUM

< 6.5.1.5 The quorum of the PSRC necessary for the performance of the PSRC f

responsibility and authority provisions of these Technical Specifications shall  :

consist of the Chairman or Vice Chairman and four memoers including alternates. l RESPONSIBILITIES l

6.5.1.6 The PSRC shall be responsible for review of: i

a. Station administrative procedures and changes thereto, ,

b. The safety evaluations for (1) procedures, (2) changes to procedures, l equipment or systems, and (3) tests or experiments completed under r the provision of Section 50.59, 10 CFR, to verify that such actions (

did not constitute an unreviewed safety question and all programs j

i required by Specification 6.8 and changes thereto. .

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c. Procesed procedures and changes to procedures, equipment or srttus which may involve an unreviewed safety question as def..it,J in Section 50.59, 10 CFR.

d. Proposed tests or experiments which may involve an unreviewed nfe';- I I

question as defined in f action 50.59, 10 CFR.

e. Proposed changes to TecMical' Specifications or the uper::im i kense.

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• GRAND GULF-UNIT 1 6-7 ken 6ent He vt.e ._ i

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NL 88/10 ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued)

f. Reports of violations of codes, regulations, orders, Technical Specifications, or Operating License requirements having n.: clear safety significance or reports of abnormal degradation of systems designed to contain radioactive material.

g. Reports of significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

h. Review of all REPORTABLE EVENTS.

1. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, l i

or components. .

j. The plant Security Plan and changes thereto.

k. The Emergency Plan and changes thereto. .

1. Items which may constitute a potential nuclear safety hazard as identified during review of facility operations,

m. Investigations or analyses of special subjects as requested by the Chairman of the Safety Review Cormittee.

n. Changes to the PROCESS CChTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste systems. .

)

< AUTHORITY q 6.5.1.7 The PSRC shall:

a. Reccenend in writing to the 00Z C,o...fManager approval or disapproval of items considered under 6.5.1.6.a. c, d, e, j, and k, above.

Plant b. Render determinations in writing to the C .N", 0,.,ere%anager with regard to whether or not each item considered under 6.5 .6.a. c and d, above, constitutes an unreviewed safety question. Plant 6

c. Provide written notification within 24/ hours to the SRC of disagreement between the PSRC and the COMO 0;mmi"Nanager;'however, the il0We j EW4 Manager shr.11 have responsibility for resolution of such ,

OPlant disagreements pursuant to 6.1.1 above. (

RECOROS 6.5.1.8 The PSRC shall maintain written minutes of each PSRC meeting that, at 4

i a minimum, document the results of all PSRC activities performed under the i

! responsibility and autnerity provisions of these Te.:nnical Specifications, J Copies shall be provided to the SRC.

s 1

GRAND GULF-UNIT 1 6-8 Amendment No.

1

NL 88/10 ADMINISTRATIVE CONTROLS

- ~ ~

6. 5. 2 SAFETY REVIEV Co miTTEE (SRC)

FUNCTION 6.5.2.1 designated activities in the areas of:The SRC shall function to provide in

a. nuclear power plant operations

b. nuclear engineering

c. chemistry and radiochemistry

d. eetallurgy

e. instrumentation and control *

f. radiological safety g, '

j mechanical and electrical engineering #

\

h. quality assurance practices

• COMPOSITION j

• Wn e 10 members ho shall v

6.5.2.2 The SRC shall be composed of Y tte. [exceepthe or i s cation Cha Vice Prestcent., nucica,w.. .Lrectf r

- s of Secti t

Messer: . . . 7 -

Vice President, Nuclear Engineering 3,g.

Memeer: or > 1981 4

Member:

or, Nuclear Plant Engine Site D r GGNS ~. .

Meber: Director,Qua Member: ans (Deleied)

Designated senta Member: GGNS MSU System Services, In .

al Manager Member:

Member:

ector Nuclear Licensing Manager , Radiological and Environmental e s

Manager,, Operational Analysis I .  ;

Resources, Inc, <:onsistent with the recomendations o on Reactor Safeguards letter, Mark to Palladino dated October 20, 1951.

• 1 science field or equivalent experience and a minimum o disciplines of 6.5.2.la through h. experience of which a minimum of th; In the aggregate, the membership of the j

disciplines of 6.5.2.la through h.

comittee shall provide specific g q,7,l practical M experien

_ ALTERNATES ( g pa#

t/p.AW '

6.5.2.3 All alternate mesters shall be appointed in writing by the SRC i Chairean to serve on a te.eperary basis; however, no more than two alternates shall participate as voting memsers in SRC activities at any one time.

r i

GRAND GULF-UNIT 1 Y

6-9 Amendment No. 12, _

4 NL 88/10 ADMINISTRATIVE CONTROLS AUO!TS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the SRC. These audits shall encomoass:

a. The conformance of unit operation to provisions contained within the Appenoix A Technical Specifications and applicable license conditions l at least once per 12 months.

1

b. The performance, training and qualifications of the entire unit staff at least once per 12 months.

c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months,

d. The performance of activities required by the Operational Quality (

Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months. t I

e. The Emergency Plan and implementing procedures at least once per  !

12 months,

f. The Security Plan and implementing procedures at least once per 1 1

12 months.  !

t

g. Any other area of unit operation considered appropriate by the SRC j or thep' ice P. eef dent, Lc';;r 4retiene.

I e h. The Fire Protection Program and implementing procedures at least once i r Officer per 24 months.

l 1. An independent fire protection and loss prevention inspection and

audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.

j. An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant i) at intervals no greater than 36 months. i i

! k. The radiological environmental monitoring program and the results  :

thereof at least once per 12 months.  !

]

i

1. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.  !

m. The PROCESS CONTROL PROGRAM and implementing procedures for ,

i solidification of radioactive wastes at least once per 24 months. l l n. The performance of activities required by the Quality Assurance i Program to meet the criteria of Regulatory Guide 4.15, February 1979. ,

' at least once per 12 months. -

GRAND GULF-UNIT 1 6-11 Amendment No.103 l

. ,, [

NL 88/10 ADMINISTRATIVE CONTROLS i orh N[c'le' 6.5.2.9 The SRC shall report to and advise thehice fregi4;a:TTL;l;;r -

0;;r;ti = on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8..

i RECORDS 6.5.2.10 Records of SRC activities shall be prepared, approved and distributed as indicated below:

a. Minutes of each SRC meeting shall be prepared, approved and forwarded to the m . 7..eident, t6cieer Operatien; _hin 14 d g wing nior_ Corporate Nuclear Off, each meeting,

b. Reports of reviews encompassed by Section l6.5. 7 above, shall be prepared, approved and forwarded to the # c. 7,esident, tL;1;;r l

Op.retien; within 14 days following completion of the review.

-+ c. . wc . Audit reeorts encompassed by Section 6.5.2.8 above, shall be Senior Corporate forwarceo to tnetVic. Presideas, Lci;;r Op;retiea; and to the (;

! Nuclear Officer management positions responsible for the areas audited within 30' days

! after completion of the audit by the auditing organization. l 6.5.3 TECHNICAL REVIEV AND CONTROL ACTIVITIES 1 6.5.3.1 Activities which affect nuclear safety shall be conducted as follows:

a. Procedures required by Technical Specification 6.8 and other proce-1 dures which affect plant nuclear safety, and changes thereto, shall (

be prepared, reviewed and approved. Each such procecure or procedure l j change shall be reviewed by an individual / group other than the

' individual / group which prepared the procedure or procecure change, but who may be from the same organization as the individual / group '

which prepared the procedure or procedure change. Proceduros othe than Administrative Procedures shall be approved as delinanted in PI_a nt  :

lant writing ey the ".0" 0;ner;i Manager. The "" " ' Manager shall  !

approve administrative procecures, security implementing procacures l

and emergency plant implementing precedures. Temporary approval to l

procedures which clearly do not change the intent of the approved procedures may be made by two members of the plant manegement staff, at least one of whom holds a Senior Reactor Operator's License.

i Temporary changes shall be reviewed by the reviewing authority within l 14 days of being issued. For changes to procedures which may involve a change in intent of the approved procedures, the person authorized '

above to approve the procedure shall approve the change.

I 6 -r b. Proposed changes or modifications to plant nuclear safety-related Senior Onsite structures, systems and components shall be reviewed as designated by thenit. Oi.eciei 00 % Each such modification shall be reviewed j

Management Re r enta tivV by an individual / group other than the individual / group which designed the modification, but who may be from the sa.se organization as the individual / group which designed the modifications. Implementation of proposed modifications to plant nucleaa safety-related structures, systems and components shall be approved by the y 0;.;rei Manager. l Plant GRAND GULF-UNIT 1 6-12 A.mendment N 0l j

(

e* NL 88/10 ADMINISTRATIVE CONTROLS ACTIVITIES (Continued)

c. Proposed tests and experiments which affect plant nuclear safety and are not addressed in the Final Safety Analysis Report shall be reviewed by an individral/ group other than the individual / group which prepared the proposed test or experiment. ,

d. Events reportable pursuant to Section 50,73 to 10 CFR Part 50 shall be investigated and a report prepared which evaluates the event and whict crovides recommendations to prevent recurrence. Such report '

lant shall be approved by the'00NS Oer,,ie' Manager.

e. Individuals responsible for reviews performed in accordance with

> Senior Onsite 6.5.3.1.a. 6.5.3.1.b, 6.5.3.1.c and 6.5.3.1.d shall meet or exceed the cualification recuirements of Section 4.4 of ANSI 18.1, 1971,

(' ManagerrentRepresentative) as previously designated by the' Site Oirecter, 00NS or 00NS 0; ;ref

- - p'PManager, as applicable. Each such review shall include a determina-tion of whether or not additional, cross-disciplinary review is necessary. If deemed necessary, such review shall be performed by

+

the review personnel of the appropriate discipline,

f. Each review shall include a determination of whether or not an .

unreviewed safety question is involved.

g. Records of the above activities shall be provided to the COMO Oer,erei=

> Manager, PSRC and/or as necessary for required reviews.

l 6.6 REPORTABLE EVENT ACTION j 6.6.1 The following actions shall be taken for REPORTABLE EVENTS: l l

a. The Commission shall be notified pursuant to the requirements of l i Section 50.72 to 10 CFR Part 50, and a report submitted pursuant t to the requirements of Section 50.73 to 10 CFR Part 50, and  ;

b. Each REPORTABLE EVENT shall be reviewed by the PSRC and submitted

to the SRC and the Vice Tieeider,t._M-cieer Oy...i.ivu.. l1t 4

6.7 SAFETY LIMIT VIOLATION Qen{or]o[p a[e ea} fQ I 6.7.1 The following actions shall be taken in the event a Safety Limit is 1 j violated:

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice Ti. eider.t, Senior Cohpg =  ; y and the SRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l Nuclear b. A Safety Limit Violation Report shall be prepared. The report sna11  !

t Officer be reviewed by the PSRC. This report shall describe (1) applicable  :

)

circumstances preceding the violation, (2) effects of the violation upon unit components, systems or structures, and (3) corrective action taken to prevent recurrence, j j -

i I

). ,

GRAND GULF-UNIT 1 Amendment No. 10j

NL 88/10 ADMINISTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued) '

c. The Safety Limit Violation Report shall be submitted to the Commission,theSRCandthegaceFc..idec.;,L.ci.-

14 days of the violation. " ,e;iec. within 6(( C[o[po[r Critical operation of the unit shall not be resumed until authorized i

d.

by the Commission.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented maintained l

covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory j Guide 1.33, Revision 2. February 1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation. -

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation. f

h. OFFSITE DOSE CALCULATION MANUAL implementation.  !

] i. Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 4.15, February 1979. '

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed

as required by 6.5, above, prior to implementation and shall be reviewed periodically as set forth in administrative procedures.  !

6.8.3 The following programs shall be established, implemented, and maintained:

) 4. Primary Coolant Sources Outside Containment i l A program to reduce leakage from those portions of systems outside

' )

containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the:

1. RCIC system outside containment containing steam or watar, except the drain line to the main condenser.

2. RHR systes outside containment containing steam or water, except j the line to the LRV system and headers that are isolated by j manual valves. ,

3, HPCS system.

4. LPCS system.

5. Hydrogen analyzers of the combustible gas control system.

GRAND GULF-UNIT 1 6-14 Amendment No.10 j It

ee O

. ENCLOSURE 4 ALTERNATE DECAY HEAT REMOVAL SYSTEM INTRODUCTION DESIGN 0BJECT!YES AND REQUIREMENTS SYSTEM DESCRIPTION OPERATING MODES DESIGN RCv!EWS CONCLUSIONS / QUESTIONS I NMAUNC 0206 ASDCS PRESENTATION

• • O l

l l

INTRODUCTION 1

TECHNICAL SPECIFICATION 3/41.4.9 Basts:

, REFUELING RESIDUAL HEAT REMOVAL ALTERNATE DECAY HLAT REMOVAL HETH0ns i

PREVIOUS ALTERNATIVE DECAY HEAT REMOVAL METH0ns

l i REACTQA WATER CLEANUP 1

j FUEL POOL COOLING ANJ CLEANUP t

CRD SYSTEM

{

- p I

• PROBLEM AREAS i i

l!MITED CAPACITY i

• OPERATING FLEXtBILITY i

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NMAUNC 0206 ASDCS PRESENTATION (];) j

,. . .' j 1

l M. W /.iE DECAY HEAT REMOVAL SYSTEM l

l des!GN OsJECTIVES l

i ALTERNATE DECAY HEAT RnM0 VAL CAPACITY AvA!LAstE sY ,

THE END OF OUTAGE DAY ;. i As INDEPENDENT As P0ssistE FROM OTHER PLANT SYSTEMS des!GN REQUIREMENTS MAINTA1NINGIEMPERATURg1.!MITS IN IECHNICAL SPECIFICATION IABLE 1. j 1, 200F DURING MODE 4 .

2. 140F DuRING MODE 5  ;

l CPERATIONAL IN MODES 4 AND 5 ONLY  !

i NO SAFETY FUNCTION RELATED Iot

1. SHUTDOWN CAPAs!LITY i

2. ACCIDENT MIT!GATION 1

NO ADVERSE INTERACTION WITH Ex!$T!No PLANT SYSTEMS  ;

• PRES $UAE BOUNDARY - ASME SECTION !!! class 3, SE!$MIC CATEGORY I  !

i l

• OPERATED FROM THE CONTROL ROOM i

l NMAUNC 0206 ASDCS PRESENTATION

= -

t SYSTEM DESCRIPTION 4

PRIMARY FLOWPATHI l

SUCTION PATHS RHR COMMON SUCTION f

SPENT FufL POOL

• Two pumps AND HEAT ExCHANGERS

• D!$ CHARGE PATH: '

RHR "C" i i

! SYSTEM INTERFACE  !

.l I

SECONDARY FLOWPATH1

• - PLANT SERVICE WATER SYSTEM ,

1 RADIATION MON! TOR ON EFFLUENT 1

• i ELECTRICAL power: i NON-class IE i

i EXCEPTION SYSTEM ls0LAT!0N VALVE ,

CONTROL AND INSTRUMENTATIONI  !:

FL0w CONTROL VALVE

, IEMPERATURE INDICATION l 1 1 PUMP CONTROLS AND INDICATION  !

HVAC:

J AIR HANDLING UNIT  !

I

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• PSW SUPPLIED 4

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OPERATING MODES LAYUP

• MODES 1, 2, 3 MECHANICALLY AND ELECTAICALLY ISOLATED FROM INTERFACING SYSTEMS FLUSH /IEST SUPPRESSION POOL TO 'S!JPPRESSION POOL RPV TO RPV COOLING NORMAL COOLING l!NEUP ,

COMMON SUCTION TO LPCI "C" FLOWPATH

• SPENT PUEL POOL TO RPV COOLING USED WHEN "COMMON-SUCTION" LINE UNAVAILABLE f

9 1

b

DESIGN REVIEWS AND SYSTEM INTERACTION EVALVATION i l

INTERFACING SYSTEMS CRITERI.A CONDUCTED IN TWo PARTS ,

FUNCTIONAL INTERACTIONS PHYSICAL INTERACi10NS i

I e

1 i

I

i FUNCTIONAL INTERACTION EVALUATION MAINTAINING THE OPERABILITY OF SAFETY-RELATED SYSTEMS AND FUNCTIONS OPERATING MODES AND COMBINATIONS OF CPERATING MODES POTENTIAL FOR INADVERTENT DRAINAGE CONTROLS AMD OPERATIONAL INTERACTIONS 9

l i

t e

l l

l l

1

, i NMAUNC 0206 ASDCS PRESENTATION

~ .

I.

POTENTIAL FOR INADVERTENT DRAINAGE EVALUATION CRITERIA SINGLE ACTIVE COMPONENT FAILURE SINGLE OPERATOR ERROR EVALUATION REVIEW DRAIN PATHS OVER 1 INCH REVIEW OPERATING M0DE COMBINATIONS P,SULTS T

FEEDBACK INTO DESIGN CHANGE CLARIFIED ADMINISTRAT!YE/ PROCEDURAL CONTROLS  ;

REQUIREMENTS CONCLUSION ALTERNATE DECAY HEAT REMOVAL SYSTEM PRESENTS NO GREATER POTENTIAL FOR INADVERTENT DRAINING THAN EXISTING SYSTEMS 1

0 1

i l xxAune o20s Asocs rassen:Arion 3

s_ -

i e.

,s.....

)

1 CONTROLS /0PERATIONAL INTERACTION EVALUATION EVALUATION CRITERIA PREVENT ADVERSE IMPACT ON SAFETY RELATED SYSTEMS ASSUME SINGLE ACTIVE FAILURE SINGLE OPERATOR ERROR EVALUATION.

IDENTIFY POTENTIAL FUNCTIONAL INTERFACES CONSIDER OPERATING MODE COMBINATIONS CONSIDER DESIGN BASIS EVENTS / ACCIDENTS DEVELOPED CONTROL INTERACTION MATRIX  ;

CONCLUSION ALTERNATE DECAY HEAT REMOVAL SYSTEM lhrRODUCES NO ADVERSE CONTROL /0PERATIONAL INTERACTIONS GW

4 m.

o s...

PHYSICAL INTERACTIONS EVALUATION

• EFFECTS ON PLANT AMBIENT CONDITIONS (TEMPERATURE, CHEMISTRY, ETC.)

• IMPOSED LOADINGS (NORMAL, TRANS!ENT, SEISMIC, ETC.

LOADS)

HAZARDS CONDITIONS (LINE BREAKS, FLOODING / SPRAY, MISSILES, FIRE, EYC.)

PROCESS CONDITIONS (EFFECTS ON REACTOR WATER FLOW, TEHPERATURE, PRESSORE, ETC.)

RADIOLOGICAL EFFECTS 1

1 l

l 1 .

NMAUNC 0206 ASDCS PRESENTATION (2

3 ,

dM i

DISTRIBUTION FOR f tEETING SUldt4ARY DATED: October 18, 1988 Facility: Grand Gulf fluclear Station

Docket File '

NRC PDR Local PDR PDIl-1 Reading E. Adensam P. Anderson A. Chu (8D1) fl. Hartzman (9H3)

W. Hodges (8E23)

OGC E. Jordan (l:NBB3302)

B. Grimes (9A2)

ACRS (10)

B. Troskoski (17D19) cc: Licensee / Applicant Service List 0f i(