ML20127E648
| ML20127E648 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/30/1985 |
| From: | CONSOLIDATED EDISON CO. OF NEW YORK, INC. |
| To: | |
| Shared Package | |
| ML20127E603 | List: |
| References | |
| NUDOCS 8506240607 | |
| Download: ML20127E648 (23) | |
Text
_ _
'9 -
ATTACHMENT A APPLICATION FOR AMENDMENT TO OPERATING LICENSE Technical Specification Page Revisions Consolidated Edison Company of New York, Inc.
l Indian Point Unit No. 1 Docket No. 50-3 I Facility Operating License No. DPR-5 June, 1985 l
l l
l 8506240607 8506 3 j PDR ADOCK 0500 p
( P i
s E E T T I I S S F N F
O mO a
=
=
- T N R m E E AR w D tE C I t N lv S tO
= E R
uP N
= P
. = n
= o i
- t
- a z
. i S
= n R El f t t = a A l l il m g Ei L E l - .
. r CI U C L F A
S n
n - .
. O o .
N A C - -
F t
. r
. o
- . p
. p
- u
- - S a .
a .
l pggIIIIIIllIg388 a
- gIB8IJ c
- - i
- - n I
- h N
E S - c I
N t
v D I L0 N
a - e A
E D 1 S 1
ut R ll I
Tg % gI%4 - T N a I
i S t fF R -
t t E E - d R
F t u
E C
C O
P - n I
- - a V
- t
- - n
- - e
- - m e
- g
- - a
- - n
- - a M
- - y
- t
- - i
- l
- i
- - c
- - a
- - F
- I 1
G - 3 N E t C l E l C - l INRI Nl Nl h N e E
t i
Ai R L R Ai R t - t 6 N h r D B O u 3
I N u!
s0 IO ul
$0 - t P 5 u t
G sA C SA C i g R N aI E AI . h R h
( L L h i P t E R
I E - a h I E R liR I
D I R n t i I I F C
l (
H t D ItD - C u L v t t N N uA nh uA - N R u
u D G
- d N
Y3.oD r &GDf t
Zo.
ATTACHMENT B APPLICATION FOR AMENDMENT TO OPERATING LICENSE Technical Specification Page Revisions l
t Consolidated Edison Company of New York, Inc.
Indian Point Unit No. 2 Docket No. 50-247 Facility Operating License No. DPR-26 June, 1985 l
l l
n l
l TABLE OF CONTENTS (Cont'd)
Section Title Page Nuclear Facilities Safety Committee 6-7
. 1) Function 6-7
- 2) Composition 6-8
- 3) Alternates 6-8
- 4) Consultants 6-9
- 5) Meeting Frequency 6-9
- 6) Quorum 6-9
- 7) Review 6-10
- 8) -Audits 6-11
- 9) Authority 6-11
- 10) Records 6-11 66 Reportable Event Action 6-12 67 Safety Limit violation 6-12 68 Procedures 6-12 6.9 Reporting Requirements 6-13
-Routine Reports 6-13 Startup Reports 6-14 Annual Radiological Env. Operating 6-15 Report Semiannual Radioactive Effluent Report 6-16 Monthly Operating Report 6-18 Special Report 6-18 '
6.10 Record Retention 6-19 6.11 Radiation Protection Program 6-20 6 12 High Radiation Area 6-20 6 13 Environmental Qualification 6-21 6 14 Process Control Program (PCP) 6-21 6.15 Offsite Dose Calculation Manual (ODCM) 6-21 6.16 Major Changes to Radioactive Wastes Systems 6-22 I
Amendment No. iv i
I 4
I.
i I
-_,_--~, -- , , , - , , . . - - . . , , , . ,,__ ,. ~ -. . , , - . , - , , - - . - . . . _ . . . - _ - _ . . - . . - . _ _ . . _ _ _ _ _ - . , _ . . . , _ . _ - - - - - - _ - _ - .
60 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Vice President-Nuclear Power shall be responsible for overall facility activities as shown in Figure 6 2-2 and shall delegate in writing the succession to this responsibility during his absence.
6.1.2 The General Manager-Nuclear Power Generation shall be responsible for facility operations as shown in Figure 6.2-2 and shall delegate in writing the succession to this responsibility during his absence.
62 ORGANIZATION FACILITY MANAGEMENT AND TECHNICAL SUPPORT 6 2.1 The organization for facility management and technical support shall be as shown on Figure 6.2-1.
FACILITY STAFF 6 2.2 The Facility organization shall be as shown on Figure 6 2-2 ands
- a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6 2-1.
- b. At least one licensed Operator shall be in the control room when fuel is in the reactor.
- c. At least two licensed Operators shall be present in the control room during reactor startup, scheduled reactor shutdown, and during recovery from reactor trips.
- d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
- e. ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling. This individual shall have no other concurrent responsibilities during this operation.
Amendment No. 6-1
, ** i
- f. A Fire Brigade of at least five members shall be maintained on site at all times *. This excludes four members of the minimum shift crew necessary for safe shutdown of the plant and any personnel required for other essential functions during a fire emergency.
During periods of cold shutdown, the Fire Brigade will exclude two - hers of the minimum shift crew.
shall be developed 'and
- g. Administrittive procedures implemented to limit the working hours of unit staff who perform safety-related functions (e.g., licensed Senior auxiliary Operators, licensed Operators, health physicists, operators, and key maintenance personnel). .
The amount of overtime worked by unit staff members performing safety-related functions shall be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).
\
l
(
- F3.re Brigade composition may be one member less 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 of Fire Brigade members provided immediate action is taken to restore the Fire- Brigade to within the minimum l
requirements.
j l
s l . .
l
)
6-1(a)
Amendment No.
i
~ - ~ * - - - - - - _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _
~ - - - - _ _ - _ - _ _ _ _ _ _ .
~ .
E - E T T I
S F
- SI F -N O
- O-
- T
- N n
( 0t Ah w ( l Et
. L n 1 S Co
- 3 E R
uP m
- P
- n
- o i
- t .
- . a
- - z i
- - n S
t g
- - a Mt i 5
1 tg
. g Ei L t i
r CI u
L Mnl n .. O mC h S o C -
F . t
. r
- .. o
- p
. p
- . u
- .. S
. l
. a
- ygg3iIglEI3EgaBIagIIaIJ c
- i
, - n T
- h N
S - c t
u t
u OI lD a
u - - e t
o ll S n l ih1 - T k ! I'l
)ggItA l
l t
S uMI C P naA l 1 \
t E i E - d a
P X
E E L P - n C
1 O
- - a v
- t n
- e m
- e g
a n
a
- - M y
t
- - i
- l
- i
- c
- F a
- - 1 2
l m
G N
I R
E C I mI nI t
ci ev ni
~ h m t h E
C 6
~
f
[
0 t n
nL u1 n o
nt ut n mR 0 U e 4SGi $8 SA i
$e Sn - e G
P S S
r Mu eI C t
ai a m
RA R
u P E t
L n L
- a E I g A t M : :M 0 n L II i E
C R t t iL0 to - C U L F v Ca em nn - NR u
uA ua Q u Q Q m
aD&aDt cZ' O .
G8JB l
63 FACILITY STAFF OUALIFICATIONS 6.3.1 Each mem'ber of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiation Protection Manager who shall meet or exceed the minimum qualifications of Regulatory Guide 1.8, September 1975.
632 The General Manager-Nuclear Power Generation shall meet or exceed the minimum qualifications specified for Plant Manager in ANSI N18.1-1971.
633 The Shift Technica l Advisor (STA) shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.
64 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Nuclear Training Director and shall meet or exceed the requirements and recommendations of Section 55 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6 4.2 A training program for the Fire Brigade shall be maintained under the direction of the Nuclear Training Director and shall meet or oxceed the requirements of Section 27 of the NFPA Code-1976 with the exception of the training program schedule.
65 REVIEW AND AUDIT .
6.5.1 STATION NUCLEAR SAFETY COMMITTEE (SNSC)
FUNCTION 6 5.1.1 The Station Nuclear Safety Commitee shall function to advise the Vice President-Nuclear Power on all matters related to nuclear safety.
Amendment No. 6-5
COMPOSITION 6 5.1 2 The Station Nuclear Safety Committee shall, as a minimum, be composed as follows:
Chairman: General Manager-Technical Support Member: Chief Technical Engineer Member: Operations Manager Member: Maintenance Engineer Member: ' Instrument and Control Engineer Member: Radiation Protection Manager Member: Reactor Engineer 6.5.1.2.1 In addition, other technically competent individuals may be appointed by the SNSC Chairman to serve as SNSC members.
ALTERNATES 6.5.1 3 Alternate members shall be appointed in writing by the SNSC Chairman to serve on a temporary basis, and must have qualifications similar to the member being replaced. .
MEETING FREQUENCY 6.5.1 4 The SNSC shall meet at least once per calendar month and as convened by the SNSC Chairman or his designated alternate.
QUORUM 6.5.1.5 A quorum of the SNSC shall consist of the Chairman or his designated alternate and four members. No more than two alternates shall be included in the quorum.
RESPONSIBILITIES 6.5.1.6 The Station Nuclear Safety Committee shall be responsible for: ,
- a. Review of 1) all procedures required by Specification 6.8 and changes thereto, and 2) any other proposed procedures or changes thereto as determined by the General Manager-Technical Support to affect nuclear safety.
- b. Review of all proposed tests and experiments that affect nuclear safety.
- c. Review of all proposed changes to the Technical Specifications.
Amendment No. 6-6
, ** a
- d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
- e. Investigation of all violations of the Technical Specifications and preparation and forwarding of a report covering evaluation and recommendations to prevent recurrence to the Vice President-Nuclear Power and to the Chairman of the Nuclear Facilities Sa'fety Committee.
- f. Review of facility operations to dc F.ect potential nuclear safety hazards.
- g. Performance of special reviews and inveutigations and the issuance of reports thereon as required by the Chairman of the Nuclear Facilities Safety Committee.
- h. Review of the Plant Security Plan and implementing procedures and submission of recommended changes to the Chairman of the Nuclear Facilities Safety Committee.
- i. Review of the Emergency Plan and implementing procedures and submission of recommended changes to the Chairman of the Nuclear Facilities Safety Committee.
- j. Review of any unplanned, radioactive release, including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President -
Nuclear Power and to the Nuclear Facility Safety Committee.
- k. Review of changes to the Process Control Program and the Offsite Dose Calculation Manual.
Amendment No. 6-6(a)
REVIEW (Continued) 9 Reportable Events, as specified by 10 CFR 50.73. ,
- h. Any indication of an unanticipated deficiency in some aspect of design, or operation of safety related structures, systems, or components.
- 1. Reports and meeting minutes of the Station Nuclear Safety Committee.
- j. Environmental surveillance program pertaining to radiological matters.
AUDITS 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the NFSC. These audits shall encompass:
- a. The conformance of facility operation to all provisions contained within the Radiological Technical Specifications (Appendix A) and applicable license conditions at least once per 12 months.
- b. The conformance to all provisions contained within the Environmental Technical Specifications (Appendix B) pertaining to radiological matters and applicable license conditions at least once per 12 months.
- c. The performance, training and qualifications of the entire facility staff at least once per 12 months.
- d. The results of all actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.
- e. The performance of all activities required by the Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.
- f. The Facility Emergency Plan and implementing procedures at least once per 12 months.
[
- g. The Facility Security Plan and implementing procedures at least once per 12 months. g
- h. The Facility Fire Protection Program and implementing procedures at least once per 24 months.
Amendment No. 6-10
., 'a e
- d. Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1, April 1975.
6.8.2 Each procedure and administrative policy of 6 8.1 above, and any changes to them shall be reviewed and approved for implementation in accordance with a written administrative control procedure approved by the appropriate General Manager," with the concurrence of the Station Nuclear Safety Committee and the Nuclear Facilities Safety Committee. The administrative control procedure required by this specification shall, as a minimum, require that:
- a. Each proposed procedure / procedure change involving safety related components and/or operation of same receives a pre-implementation review by the SNSC except in case of an emergency.
- b. Each proposed procedure / procedure change which renders or may render Final Safety Analysis Report or subsequent safety analysis report inaccurate and those which involve or may involve potential unreviewed safety questions are approved by the SNSC prior to implementation.
,c. The approval of the Nuclear Facilities Safety Committee shall be sought if, following its review, the Station Nuclear Safety Committee, finds that the proposed procedure / procedure changes either involves an unreviewed safety cuestion or if it is in doubt as to whether or not an unreviewed safety question is involved.
6.8.3 A mechanism shall exist for making temporary changes and they shall only be made by approved management personnel in accordance with the requirements of ANSI 18 7-1972. The change shall be documented, and reviewed by the SNSC within 14 days of implementation.
69 Reporting Requirements Routine Reports 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the NRC Regional Administrator of the Region I Office unless otherwise noted.
Amendment No. 6-13
a 's e Monthly Operating Report 6 9.1 7 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or Pressurizer safety valves shall be submitted on a monthly basis to the Director, office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D C. 20555, no later than the 15th of each month following the calendar month covered by the report.
Special Reports 6.9.2 Special reports shall be submitted to the NRC Regional Administrator of the Region I Cffice within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specifications
- a. Each containment integrated leak rate test shall be the subject of a sununary technical report including results of the local leak rate test since the last report. The report shall include analyses and interpretations of the results which demonstrato compliance in meeting the leak rate limits specified in the Technical Specifications.
- b. Inoperable ,
- fire protection and detection equipment (Specificat' ion 3.13).
- c. Sealed source leakage in excess of limits (Specification 4.15).
.d. The complete results of the steam generator tube inservice inspection (Specification 4.13.C).
- e. Radioactive effluents (Specification 3.9).
- f. Radiological environment monitoring (Specification 4.11)
- g. Meteorological monitoring instrumentation (Specification 3.15).
Amendment No. 6-18
6.10 Record Retention 6.10.1 The following records shall be retained for at least five years:
- a. Records ar.d logs of facility operation covering time interval at each power level.
- b. Records and logs of principal maintenance activities, i
inspectionsi repair and replacement of principal items of equipment related to nuclear safety.
- c. Reportable Event Reports.
- d. Records of surveillance activities, inspections and 2
calibrations required by these Technical Specifications.
- e. Records of reactor tests and experiments.
- f. Records of changes made to Operating Procedures.
I
- g. Records of radioactive shipments.
- h. Records of sealed source leak tests and rescits.
- 1. Records of annual physical inventory of all source material or record.
i 6.10 2 The following record shall be retained for the duration of the Facility Operating Licenser
- 4. Record and drawing changes reflecting facility design ,
modifications made to systems and equipment described in the Final Safety Analysis Report.
I b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
- c. Records of facility radiation and contamination surveys.
d
- d. Records of radiation exposure for all individuals entering radiation control areas.
- e. Records of gaseous and liquid radioactive material releases to the environs.
- f. Records of transient or operational cycles for thoce facility components designed for a limited number of transients or cycles.
j Amendment No. 6-19 s
i
-_ - . , _ _ __-._.,,.-__,_,.___._.___,_m-_,,_,_.~_ , _ . , _ _ _ , . - - _ , . _ _ _ , _ . . _ _ . - . _ _ ,,._,.,mr .m.m . . _ - - _ _ ,
. . . . - ~. -_. . ..-~_- ~. - . - - - - - - . - . . . . . _ _ - - - . - .
- . e.
I' i
{ Record Retention (continued)
Records of training and qualification for current members of 4
g.
the plant staff.
- h. . Records of in-service inspections performed pursuant to these Technical Specifications.
- i. Records of. Quality Assurance activities required by the QA Manual except as noted in 6 10.1.
- j. Records of reviews performed for changes made to procedures or j equipment or reviews of tests and experiments pursuant to 10 l - CFR 50.59.
I
- k. Records of meetings of the SNSC and the NFSC.
- 1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13.
{ ,
- m. Record of analyses required by the radiological environmental
- ' the monitoring program that would permit evaluation of accuracy of the analysis at a later date. This should include i procedures effective at specified times and QA records showing that these procedures were followed.
6.11 Radiation Protection Program e
! Procedure for personnel radiation protection shall be prepared
! consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
6.12 High Radiation Area i 6.12.1 As an acceptable alternate to the " control device" or " alarm l
signal" required by paragraph 20 203(c)(2) of 10 CFR 20: ,
- a. Each High Radiation Area in which the intensity of radiation is greater tha6 100 mres/hr but less than 1000 mres/hr shall l' l be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation Work Permit and any individual or group of
- individuale permitted to enter such areas shall be provided l with - a radiation monitoring device which continuously indicates the radiacion dose rate in the area. ,
- b. Each High Radiation Area in which the intensity of radiation is greater than 1000 mrem /hr shall be subject to the provisions of 6.12.l(a) above, and in addition locked doors
- shall be provided to prevent unauthorized entry to such areas and the keys shall be maintained under the administrative l control of the Watch Supervisor on duty.
i .
l Amendment No. 6-20 l
l.
sw-y e.e- .- c ,-.--m,.,,,..uy. ,... .-.~e.,--.-,--m.,. e,%my-.-,,.g.,%,, ,y ~,,-.rv.e%,we,--.a.w -.=ww.
l i
- o o I
l 6.13 Environmental Qualification 6.13.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions oft Division of Operating Reactors
" Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," December 1979. Copies of these documents are attached to Order for Modification of License No. DPR-26 dated October 24, 1980.
6.13 2. By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines of NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.
6.14 Process Control Program (PCP) 6.14.1 Licensee initiated changes to the PCP:
- 1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contains
- a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental informations
- b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the (SNSC).
- 2. Shall become effective upon review and acceptance by the (SNSC).
6.15 Offuite Dose Calculation Manual (ODCM) 6.15.1 The ODCM shall be approved by the Commission prior to implement 1cien.
Amendment No. 6-21
. '. e 6.15.2 Licensee initiated changes to the ODCM:
- 1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain
- a. Sufficiently detailed information to totally support the. rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluation justifying the change (s);
- b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
- c. Documentation cf the fact the change has been revised and found acceptable by the (SNSC).
- 2. Shall become effective upon ~ review and acceptance by the (SNSC).
6.16 Major Changes to Radioactive Liquid, Gaseous and Solid Waste 6.16.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change was made. The discussion of each change shall contains
- a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50 59.
- b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
- c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;
- d. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; Amendment No. 6-22
, '+ .
- e. An evaluation of the change, which shows the expected maximum exposures to individual in the Unrestricted Area and to the general population that differ from those previously estimated in the license application and amendments thereto;
- f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period to when the changes are to be mades
- g. An estimate of the exposure to plant operating personnel as a result of the changer and
- h. Documentation of the fact that the change was reviewed and found acceptable by the (SNSC).
Amendment No. 6-23
. ** e ATTACIDENT C APPLICATION FOR AMENDMENT TO OPERATING LICENSE Safety Assessment 1
4 Consolidated Edison Company of New York, Inc.
Indian Point Unit Nos.1 and 2 Docket Nos. 50-3 and 50-247 June, 1985
. t. o Safety Assessment A number of proposed technical specification revisions are set forth in Attachments A and B to this application. Several changes are being proposed at the request of the United Statea Nuclear Regulatory Commission (NRC) while other proposed changes are initiated by, Consolidated Edison. The apecific technical apecification are delineated and diacussed below:
a) The proposed reviaions to Figure 3-1 of the Indian Point Unit No. 1 Technical Specifications and Figure 6.2.1 of the Indian Point Unit No. 2 Technical Specifications would incorporate certain adminiatrative changes to the Facility Organization.
The proposed revisions to figures 3.1 and 6.2.1 reflect a detailed representation of the offsite organization indicating the organizational structure and lines of responsibility for the offsite groups that provide technical and management support for the unit. The organizational arrangement for performance and monitoring Quality Assurance activities is also indicated in detail.
The proposed revisions also reflect a recent Consolidated Edison organizational change. This change permits the Vice President of Nuclear Engineering Quality Assurance and Reliability to report directly to the Executive Vice President of Central Operations. The change provided more direct communication between the offaite organization and the unit, thereby facilitating technical and management support for the unit.
b) The propoaed change to Jpecification 6.2.2g waJ requested by NRC letter, dated February 22, l985, from Mr. Steven A.
Varga to Mr. John D. O'Toole. The purpoae of thia change la to limit overtime for critical shift job positions, which ia a proviaion af NUREG-0737, " Clarification of TMI Action Plan Requirements" Item I.A.l.3 and is also the subject of Commission Policy as Jtated in Generic Letter 82-12 dated June 15, 1982. The objective of the change la to prevent situations where fatigue could reduce the ability of operating personnel to keep the reactor in a safe condition. This change reflects administrative controls already in place to assure that, to the extent practicable, personnel are not ddJigned to shift duties while in a fatigued condition that could significantly reduce their mental alternesa or their decision making ability. Generic Letter 82-12 allows deviation from the guidelinea for very unusual circumstances.
C-1
I 4 4 o I
c) The propoaed revisions to specification 6 5.1.2 changes the composition of the Station Nuclear Safety Committee The memberahip of SNSC would be altered to (SNSC).
conform to the memberahip identified in the Standard Technical Specifications for Westinghouae Pressurized Water Reactors, NUREG-0452, Revision 4. The memberahip la supplemented by the Chief Technical Engineer and Radiatimi Protection Manager and removea thoaa membera who are only required for specific safety related matters. Theae individuals who are to be removed aa membera can be appointed by the SNSC Chairman to aerve as SNSC members aJ per apecification 6.5.1.2.1, when specific matters warrant their technical expertise. The change eliminates the position of Vice Chairman but provides the Chairman the authority to designate an alternate to serve as Chairman in his absence. The change clarifies the definition of the Quorum by being specific to the number of members required.
The proposed change to apecification 6.8 3 enablea the SNSC to document and review temporary changes within 14 days of implementation in order to allow adequata time for review.
These changes when implemented will increaae the effectiveneas of SNSC since it will be compoaed of thoJe individuals most related to mattera of nuclear Jafety and will improve the communication necedaary at SNSC meetings.
d) The propoaed change to apecifications 6 5.2 8F and 6.5.2.8g was requested by NRC Generic Letter 82-17, dated October 1, 1982 and NRC Generic Letter 82-23, dated October 30, 1982.
The purpose of the change to speci!! cation 6 5 2 8F is to conform to the requirements of Section 50.54(t) of Title 20 of the Code of Federal Regu'ationJ (10CFR50.54(t)). 10CFR 50 54(t) requires that each nuclear power reactor licanaee provide for an indeponlaat r w l .a of its lawfio/
Preparedneed Program at least every 12 tsont'13 The purpose of the change to apecification 6 5 2 8g is to conform to the requirements of Section 73 40(d) of Title 10 of the Code of Federal Regulations (10CFR73 40(d)).
10CFR73.40(d) requires that each nuclear power reactor licensee provide for an independent review of ita Safeguards Contingency Plan at least every 12 monthJe e) The proposed change to Jpecifications 6 9 1.7 was requeated by NRC letter dated March 20, 1985, from Mr. Steven A.
Varga to Mr. John D. O'Toole. The purpose of thia change C-2 i
. 6 e e
is to report relief and safety valve challenges which is a provision of NUREG-0737, " Clarification of TMI Action Plan Requirements" Item II.K.3.3. The deletion of specification 6 9 1.8 is proposed to conform specification 6 9.1.7 to the Standard Technical Specifications refered to above in item c.
f) The proposed change to Specification 6 10 2 1 would clarify the record retention requirements. The purpose of the change is to exempt records of Quality Assurance activities under specification 6.10.1, from the record retention requirement of 6.10.2. This clairification in the specification will reduce the unnecessary record retention burden and alleviata any inspection discrepancies that could be encountered with previously conflicting requirements.
Basis for no significant hazards consideration determination The proposed changes when implemented would reflect (a) organization change (b) overtime limits for critical job positions (c) Station Nuclear Safety Cormittee (SNSC) membership changes (d) more frequent auditing of the Emergency Plan and Security Plan (e) reporting requirer.ent for relief and Pressurizer safety valve challenges and (f) record retention clarification. The organization changes are administrative in nature and will not reduce the effectiveness of the facility organization nor would the changes decrease the required qualification of personnel. The overtime limits for critical positions constitutes an additional limitation and control not presently included in the technical specification but, implemented for some time thru administrative controls. The changes to the SNSC membership are mainly administrative in nature and will not reduce the effectiveness of the committee nor would the changes decrease the qualifications of the members. The change in frequency of the mnergency and Security Plan audits is to conform to the regulations of 10 CFR and is conservative. The reportino of relief and safety valve challenges constitutes an additional limitation and restriction not presently included in the technical specifications and conforms the specification to the Standard Technical Specification. The clarification of record retention requiremenus is purely administrative in nature and achieves consistency in the technical specifications.
Therefore, we believe the requested action involves no significant hazards consideration, because it does not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident reduction in a margin of safety.
C-3
7 ,,.
. #' s j ,
t s'
The proposed changes have been reviewed by the Station Nuclear Safety Committee and the Consoli64ted Edison Nuclear Facilities safety C.maittee. ' Both coeuittee's concur that these changes do not represent a significant/ hazards consideration and will not cause any change in,the types .or increase in the amounts of effluents or any . change in the authorized power level of the facility.
f a
S
[ ,
r .
f5
(
f s,Y J
9 f
f
/
2 C-4
.