ML20070N695
| ML20070N695 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 04/26/1994 |
| From: | ILLINOIS POWER CO. |
| To: | |
| Shared Package | |
| ML20070N693 | List: |
| References | |
| RTR-NUREG-1434 NUDOCS 9405090092 | |
| Download: ML20070N695 (100) | |
Text
f U-602203 Att. 2 Page 3 of 103
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ATTACHMENT 1A l
CTS -
PSTS COMPARISON DOCUMENT l MARKUP OF CTS 9405090092 940426 PDR P ADOCK 05000461 PDR e ... -. - - . . . . _ . .
U-602283 Att. 2 Page_ #4 of 183 re 2 t 196
_ Sec1[on e % 56 I 6.0 ACHINISTRATIVE CONTROLS
.5./ $ 5-2 5I 6.1 RESPONSIBILITY A\
- 6. I ! 6.1.1 The kanacer - C13ntnn Pavar 9tlit Q shall be responsible for overall unit operation and shall delegate in writing the succession to this responsi-bility during his absence.
6,l.1 6.1.2 The Shift Supervisor or during.his absence from the control room, a y designated individuai shall be responsible for the enntrol room Commad-- > <'
. functionJ
/HucTear] stra/il bdmanAgementy reiss6ed to/alldirec).ive tojthis effpct, st(tion per4onnel od ansigp'ed annu61by basis.
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- 6. 2 6.2 ORGANIZATION 5 2.I 6.2.1 0FFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.
6,7,ld a. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, 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 CPS Updated Safety Analysis Report (USAR).
6, '2,j ./2 b. The .anager - C14a+nn Pnwor Statio3shall be responsible for overall
- l unit safe operation and shall have control over those onsite ,
Ak activities necessary for safe operation and maintenance of the plant.
6,2,/.C c. The N D-aUrfant - hele2shall have corporate responsibility for ,
overall plant nuclear safety and shall take any measures needed to I ensure acceptable performance of the staf f in operating, I maintaining, and oviding technic _ql.jwpngrt to the nlant tun l'g f
nuc r safetv. , Sj w nc d cor poM exe cu-hue. CM JAL d o w ~ uh s~ rk. US All / _ . s l
5 2,I,d d . 1ne ineiyiduais uno train tne operating staff and tnose who carry out radiation protection and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.
t CLINTON w I 6-1 Amendment No.'34
U-602283 Act, 2 Page 5 or 183 INSERT 1A .
The plant manager, or his designee, shall approve, prior to implementation, each proposed test, experiment, and modification to systems or equipment that affect nuclear safety, s
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! INSERT CLINTON 6-1 4/15/94
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U-602203 Att. 2 Page 6 of 103 Tedos a to E d 196 q ec,tionSPage S o %S 6.0 ADMINISTRATIVE CONTROLS ~
h , 2. 7. 6.2.2 UNIT STAFF ,
Each on duty shift shall be composed of at least the minimum shift crew t composition shown in Table 6.2.2-1; g,2,L 6 b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in OPERATIONAL CONDITION 1, 2 or 3, at least one licensed Senior Operator shall be in the control room; 6,L 1b c. A Radiation Protection Technician
- shall be on site when fuel is in the reactor;
- 6. All CORE ALTERATIONS shall be observed and directly supervised by either
a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this p0 ooeration:-
- e. A ite fire bKgade of at1 east five members-shall be maintained on site ]
at all times *. The fire brigade shall not include the Shift Supervisor, j the Shift Technical Advisor, nor the two other members of the minimum i shift crew necessary for safe shutdown of the unit and any personnel re uired for other essential functions during a fire emergency; and
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g,2,2,d *The Radiation Protection Technician and hire brica3compositionmaybeless than the minimum requirements for a peri 6d 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 ,
l order to accommodate unexpected absence, provided immediate action is taken l l
to fill the required positions. -
CLINTON 6-la Amendment No. 34
U-602283 Att. 2-Page 7 of 103 _ INSERT laA
- a. A non-licensed operator shall be on site when fuel is in the reactor and an additional non-licensed operator shall be on site while the unit is in MODE.1, 2, or 3.
l 1 I i 1 l l l l INSERT
< CLINTON 6-la 4/15/94
u-002283 Att. 2 Page 8 of 183 f.@ 55 4 ADMINISTRATIVE CONTROLS i UNIT STAFF (Continued) - h.7, M . f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions ~ e.g. , licensed Senior Operators, licensed Operators, health physicists, auxiliary operators, and key maintenance personnel. 4
&eamount Th ertime worked by unit sta .
embers performing safety-
'related ions shall be limited cordance with the NRC Policy g8 (Stateer t on workin hours (Gene c letter No. 82-12).;
Adequate shift coverfge shall be maintained without routine heavy use Y
\of overtime._Ifha ^uan ma +'"t ;,sw oper;;;r.g perwane+-worA-aj (no-1 M. cur day, 40 hour week M.L use unii. is uvui a ting.Pfowever, in the event that unforeseen problems require substantial , amounts, of . i overtime to be used, or during extended peri.ods of s'hutdown for refueling, I major maintenance, or major unit modifications, on a temporary basis the i following guidelines shall be followed:
- 1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time.
- 2. An individual should not be permitted to work more than 16 hours in any 24-hour tieriod, nor more than 24 hours in any 48-hour period, ;
nor. more than 72 hours in any seven day period, all excluding shift u turnover time. I l
- 3. A break of at least eight hours should be allowed between work periods, including shift turnover time.
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- 4. Except during extended shutdown periods, the use of overtime should i
be considered on an individual basis and not for the entire staff on l a shift.
)
M Any__ deviation from the above guidelines shall be authorized by the anacer-c a nt an onwr StatioiDor his deputy, or higher levels of -l management, in accordance with established procedures and with documenta-l tion of the basis for granting the deviation. ' Controls shall be included ' in the rocedure's such that individual overtime shall be reviewed monthly by e ananar-ointon Power Station)or his designee to assure that l excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.
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! i I e CLINTON O Ii4T M 6-2 Amendment No. 26
' k U-so2283 Att. 2 Tnelo 2 to i . 6 Page 9 of 183 Secderr$ Pagu l
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N 5) L f A % 1, 1 4 CLINTON (ii;iIT 1 6-3 Amendment No. 34
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3 O be t 4 l l 4 1 I t e 1 CLINT0H h IT M 6-4 Amendment flo. 34 i
F u-602283 Att. 2 Page il of 183 , TABLE 6.2.2-1 l- MINIMUM SHIFT CREW COMPOSITION SINGLE UNIT FACILITY l NUMBER OF INDIVIDUALS REOUIRED TO FILL POSITION l POSITION CONDITION 1, 2, or 3 CONDITION 4 or 5 SS 1 1 SRO 1 None I R0 2 1 6.2i2 A. AD. 2 1~ STA 1 None] \ TABLE NOTATIONS l
- -- SS - Shif t Supervisor with a Senior Operator license on Unit 1.
SRP- Individual with a Senior Operator license on Unit 1. I R0 Individual with a Operator license on Unit 1. AO - Auxiliary Operator Q - Shift Technical Advisor - l i The shift crew composition may be one less than the minimum requ.irements of Table 6.2.2-1 for a period of time not to exceed 2 hours in order to accom-modate unexpected absence of on-duty shift crew members provided immediate 5224 action is taken to restore tne snift crew composition to witnin the minimum requ,irements of Table 6.2.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. During any absence of the Shift Supervisor from the control room while the unit i is in OPERATIONAL CONDITION 1, 2, or 3, an individual with a valid Senior Opera-5'11 t r icense sh 1 be designated t assume the control room command function. During any absence of the Shift Supervisor from the control room while the unit is in OPERATIONAL CONDITION 4 or 5, an individual with a valid Senior Operator license or Operator license shall. be designated to assume the control room command function. w t CLINTON - UNIT 1 6-5
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U-502263 Att. 2 Page 12 09 163 I ADMINISTRATIVE CONTROLS I 6.2.3 INDEPENDENT SAFETY EAGINEERING GROUP (ISEG1 FUNCTION 6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of unit design and operating experience information, including units of i similar design, which may indicate areas for improving unit safety. The ISEG l shall make detailed recommendations for revised procedures, equipment l modifications, maintenance activities, operations activities, or other means l of improving unit safety to. the Manager-Nuclear Assessment, and to members of - l ) the Nuclear Review and Audit Group (NRAG). ' COMPOSITION l p b 6.2.3.2 The ISEG shall be composed of at least five, dedicated, fulb time l 9 members located on site. Each ISEG member shall have either: (1) A bachelor's degree in engineering or related science and at least l three years of professional-level experience in his field which , shall include at least one year of nuclear power experience, or l (2) At least five years of nuclear experience and hold or have held a Clinton Power Station Senior Reactor Operator license.
/ As a minimum, four of the I'SEG members shall have the qualifications specified l
[ in (1) above. I RESPONSIBillTIES l 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. l RECORDS 6.2.3.4 Record!. of activities performed by the ISEG shall be prepared, maintained, and ' forwarded each calendar month to the Manager-Nuclear ! Assessment and to members of the NRAG. l l esponsitde for signoff function. CLINTON - UNIT 1 6-6 Amendment No. 86 i
.i f U-602203 Att. 2 Page 13 or 183 ADMINISTRATIVE C0f1TROLS 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support , g'g, to the Shift Supervisor in the areas of thermal hydraulics, reactor l engineering, and plant analysis with regard to ufo nnnrntion nf the unit. f The Shift Technical Advisor shall have a bachelor's degree or equivalent in a] scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in I,{ unit design and layout, including the capabilities of instrumentation and d;ontrols in the control room. . l
; 6.3 UNIT STAFF OVAllFICAT!'ONS 1
6.3.1 Each member of the unit staff shall meet or exceed minimum _ i 5,3,\ aualifications of ANSI /ANS 3.1-1978. [Tha airec4w Plant Ops,du uu> shall - I have-held--ar SRO lin nae fei- Clinten Pcwcr Stationt the Assistant Director - , grg g,Q {glant Operations shall hold an SR0 license for Clinton Power Sta F 6.4 TRAINING} 1
~ 6.4.1 A retraining and replacement training program for the unit staff shall
, lM be maintained under the direction of the Director - Operations Training and l shall meet or exceed the requirements of 10 CFR Part 55. REVIEW AND AUDIT j { 6.5
. 6.5.1 FACILITY REVIEW GROUP (FRG P
. 6.5.1.1 The FRG shall function to advise the Manager - Clinton Power Station on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The FRG shall be composed of a chairman and supervisory (supervisor ! or higher) members collectively having the experience and competence in areas j l 1 . of plant operations, maintenance (control and instrumentation, electrical maintenance, mechanical maintenance), radiation protection, chemistry and (nuclear engineering, necessary to perform the required review function. CLINTON - UNIT 1 6-- Amendment No. 86
U-502203 Att. 2 Page 14 of 183 ADMINISTRATIVE CONTROLS g ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the FRG Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in FRG activities at any one time. MEETING FREQUENCY 4 6.5.1.4 The FRG shall meet at least once per calendar month and as convened by the FRG Chairman or his designated alternate.
- V QUORUM i
6.5.1.5 The quorum of the FRG necessary for the performance of the FRG
, responsibility and authority provisions of these Technical Specifications shall. consist of the Chairman or his designated alternate and four members including alternates.
1 . RESPONSIBILITIES 6.5.1.6 The FRG shall be responsible for: I
- a. Review of all' administrative procedures and changes thereto,
- b. Review of all proposed tests and experiments that affect nuclear safety;
- c. Review of all proposed changes to Appendix A Technical Specifications;
- d. Review of all proposed changes or modifications to unit systems or equip-ment that affect nuclear safety;
- e. Review of all proposed programs required by Technical Specification 6.8.4 t and changes thereto;
- f. Review of the safety evaluation for procedures, and changes thereto, re-quired by Technical Specification 6.5.3.1.c;
- g. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recom-mendations to prevent recurrence, to the Vice President Nuclear and to the Nuclear Review and Audit Group;
- h. Review of all REPORTABLE EVENTS; Review of unit operations to detect potential hazards to nuc kar safety; l
6
CLINTON - UNIT 1 '
6-8 - y
X U-002283 Att. 2 Page 15 of 183 ADMINISTRATIVE CONTROLS
] 'rRESPONSIBILITIES (Continued) -
- j. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Manager-Clinton Power Station or the Nuclear Review and Audit Group;
- k. Review of the Security Plan and implementing procedures and submittal of recommended changes to the Nuclear Review and Audit Group; and 1 !
l
- l. Review of the Emergency Plan and implementing procedures and submittal of l
{ recommended changes to the Nuclear Review and Audit Group. v m. Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations f and disposition of the corrective action to prevent recurrence, and the forwarding of the reports to the Vice President Nuclear and to the Nuclear Review and Audit Group.
- n. Review of the changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL.
l
- o. Review of the Fire Protection Program and implementing procedures and
- submittal of recommended changes to the Nuclear Review and Audit Group.
l DUTIES ) k l l 6.5.1.7 The FRG shall: ;
- a. Recommend in writing to the Manager-Clinton Power Station approval or dis- i approval of items considered under Specification 6.5.1.6a through e prior '
to their implementation.
- b. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6a through g constitutes an unreviewed safety question.
- c. Provide written notification within 24 hours to the Vice President Nuclear and the Nuclear Review and Audit Group of disagreement between the FRG and the Manager-Clinton Power Station; however, the Manager-Clinton Power Station shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.
RECORDS 6.5.1.8 The FRG shall maintain written minutes of each FRG meeting that, at' a minimum, document the results of all FRG activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Vice President Nuclear and the Nuclear Review and Audit Group. i CLINTON - UNIT 1 6-9
U-002203 Att. 2 Page 16 or 183 ADMINISTRATIVE CONTROLS k.5.2NUCLEARREVIEWANDAUDITGROUP(NRAG) l FUNCTION 6.5.2.1 The NRAG shall function to provide independent review and audit of designated activities in the areas of:
- a. Nuclear power plant operations,
- b. Nuclear engineering,
- c. Chemistry and radiochemistry, ,
- d. Metallurgy, !
- e. Instrumentation and control,
- f. Radiological safety, 9 Mechanical and electrical engineering,
.. h. Quality assurance practices,
- i. Non-destructive testing,
- j. Administrative controls, and -
- k. Other appropriate fields associated with the unique characteristics of Clinton Unit 1.
t COMPOSITION 6.5.2.2 The NRAG, as a minimum, shall be composed of a Director and four mem-bers. Individuals filling positions as required by ANSI /ANS 3.1-1978, Section 4.7.2, shall meet the education and experience requirements of the same
) section. The NRAG Director shall be appointed by the Vice President. The members shall be appointed by the Director. Current membership rosters, signed by the Director, shall be maintained.
ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NRAG Director to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NRAG activities at any one time. CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NRAG Director to provide expert advice to the NRAG. MEETING FRE0VENCY 6.5.2.5 The NRAG shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per l 6 months thereafter. j CLINTON - UNIT 1 6-10 Amendment No. 86
r ' U-602283 Att. 2 Page 17 or 183 . O ADMINISTRATIVE CONTROLS QUORUM 6.5.2.6 The quorum of the NRAG necessary for the performance of the NRAG review and audit functions of these Technical Specifications shall consist of the Director or his designated alternate and at least four NRAG members including alternates. No 'more than a minority of the quorum shall have line responsibility for operation of the unit. REVIEW 6.5.2.7 The NRAG shall be responsible for the review of:
- a. The safety evaluations for (1) changes to procedures, equipment or system,
-- and (2) tests or experiments completed under the provision of 10 CFR 50.59 to ver'.fy that such actions 'did not constitute an unreviewed safety question;
- b. Fraposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defi.ned in 10 CFR 50.59;
- c. Proposed tests or experiments which involve an unreviewed safety question.
as defined in 10 CFR 50.59; j
- d. Proposed changes to Technical Specifications or this Operating License; j e. Violations of codes, regulations, orders, Technical Specifications, )
license requirements, or of internal procedures or instructions having nuclear safety significance;
- f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety;
- g. All REPORTABLE EVENT 5;
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and I, Reports and meeting minutes of the FRG and ISEG.
AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the NRAG. These audits shall encompass: i
- a. The conformance of unit operation to provisions contained within the Tech-nical Specifications and applicable license conditions at least once per j 12 months; CLINTON - Un!T 1. 6-11 b -
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U-002283 Att. 2 Page 18 of 183 ( ( ADMINISTRATIVE CONTROLS [ AUDITS (Continued)
- 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 Assur-ance Program to meet the criteria of Appendix B,10 CFR Part 50, at least once per 24 months;
- e. The Emergency Plan and implementing procedures at least once per 12 months.
& -- f. The Security Plan and imp,lementing procedures at least once per 12 months.
- g. Any other area of unit operation considered appropriate by the NRAG or the
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Vice President.
- h. The fire protection programmatic controls including the implementing procc- 1
( dures at least once per 24 months by qualified licensee QA personnel; d' i. The fire protection equipment and program implementation shall be performed
\' at least once per 12 months utilizing either qualified offsite licensee fire protection engineer (s) or an outside independent fire protection con-sul tant.
J. An inspection and audit of the fire protection and loss prevention program shall be performed by an outside' qualified fire consultant at intervals . no greater than 36 months. 1
- k. The radiological environmental monitoring program and the results thereof at least ance per 12 months. -
- 1. The OFFS?.TE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.
l m .- The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months. ,
- n. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977, at least once per 12 months. -
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a i m mvm, _ mir , 6-12
U-002283 Att. 2 Page 19 or te3 ADMINISTRATIVE CONTROLS 's IkUTHORITY 6.'5. 2. 9 The NRAG shall report to and advise the Vice President on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8. RECORDS l l 6.5.2.10 Records of HRAG activities shall be prepared, approved, and distrib- i uted as indicated below:
- a. Minutes of each NRAG meeting shall be prepared, approved, and forwarded to the Vice President within 14 days following each meeting.
- b. Reports of reviews encompassed by Specification 6.5.2.7 shall be prepared, approved, and forwarded to the Vice President within 14 days following completion of the review.
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- c. Audit reports encompassed by Specification 6.5.2.8 shall be forwarded to h the Vice President and to the management positions responsible for the l
\/ areas audited within 30 days after completion of the audit by the auditing organization.
6.5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES Procedures required by Technical Specificction 6.8 and other procedures which affect plant nuclear safety as determined by the Manager - Clinton Power Station or the responsible mar.ager* and changes thereto, other than editorial or typo- l graphical changes, shall be reviewed as follows: 6.5.3.1 TECHNICAL REVIEW -
- a. Each such procedure 'or procedure change shall be independently reviewed by an individual knowledgeable in the area affected other than the individ- )
ual who prepared the procedure, or procedure change. The Manager - Clinton Power Station or the responsible manager
- shall prior to implementation [
approve all plant procedures and changes thereto. )
- b. Individuals responsible for reviews performed in accordance with Item 6.5.3.la. above shall be designated by the 'Hanager - Clinton Power i Station or the responsible manager *. Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary. If deemed necessary, such review shall be performed by the l review personnel of the appropriate discipline. ,
Individuals performing these rev~iews shall meet or exceed the qualifi-j cations stated in ANSI /ANS 3.1-1978 for the' appropriate discipline. f . . .
*The responsible manager must be the equivalent of the Manager - Clinton Power j Station with respect to his/her level of responsibility within the corporate
( structure. i l f CLINTON - UNIT 1 6-13 Amendment No. 34 l
u-002283 Att. 2 Page 20 or jg3 ADMINISTRATIVE CONTROLS r (TECHNICAL REVIEW (Continued)
- c. When required by 10 CFR 50.59, a saf ety evaluation to determine whether n the or not an unreviewed safety question is involved shall be included i,.
procedure or the procedure change review. If it is determined that an unreviewed safety question is not involved, a written safety evaluation j to support that decision will be prepared and submitted to the FRG for review. Pursuant to 10 CFR 50.59, NRC approval of items involving unre-viewed safety questions shall be obtained prior to the Manager-Clinton Power Station approval for implementation.
- d. Written records of reviews performed in accordance with item 6.5.3.1.a.
above, including recommendations for approval or disapproval, shall be ). prepared and maintained. k6 REPORTABLEEVENTACTION1 6.6.1 The following actions shall be taken for. REPORTABLE EVENTS: h a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and (b. Each REPORTABLE EVENT shall be reviewed by the FRG, and submitted to L.M the NRAG and the Vice President. ( I 6.7 SAFETY LIMIT VIOLATION \ 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
- a. In accordance with 10 CFR 50.72, the NRC Operations Center shall be noti-fied by telephone as soon as possible and in all cases within 1 hour af ter the violation has been determined. The Vice President and the NRAG shall be notified within 24 hours.
- b. A License Event Report shall be prepared in accordance with 10 CFR 50.73.
]b c. A Safety Limit Violation Report shall be prepared. The report shall be re-ggM viewed by the FRG. This report shall describe (1) applicable circumstances TD preceding the violation, (2) effects of the viclation upon unit components, systems, or structures, and (3) corrective action taxen to prevent recurrence.
@p The Safety Limit Violation Report shall be submitted to the Commission d.
within 30 days and to the NRAG, and the Vice Presirent within 30 days 'of the violation.
- e. Critical operation of the unit shall not be resumed until authorized by ..
the Commission. 6-14 Amendment No. 26 CLINTON - UNIT 1
U+602203 Att. 2 Page 21 of 183 m ;
. ADMINISTRATIVE CONTROLS
( 5.g 6.8 PROCEDURES AND PROGRAMS PROCEDURES 6 g y,l co.8.1 veringWritten the procedures activities referenced shall be established, below: implemented, and maintained
- a. The applicable procedures recommended in Appendix A of Regulatory
$N'A Guide 1.33, Revision 2, February 1978.
g'g b. The M p ocedures required to implement the re uirements of NUREG-0737 and supplement, ereto. N d m W W p d L 2- l F. c Refueling operations k l M d Surveillance and test activities of safetv-related equipment) L.
, e. Security Plan implementation. / l N f. Emergency Plan implementation.
g, /))/ r fo M 5 *: cM 1 (g. rire Protection Proaram implementation.] i r' peciSc A N.
. PROCESSCONTROLPROGRAMimplementation]
- i. OFFSITE DOSE CALCULATION MANUAL implementatioii) g,),c, j. Quality Assurance Program for effluent and environmental monitoring.
( REVIEW AND APPROVAL 1 k 6.8.2 Each procedure of-5)ecification 6.8.1, and changes thereto, shall be D reviewed in accordance witi 6.5.1.6 and 6.5.3 as approariate and shall be approved by the Manager - Clinton Power Station or the responsi)le manager
- prior to l implementation and reviewed periodically as set forth in administrative procedures.
TEMPORARY CHANGES 6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made provided:
- a. The intent of the original procedure is not altered;
- b. The change is approved by two members of the unit management staff, at least one of whom holds a Senior Operator license on the unit affected; and
- c. The change is documented, reviewed in accordance with 6.5.1.6 and 6.5.3 as appropriate, and app
- roved by the Manager - Clinton Power Station or the responsible manager within 14 days of implementation.
*The responsible manager must be the equivalent of the Manager - Clinton Power !
Station with respect to their level of responsibility within the corporate ( structure. ( CLINTON - UNIT 1 6-15 Amendment No.19
I u-602283 Att. 2
, Page 22 or 183 ~
ADHIHISTRATIVE CONTROLS PROGRAHS
$b 6.8.4 The following programs shall be established, implemented, and maintained:
- a. Primary Coolant Sources Outside Containment 4
A program to reduce leakage from those portions of systems outside containment that* could contain highly radioactive fluids during a serious transient or acci-dent to as low as practical levels. The systems. include the LPCS, HPCS, RHR, RCIC, Suppression Pool Hakeup,, Combustible Gas Control, Containment Monitoring and Post-Accident Sampling. The program shail include the following:
- 1. Preventive maintenance and periodic visual inspection requirements, and l
- 2. Integrated leak test requirements for each system at refueling cycle intervals or less, l
- b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the
- q. airborne iodine concentration in vital areas under accident conditions.
This program shall include the following: Training of personnel, 1.
- 2. Procedures for monitoring, and
- 3. Provisions for maintenance of sampling and analysis equipment._
g c. Postaccident Samoling
- A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples, under accident conditions. The program shall include the following:
- 1. Training of personnel,
- 2. Procedures for sample and analysis, and
- 3. Provisions for maintenance of sampling and analysis equipment.
Ultimate Heat Sink Erosion, Sediment Honitoring and Oredging Procram
} { d.
A program to provide maintenance on the ultimate heat sink (UHS) in the event inspections of the UHS dam, its abutments or the UHS shoreline indicates erosion or local instability. This program shall ensure that the UHS is maintained in such a way to achieve the following objectives: (1) During normal operation, there will be a volume of water in the UHS below elevation 675 sufficient to receive the sediment load from a once in 25 year flood event, and (2) Still be adequate to maintain the plant in a safe-shutdown condition for 30 days under meteorological conditions of the severity suggested by Regulatory Guide 1.27.
- e. Fire Protection Program A program to implement and maintain in ef fect all provisions of the approved fire protection program as described in the finil Safety Analysis Report as amended, and a~s approved in the Saf ety Evaluation Report (HUREG-0853) dated February 1982 as supplemented. Honcompliance with the above Fire Protection Systems described in plant procedure CPS Ho. 1893.01 shall be reported as a REPORTA8LE EVENT in accordance with Section 6.6.1 of these Technical Specifications.
CLINTON - UNIT 1 6-16 p)2 }MSCP:r \% }
u-no2283 Att. 2 'Lhol ure 2 to ' 96
- chit Pege 23 or 183 Sectio 56 INSERT 16A (Add new programs per NUREG-1434 markup.)
5.5.T 4.".10- Safety Function Determination Program (SFDP)
-0.',17 Technical specifications (TS) Bases control
$ $.lO e
as INSERT CLINTON 6-16 10/1/93
U-oo2283 Atc. 2 % ge2t 19M Page 24 of Ig3 - ' f }56 {Secti ADMINISTRATIVE CONTROLS PROGRAMS 1 gtpf. Radioactive Effluent Controls Proaram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEM8ERS OF THE l PUBLIC from radioactive effluents as low as reasonably achievable. The ; program (1) shall be contained in the 00CM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- 1. Limitations on the operability of radioactive liquid and gaseous ,
monitoring instrumentation including surveillance tests and setpoint I determination in accordance with the methodology in the 00CM;
- 2. Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to ten times the concentration values in 10 CFR Part 20.1001 - 20.2401, Appendix B, Table 2, Column 2;
- 3. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology l and parameters in the 00CM;
- 4. Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from the unit to UNRESTRICTED AREAS conforming to Appendix'I to 10 CFR Part 50;
- 5. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current.
calendar year in accordance with the methodology and parameters in ; the 00CM at least every 31 days; -
- 6. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix ! to 10 CFR Part 50; l
- 7. Limitations on the dose rate resulting from radioactive material I released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be limited to the following.:
1
- a. For noble gases: Less than or equal to 500 mrem /yr to the total i body and less than or equal to 3000 mrem /yr to the skin, and ,
- b. For Iodine-131, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ; CLINTON h 6-16a Amendment No. 69
, . . _ . . . - . _ _ . ..--..-....._.m U-002283 Att. 2 Page 20 of 183 1
1 9 OESIGN FEATURES , es j 5.5 METEOROLOGICAL TOWER LOCATION - I , 5.5.1 The meteorological tower shall be 0 .own on Figure 5.1.1-1. 5.6 FUEL STORAGE ! CRITICALITY I , 1 j 5. 6.1 The spent fuel storage r /f9 sned and shall be maintained with:
- a. A k,ff equivalent to le 'l sal to 0.95 when flooded with unborated water, including all ; uncertainties and biases as described in Section 9.1.2 of tb { l 4 b. A nominal 6.4375 .o-center distance between fuel assemblies
- placed in the < in the Fuel Building storage pool. A nominal,
, ce nter-to-cer . tween rows of 12.25 inches and within the rows of 7.00 inc* .ssemblies placed in the storage rack in the Upper i Containmer j 5. 6.1.1 The fuel for the first core loading stored dry in the
) spent fue' s shall not exceed 0.98 when aqueous foam moderation is .
4 assumed. ? i - ORAIN' S f 5. { . fuel storage pool is designed and shall be maintained to prevent aining of the pool below elevation 754'0". j k . i 5 i. ne spent fuel storage pool is designed and shall be maintained with a stora. a capacity limited to no more than 2522 fuel assembif es. j - I 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The' components identified in b h are %=M _g; A-shab'M 4 maintained within the cyclic or transient limits 4TTiuiu 3.7"Fp Mb i P 1 h . A e i
} }
j , .... 5-b i 1H5E 2 / 4 CL]NTON 0-6 i G - / C600 i
u-602203 Att. 2 Page 25 of 183
, ec 6 ADMINISTRATIVE CONTROLS PROGRAMS z g '_ Radioactive Effluent Controls Program (Continued)
(tal 8. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE B00H0ARY conforming to Appendix I to 10 CFR Part 50; 9. Limitations on the annual and quarterly doses to a MEM8ER OF THE PUBLIC from I6 dine-131, Iodine-133, tritium, and all radionuclides l in particulate form with hal(-lives greater than 8 days in gaseous effluents released from the unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50; and 10. 1, imitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190. h Radiological Environmental Monitoring Program A program shall be provided to monitor the radiati'on and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. x be contained in the OOCH, (2) conform to the guidance of Appendix I toThe progr gl 10 CFR Part 50, and (3) include the following: 1. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCH; 2. A Land Use Census to ensure that. changes in the use of areas at and l beyond the SITE BOUNDARY are identified and that modifications to i the monitoring program are mad ~e if required by the results of this census; and i 3. Participation in an Interlaboratory Comparison Program to i:nsure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are I performed onitoring. as part of the quality assurance program for environmental g' l5osq& l% N!S% W l a CLINT0thtti 6-16b Amendment No.40 1
- U-602283 Att 2 Ens - re 2 to 4 &219(
Page 27 or Sectio gqof gy l ~~s )
- l l
;- TABLE 5.7.1'-1 s l
COMPONENT CYCLIC OR TRANSIEN L1 HITS CYCLIC OR DESIGH CYCLE , COMPONENT TRANSIENT LIMIT , OR TRANSIENT Reactor 120 heatup and cooldown cycles 70*F to 560*F t6,70*F 80 step change cycles Loss of feedsater heaters 180 reactor trip cycles 100% to 0% of RATED THERMAL POWER 40 hydrostatic' pressure or Pressurized to leak tests 930 psig and 1250 psig h 9 o y .
)
Se t l 1 1 I l
. l s
CLINTON( 5'- G - /66G) 1 1 1
U-co2283 Att. 2 APPLICABILITY 1 j SURVEILLANCE REQUIREMENTS l 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS ' , or other conditions specified for individual Limiting Conditions for Operation l unless otherwise stated in an individual Surveillance Requirement. ' 4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance. interval with a maximum allowable extension not to exceed 25 per-g cent of the specified surveillance interval. %h 4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a . Qs Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications. Surveillance requirements do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicabic condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified. J
- f. 3. 5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservic o testing of ASME .Co.de Class 1, 2, and 3 pumps and valves shall be performed - -
in accordance with Section XI of, the, ASME Boiler anii Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel ' Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel y -
Code and applicabic. Addenda shall be applicable as follows in these Technical Specifications: ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice inspect. ion.and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days / Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days J TW Amendment No. 53 CLINTON - UNIT 1 J/4-C 2 (o M
s. U 602283 Att, 2 Page 2g or pg3 APPLICABILITY SURVEILLANCE RE0VIREMENTS (Continued) k0.5 (Continued)
- c. The provisions of Specification 4.T.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.
- d. Performance of the above inservice inspection and testing activities shall
% 1 be in addition to 'other specified Surveillance Requirements.
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
- f. The Inservice Inspection Program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC Staff positions on schedule, methods and personnel, and sample expansion included in the generic letter.
DJ6ERT' , CLINTON - UNIT 1 -2/', 0 3-- Amendment No. 65 l b'/kb[y) l i
.._...__-._._-,,..~_.._..-____..._..-_--j
U-602283 Att. 2 Page 30 of 103 2 -6021 CONTAINMENT SYSTEMS ( Sejc oerStage 4 15:
$3NDBY CAS TREATMENT SYSTEM
,g '
LIMITING CONDITION FOR OPERATION 3.6.6.3 Two independent standby gas treatment sub i be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 ACTION:
- a. With one standby gas treatment suk inoperable subsystem to OPERABLr .
h .: able, restore the 7 days, or:
- 1. In OPERATIONAL CONDITIF oe in at least HOT SHUTDOWN within the next 12 hours and 0 ,0WN within the following 24 hours.
- 2. In OPERATIONAL C0'~ 0 . pend handling of irradiated fuel in the secondary c' dE ALTERATIONS and operations with a potential for .eactor vessel. The provisions of Specificatic at applicable, f
- b. With both str
, suspend
- ALTERATI0' C
[ . ment subsystems inoperable in OPERATIONAL CONDITION
. radiated fuel in the secondary containment, CORE The pro' q/ .ons with a potential for draining the reactor vessel.
.cification 3.0.3 are not applicable.
SURVEILLA' dTS 4.6.e .dby gas treatment subsystem shall be demonstrated OPERABLE:
- a. .
ti., once per 31 days by initiating, from the control room, flow through a filters and charcoal adsorbers and verifying that the subsystem oper. es for at least 10 hours with the heaters OPERABLE. b' b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or or chemical release in any ventilation zone commun(icating with the subsystem y gg J ,a Sechod 3/4 b
/
O/*When irradiated fuel is being handled in the secon'dary containment and during CORE (ALTERATIONS and operations with a potential for draining the reactor vessel. TNSER T CLINTON %,C5- Amendment No. 68 c-n de 4
,- - - - ~ ~ . . . . . . . . -- .- . _ _ . . . . . - - _ _ _ - _ _ - _ _ - _ _ _ . _
U-602203 Att. 2 Page 31 of 103 CONTAINHENT SYSTEMS STANDBY GAS TREATMENT SYSTEM $,$, Q SURVEILLANCE RE0VIREMENTS (Continued) l 4.6.6.3 (Continued)
- 1. Verifying that the subsystem satisfies the in-place penetration and g,g4,4 bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c and 5 $4 b C.S.d of Regulatory Guide 1.52, Revision 2 March 1978*, and the system flow rate is 4000 cfm 10%. g
- 2. Verifying,61 thin 31 days after removah that a laboratory analysis of a representative caroon sample obtaineo in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978*, meets gg ,g . c.,, the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978*, for a methyl iodide-penetr ion of less than 0.175%; when tested in accordance with ASTM h D380 methods,- with the following parameters: !
[a) Bed Depth - 4 inches 4AZ , ( b) Velocity - 40 fpm c) Temperature - 80
- C d) Relative Humidity - 70%
and l 5,5.(o. A. 5.5 M 3. verifying a subsystem flow rate of 4000 cfm 10x d ing system operation j g $ L.d when tested in accordance with ANSI N510-1980. ji _ ,
~
i c. (After every 720 hours of charcoal adsorber operation by verifying, l (days after removal) that a laboratory analysis of a representative aroon c(with
- sample oDt.ained in accordance with Regulatory Position C.6.b of Regulatory l g,f,(0.C Guide 1.52, Revision 2, March 1978*, meets the laboratory testing criteria of l Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978*, j
! for a thyl iodide penetration of less than 0.175%; in accordance with ASTM i h 0380 methods, with the following parameters: Bed Depth - 4 inches 2 Ca) b) Velocity - 40 fnm c) Temperature - 80* C d) Relative Humidity - 70% g.5,W6 MNSI N510-1980 shall be used in place of ANSI N510-1975 as referenced in Regulatory Guide 1.52, Revision 2, March 1978.
'C Wa @ T~
CLINTON - U"!T F -3/i S 29 Amendment No. 46, 68 (p #lhbf(ph
l l U-602283 Att. 2 Page 32 or 183 l CONTAINMENT SYSTEMS STANOBY GAS TREATMENT SYSTEM q b.b. b l SURVElllANCE RE0VIREMENTS (Continued) 4.6.6.3 (Continued) L.& 554 6 At least once per 18 months by: (1. Performing a system functional test which includes simulated auto-g j matic actuation of the system throughout its emergency operating sequence \ for the: 9 a) LOCA, and b) Fuel handling accident.
- 2. Verifying that the pressure drop across the combined HEPA filters 5.5.lo 8 and charcoal adsorber banks is less than 6.0 inches Water Gauge while operating the filter train at a flow rate of 4000 cfm 10%.
- 3. Verifying that the filter train starts and isolation dampers open on l ,
receipt of the following test signals:
;p6M a) Manual initiation from the control room, and l
Q b) Simulated automatic initiation signal.
- 4. Verifying that the filter cooling bypass dampers can be manually opened and the fan can be manually started.
5.6 h'g 5. Verifying that the heaters dissipate at least 18.0 kW when tested in accordance with ANSI N510-1980.
- e. (Efter each complete or oartial replacement of a HEPA filter banQ by verifying bg'4'g that the HEPA filter bank satisfies the in-place penetration ana bypass j leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at a flow rate of 4000 cfm i 10%.
difter each complete or partial replacement of a charcoal adsorber bank 3by f. verifying that the charcoal adsorber bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in
$,$.(c.b accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test l gas while operating the system at a flow rate of 4000 cfm 10%. l l
v nsd2T y papa & SR 3.02 anD SE3'03 g7 m oppnaDe to w vFTP 4e9 frepdd _p. > - TN6@ CLINTON -UNIT 4 2/1 C 10 Amendment No. 66, 68 l l 6-la(7)
U-607783 Att. 2 Pega 33 of 183 PLANT SYSTEMS CONTROL ROOM VENTILATION SYSTEM gg4 5 SURVEILLANCE REQUIREMENTS (Continued) 4.7.2 (Continued)
- c. St-leastonceper18monthsor(1)afteranystructuralmaintenanceonthe) makeup or recirculation HEPA filters or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone j ]
communicating with the makeup or recirculation filter system by:
- 1. Verifying that the makeup filter system satisfies the in place penetration and bypass leakage testing acceptance criteria of less 65.lo,q than 0.05% and uses the test procedure guidance in Regulatory Positions
d ' "' 9 " ' ' "" ' d ' * ' ' * "'" " ' " h 5' % b ' 1978*', '* '
and e "sy"s' th tem flow rate is 3000 cfm i 10%' .
- 2. Verifying that the recirculation filter system satisfies bypass leakage testing acceptance criteria of less than 2% total bypass and l I
uses test procedure guidance in Regulatory Positions C.S.a and C.S.d ' 5.5.b.b of Regulatory Guide 1.52, Revision 2, March 1978*, and the system flow rate is 64,000 cfm i 10%.
- 3. Verifying,(within 31 days after removal) that a laboratory analysis of a representative carDon sample obtained in accordance with Regula-tory Pos~ition C.6.6 of Regulatory Guide 1.52, Revision 2, March 1978*,
meets the laboratory testing criteria of Regulatory Position C.6.a of 6.$ .M Regulatory Guide 1.52, Revision 2, March 1978*, for a methyl iodide penetration of less than 0.175% for makeup filter system carbon ad-l sorber and 6% for recirculation filter ystem carbon adsorber when methods, with the following h tested; in accordance with ASTM 0380 parameters: Make Up Filter System a) Bed Depth - 4 inches 4N b) Velocity - 40 fpm c) Temperature - 30 % d) Relative Humidity - 70% Recirculation Filter Syste_m g2, a) Bed Depth - 2 inches
< b) Velocity -
80 fpm lemperature 30"C c) d) Relative Humidity - 70% Verifying flow rate of 3000 cfm i 10% for the makeup filter system 6,$,6,)4. and 64,000 cfm i 10% for the recirculation filter system during opera-g ,g.tp .b tion when tested in accordance with ANSI H510-1980. g ,$ .te d
- ANSI H510-1980 shall be used in place of ANSI H510-1975 as referenced in
($ @pio Regulatory
$ Guide 1.52, Revision 2, March 1978.
N' CLINTON 'J"!T 1 ' - S/4 7-i TsS05)(,- 14 b(4 I t-- - _ - _ _ _ _ __ - ~ . - . . - , _ , . . . _ _ , , _ . , ,_ _
. . U-602283 Att. 2 Page 314 of 183 PLAfU SYSTEMS i
; CONTROL ROOM VENTILATION SYSTEM { ], g SURVEILLANCE REQUIREHENTS (Continued) 4.7.2 (Continued)
- d. [After every 720 hours of charcoal adsorber coeration, by verifyino withi3 (31 days af ter removal f that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C;6.'b of Regulatory Guide 1.52, Revision 2, March.1978*, meets the laboratory test.-
$56.<- ing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978*, for a methyl iodide penetration of less than 0.175% (
l for the makeup filter system carbon adsorber and 6% for the recirculatio l filter system carbon adsorber when tested; in accordance with ASTH D380 49-methods, with the following parameters: l Make Up Filter System l
) Bed Depth - 4 inches-b) Velocity - 40 fpm 's c) temperature -
30"G-d) Relative Humidity - 70% - LA2 ,
)
/
Recirculation Filter System s , f,a) Bed Depth - 2 inches I (b) Velocity - 80 fom
- c) Temperature -
30 C > d) Relative Humidity - 70% i he. At least once per 18 months by
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while g,g'gd _ operating the makeup filter system at a flow rate of 3000 cfm 10%.
- b. Verifying that on a high cniorint: ci Luation'^ an6 manual initia-n tion test signal, the system automatically ** switches to the chlorine mode of operation and the dampers close within 2 seconds.***
gd) i ffo+ 3. Verifying that the control room leak rate is limited to < 4000 cfm 4 go i 10% at > 1/8-inch Water Gauge (W.G.) with respect to a'Bjacent areas.
- 4. Verifying that on a smoke mode actuation test signal, the system automatically switches to the smoke mode of operation at a flow rate
( less than or equal to 64,000 cfm 10%. Verifying that on a high radiation actuation test signal, the system automatically switches to the.high radiation node of operation and (5. - N ^ ANSI H510-1980 shall be used in place of ANSI H510-1975 as referenced in Regulat'ory Guide 1.52,* Revision 2. March 1978.
, 9 9' , ** Automatic transrer to the chlorine mode is not required when chlorine 1
- containers having a capacity of 150 pounds or less are stored 100 meters from the control room or its fresh air inlets.
'p ***This specification is not applicable af ter all chlorine containers having a capacity of 100 pounds or greater are removed from the site including the t chlorine containers located at the site sewage treatment plant.
4 -r y g 1" CLINTON - UHIT 1 ^ 3/4 7 D Amendment No.12 (o-lb bh
U-602283 Att. 2 Page 35 of 183 PLANT SYSTEMS
)
CONTROL ROOM VFNTILATION m SYSTbi gg SURVEILLANCE REQUIREMENTS (Continued) ( 4.7.2 (Continued)
%h in cKon the control room is maintained at a positive pressure of at least 1/8-inch W.G. with respect to adjacent areas during system operation j l 3.~l at a flow rate less than or equal to 3000 cfm.
- 6. Verifying that the makeup filters heaters dissipate at least 14.4 kW i
. 5.% e when tested in accordance with ANSI N510-1980.
- f. After each complete or partial replacement of a HEPA filter bank in thD ,' I makeup filter system,Jby verifying that the HEPA tilter bank satisfies The in-place penetration and bypass leakage testing acceptance criteria Ql 55G ' of less than 0.05% in accordance with ANSI N510-1980 while operating the i
' system at a flow rate of 3000 cfm i 10%. I
- g. Aftereachcompleteorpartialreplacementofacharcoaladsorberbankid the makeup or recirculation filter systemsyby verifying that the char-coal adsorcer canK satisties the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% for the makeup filter 6.6Mb system and 2% total bypass leakage for the recirculation filter system in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the makeup system at a flow rate of 3000 cfm N) i 10% and the recirculation filter system at a flow rate of 64,000 1 10%.
- h. At least once per 18 months by verifying that 'the air inleakage rate into the negative pressure portions of the Main Control Room Ventilation System ductwork located outside the Main Control Room habitability SR 313A boundary between fan OVC04CA(B) and isolation dampers OVC03YA(B) inclusive, and fire dampers OVC042YA(E), OVC042YB(F), 0VC042YC(G) and OVC042YO(H) to be s 650 cfm when tested in accordance with an NRC-a;5 proved test method. In addition, visually verify the integrity of the Recirculation Filter Housing flexible connection to fan OVC03CA(B).
r om'Go n$ of A b,0ol Gnk A $* O'b Q TC-Qg,[c(ae ag 40 % N/ Prp ,e Weyen'u 5 A7 ctdS @ X C AMENDMENT No. 82 CLINTON - U"IT I -3/4 7-0 G~lbb6o)
rEnw re 2 to 219) u-eo2283 Att. 2 Page 36 of 183 ectio 4 of I5t PLANT SYSTEMS 3/4.7.8 HAIN CONDENSER OFFGAS HONITORING_ i l 0FFGAS-EXPLOSIVE GAS HIXTURE. -
,Di' LIMITING CONDITION FOR OPERATION l
inhain condenser offgas treatment 3.7.8.1 The concentrati ' 6.gfo. system shall be limited [on of hvdreienless than or equal (3 to ! 4% bi volug hPLICA81LITY[ Whenever the main condenser air ejector is depera
)
- a. With the concentration of hydrogen in the main conde ,
b within 48 hours.
- b.
The provisions of Specifications 3.0.3 and 3.0.4 are hot applicable.
.v SURVEILLANCE REQUIREMENTS
,l.
The concentration of hydrogen in the main condenser offgas treatment b $,1, A 4.7.8.1 system shall'be determined to be within the above limits by continuously s// monitoring the waste gases in the main condenser offgas treatment systemhydrogen) whenever the main condenser evacuation system is in operatiegith the
~ j (monitors required OPERABLE by Specification 3.3.7.11. #f LAl
[nuar /tQ J fe n _m
-' p n ~ c , e ce s.oa _ s ,, w ,
' "c c / p b c d k k 1A c e yp .& die y n c/ []cg_ c h ,. gD Mf L /Mf kce
/ ,
f 4"^ r-dm-~ Yt*[ h k'uf sendes ,
'EN S E A T '
A=endment No. 40 l Ctun0t C .,:: p -itt m l 6 - /G b (/Q _
- - . w 4 ,e v w + w
u-oo2283 Att. 2 [ sure 2 t 96 PLANT SYSTEMS
#N 3/4.7.7 LIOUID STORAGE TANXS*
LIMITING CONDITION FOR OPERATION G 5. 7 3.7.7 The quantity of radioactive material contained in each of the following g'6I'g unprotected outdoor tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.
- a. Cycled Condensate Storage Tank
- b. RCIC Storage Tank
- c. Outside temporary tank ,
1 APPllCABiliTY: At all times.
- a. With the quantity of radioactive material in any of the above listed
[Al tanks exceeding the above limit, immediately suspend all additions of , radioactive material to the tank, within 48 hours reduce the tank f contents to within the limit, and describe the events leading to this condition in the next Radioactive Effluent Release Report. l
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
> \
SURVElllANCE RE0VIREMENTS < l 4.7.7 The quantity of radioactive material contained in each of the above ' listed tanks shall be determined to be within th_e above limit by arialyzing a sentativeJampirof_ adioactive materials are the tank's being addedcontonte fit least once per to the tank. - 7 days whep gg7.b : A {JRSERT f ~ L fecdsisNS e f 5/2 1.D,2 a n al 3 6, 3 s:.rs o.pp uca ble to thG EXf4cNe 6M ud ' S h e n c- f ~ d M. clio A ch; j W% r1'o c'~ f P m cat n Tc 57 frefvenoQ ~ g '6 *
- Tanks included in this' specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. I 1~A/ SEA T CLINTON P 1) 173 AMENOMENT -NO . 76 c-ics (a)
. 3 U-802203 Att. 2 ELECTRICAL POWER SYSTEMS p.g 3 or ie3 Tacio A_C, SOURCES - OPERATING SectioerStage 0219]6 (4 56}
'55,8 S'JRVElttANCE REOUTREMENis (Continued) 4.8.1.1.2 (Continued) .
- 5. Verifying the diesel generator than or equal to 3869 kW for
\h'k ,onized, loaded to greater A .nerator IA, 3875 kW for diesel
( j generator 18 and 2200 kW fr Gp,# generator 1C in less than or with this load for at ledst i equal to 90l* seconds, ane jg 60 minutes, l
- 6. Verifying the diesel hy to the associated ,h cu.ses.
is aligned to provide standby power Verifying the r a all diesel generator air start receivers 6Q 7. to be greater squal to 200 psig,
- b. At least once
- s and after each operation of the diesel where the period o r b a,was greater than or equal to I hour by checking for and rer amulated water from the day fuel tanks.
- c. At least . 92 days by removing accumulated water from the fuel storage to .xs.
ES ,g d. By sampling fuel oil and verifying that the sample meets the following l minimum requirements and is tested within the specified time limits:
- 1. By obtaining a sample from new fuel oil in accordance with ASTM-5.c5.0.ct. D270-1975 and verifying prior to addition of the fuel to the storage tanks that the sample has:
a) A water and sediment conten[ss than or equal to 0.05 C's ' G a ,'S y lume percent wnen teste n acc rdance with the tests specified in ASTH-D975-89;(or a clear and bright appearance when tested in accordance with ASTM-04176-R23
% ~
b) A kinematic viscositydt 40*C of greater than or equal to SS&, l.,g { 1.9 centistokes but less than or equal to 4.1 centistokes when tested in accordance with the tests specified in ASTH-D975-8 . c) An API gravityft 60*F of greater than or equal to 30 degrees i
'b ut le vssan or equal to 40 degrees; or an absolute specific $5.04 I kgravity at 60/60*F
_ to 0.89.of fgreater than or equal to 0.83 but less i
~
han _or equal '
$58b 2.
By obt;dning a sample from new fuel oi[1accordance with ASTH-09 975 and veritying within .it days after obtaining the sample l 1 i Surveillance testing to verify the erator start and load times I (less than or equal to 12 ser" l .oan or equal to 90 seconds ! respectively) from ambien
- p g 4, .hs .all be performed at least of once A
per 184 days. All ot' performed for the purpose meeting these sura' with warmup a' s % [OO s as recommended by. the V s; h uur'e.emen manufacturer. This is in dC,1 . ze mechanical stress and wear on the diesel - generator ca , r ast starting and load,ing of the diesel generator. , n wrnu n . 4;,_; y d(6,ER g} I _ _ ___ U
.. n a m. r%cn 3 3.en3 A.q SOURCES - OPERATING
[ [ ] 1 SURVEll.l ANCE REf)UIRFMENTS (Continued) SecJ Page %of 4.8,1,1,2 (Continued) that the other properties specified in Table 1 of ASTM-0975-89 are O'cg'b
- met when tested in accordance with the tests specified in ASTH-
~
~ --
l 0975-89. - , h 3. By obtain'ina a sample of fuel oil from the storage _tankr in ccordance with ASTH-02276-88 at least once per 31 days and g, $ ,6. c, ' verifying within one week after obtaining the sample that total particulate contamination is less than 10 mg/ liter when tested in accordance with ASTM-02276-88. f ~ l e. At least once per 18 months,' during shut 5, y:
- 1. Subjecting the diesel to an insper .ccordance with procedures prepared in conjunction with its .rer's recommendations for this class of standby service.
- 2. Verifying the diesel generate sty to reject a load of greater than or equal to 11T S diesel generators IA and 18, and greater than or equal * , V . for diesel generator IC while maintaining engine speed plus 75% of the difference j between nominal speed an g,cspeedtripsetpointor15%above l nominal whichever is le 9 l 3 l l 3. Verifying the diesel capability to reject a load of 3869 '
)
l kW" for diesel gener 3875 kW* for diesel generator 18 and l
.r IC without tripping. The generator 2200 kW* for diese' I voltage shall not /
O J00 volts for diesel generator lA and 18 and 5824 volts f /g generator IC during and following the load rejection. ()\
- 4. Simulating a sffsite power by itself, and:
a) For D' I and 11: - l l
- 1) g .ng deenergization of the emergency buses and load ing from the emergency buses, ifying the diesel generator starts on the auto-start
.gnal, energizes the emergency buses with permanently
.;onnected loads within 12 seconds, energizes the auto-connected loads required for safe shutdown through the c load sequence (individual timers), and operates for L greater than or equal to 5 minutes while its generator is M7 loaded with the shutdown loads. After energization, the steady state voltage and frequency of the emergency buses shall be maintained at 4160 1 420 volts and 60 i 1.2 Hz during this test.
/ start of a diesel, the diesel must be operated with a load in accordance ne manufacturer's recommendations.
.omentary transients due to changing bus loads shall not invalidate the test.
Cl 1NTON [T-@ %G EM Amendment No.80 r /cwo
, ,- ,-_ ,__ __,m, _ . - . . - .
i . ADMINISTRATIVE CONTR0ls - g 6.9 REPORTING RE0VIREMENTS u. sones Ati. 2 Page 810 of 103 ROUTINE REPORTS 6.9. In additi n t the applicable reporting requirements of Title 10, Code 5i G of Federal Regulations, the following reports shall be submitted \to the 7 Regional Administrator, of the Regional Office of the fiKc unless otherwise} TRU REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an Operating License. (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. 6.9.1.2 The startup report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and speci-b$ fications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. , 6.9.1.3 Startup Reports shall be submitted within (1) 90 days following com-l pletion of the startup test program, (2) 90 days following resumption or com-l mencement of commercial power operation, or (3) 9 months following initial cri- ,, ticality, whichever isi carliest. If the startup report does not cover all ' three events (i.e., initial criticality, completion of startup test program, - and resumption or commencerent of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been . i , completed. ANNUAL REPORTS g 6.9.1.4 Annual reports covering the activities of the unit as described below forthepreviouscalendaryearshallbesubmittedprierto[oadiof- h o i c c 6.9.1.5 Reports required on an annual basis shall include: I
- a. A tabul'ation on an annual basis of the number of plant, utility, and other
'N ' l personnel (including contractors), for whom monitoring was required, l receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions * (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special main-tenance, waste processing, and refuelings). The dose assignments to various duty functions may be l !
l l l 56,\ *This tabulation supplements the requirements of 520.2206 of 10 CFR Part 20. l CLINTON -Ut"7 4a -- 6-17 Amendment No. 69
u-co22e3 Att. 2 n re 2 to 19, Page 41 or 133 ScCtio ACHINISTRATIVE CONTROLS ' Q fli { ANNUAL REPORTS (Continued) estimated or based film badge on pocket dosimeter, thermoluminescent dosimeter (TLD), measurements. h'j Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions: f.(i .q b. f0ccumentation L and of all challenges to safety valves or safety / relief valves, 6 c Any otiier unit unique report.s required on an annual basis. N/2. The results of specific activity analysis in which the primary toolant exceeded the limits of Specification 3.4.5 should include the following A information: (1) Reactor power history starting 48 hours prior to the - fl{ first sample in which the limit was exceeded; (2) Results of the last iso-topic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis' after the radioiodine activity was reduced :o less than limit. Each result should include date and time of samplir.1 and the radiciodine concen-trations.; (3) Clean up system flow history stari.ing 48 hours pr'or to the . first sample in which the limit was exceeded; (4) Graph of the I-131 con- ; m, centration and one other radiciodine isotope concentration in microcuries / per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit. ? ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6. 9.1. 6 The Annual Radiological Environmental Operating Report covering the M'g operation of the unit during the previous calendar year shall be submitted bafore av $ of each year. The report shall include summaries, .
jf interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the 00CM arrd - (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. T
, 1blSE M M1 18 A 3 t
CLINTON. 6-18 Amendment No. 40
- ""f;;,' P M IN i INSERT 18A The Annual Rad.ological Environmental Operating Report shall include the results o' analyses of all radiological environmental samples and of'all enrironmental. radiation measurements taken during the period pursuar. ; to the locations specified in the table and figures in the ODCM, a' well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. The report shall identify the TLD results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
l i l i l l
~
INSERT CLINTON 6-18 10/1/93 i i A- ,. - - - , _ < - r - - - . - . - . ,,,,w,,m,4, ----,r,--, m, ,mv, ,. .w-. ,r .-.,-,,-.,---4re,,,4-~,
U-602203 Att. 2 Page 84 3 o f 183 l l ADMIT 11STRATIVE C0t1TROLS RADI0 ACTIVE EFFLUEllT RELEASE REPORT CA b30 3( VfG @ b)
~
tl 0. 9.1. 7 - The Radioactive Effluent Relear Report covering the operation of the Unit uring the previous (T2 month; cf scraticDshall be submitted prior to
- May of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCMgr" P_Cpand (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Ap.endix I to 10 CFR Part 50.
LA2 i I l l CLIt1T0t1 J.JuJ-T 6-19 AMErlDMEf1T ff0. 76
F r 2 Enc 7t 1 U-002263 Atta Page " of 183 S age f]i t i l l O e t mpm
@e s - - -
m u rout At4 6 th 61UrmLLY UtLt w. % e 1 1 i l l
-O CLINT0ft' 6-20 Amendment No. 40
s U-002283 Att. 2 Page 45 or 183 s
.O ';
t ADMINISTRATIVE CONTROLS I MONTHLY OPERATING REPORTS 5 1,4 s.9.1.8 Routine reports of operatins statistics and shutoown experience, including documentation of all challenges to the main steam system safety / relief valves, shall be submitted on a monthly basis /to the Docusein Cuatroi
#Tesk, U.S. Nuclear Regulatory Commission, Washington. 0.C. 20555, with a copy
() (,.to the Recional Administrator of the Regional Office of the NRCJo later won
' the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT ryggTJ 2M
' C]
6.9.1.9 Core oper ng limits shall be established and documented in the CORE OPERATING LIMIT nEPORT before each reload cycle or any remaining part of a reloa'd cyc.le. The analytical methods used to determine the core operating limits shall be those previous 1y reviewed and approved by the NRC in General Electric Standard Application for R_eactor Fuel (GESTAR) NEDE-240ll-P-A-8, as arrended (latest approved version)falid Maximum Extended Operating Domain anc ]
- f Feedwa ter Hea ter Ou t-or-der _v n.e analysis for Clinton Power Sta tion,
.b (NEDC-31546P, August 1988. (The core operating limits shall be determined so j
that all application limits (e.g., fuel thermal-mechanical limits, core 1 i thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUT 00WN MARGIN, and transient and accident analysis limits) of the safety analysis are met. THE CORE OPERATING LIMITS REPORT, including any mid-cycle revision or supplements thereto, for each reload cycle, shall be submitted upon issuance to the NRC .4 Document Inspector. Control Deskg- _with copies to the Regional Administrator and Resident pi7- ,
,- gg pg gg SPECIALREPORTS(
l f % ...(RegionalSpecial
... 6.9.2 reports shall be submitted to the Regional Administrator of the Office of the NRC within the time period specified for each report. '
T 10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10,3 Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. 6.10.2 The following records shall be retained for at least 5 years:
~~
a. j Records level, and logs of unit operation covering time interval at each power
- b. Records and logs of principal maintenance activities, inspections, repair, j and replacement of principal items of equipment related to nuclear safety.
k ! c. All REPORTABLE EVENTS. l l
- d. Records of surveillance activities, inspections, and calibrations required j i
l by these Technical Specifications and the Fire Protection Program. ' j CLINTON - UNIT 1 6-21 Amendment No. 40 1 c ., - w.,.,-e
U-602283 Att. 2 re 2 - 196 Page 46 of 103 Sect a M 56 INSERT 21A
, and shall be documented in the COLR for the following:
- 1) LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR);
- 2) LCO 3.2.2, Minimum Critical Power Ratio (MCPR); ;
- 3) LCO 3.2.3, Linear Heat Generation Rate (LHGR) ;and
- 4) LCO 3.3.1.1, RPS Instrumentation (SR 3.3.1.1.14).
1 1 l l l CLINTON INSERT 6-21 [\) 10/1/93 i
\
EMERGENCY CORE COOLING SYSTEMS ~ l u-co22ea Att. 2 r ! Enc re 2 to ECCS -OPERATING Page 47 or 183 , 9f.. ! ! Sectio
' f 156 */
l LIMITING CONDITION FOR OPERATION (Continued) . 3.5.1 ACTION (Continued): - '
- e. For.ECCS Divisions I and II, provided that E' .II is OPERABLE and Divisions I and II are otherwise OPERAP .
- 1. With one of the above required ADS h .able, restore the 1
inoperable ADS valve to OPERABLE La a 14 days or be in at ! least HOT SHUTDOWN within the o d ! and reduce reactor steam dome pressure to f 100 psig w' .c 24 hours.
- 2. With two or more of the ab' ADS valves inoperable, be in ;
at least HOT SHUTDOWN wi+ and reduce reactor steam dome i pressure to f 100 psig
- 0 . ext 24 hours . I
- f. With an ADS accumulator 1 C' alarm system instrumentation channel (s) inoperable:
- 1. Determine the r .3 accumulator system pressure from- l
' alternate ind g ,erify that ADS accumulator pressure is 1 greater tha-140.psig. at least once per 12 hours, y6 t
t p l 2. Restore .e ADS accumulator low pressure alarm system instru- ! Y .nnel(s) to OPERABLE status within 30 days or ! subm' '
.<eport to the Commission pursuant to Specification 6.9 ; next 10 days outlining the cause of the malfunction anc for restoring the instrument (s) to OPERABLE status, '
- 3. The pro, ,ons of Specification 3.0.4 are not applicable.
- g. In the event an ECCS system is actuated and injects water into the Reactor l
Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 wi. thin 90 days describing the circumstances of the actuation and the total accumulated actuation cycles
. to date. The current value of the usage factor for each affected safety i injection nozzle shall be provided in this Special Report whenever its lue exceeds 0.70.
j 1 SURVEILLANCE REOUIREMENT' d 4.5.1 ECCS Divi 41 shall be demonstrated OPERABLE by:
- a. At l' ssc ,I Ah.js for the~ LPCS, LPCI, and HPCS systems:
.f 1. /tO j venting at the high point vents that the system piping j h . pump discharge valve to the system isolation valve is filled sater.
IkSQ f CL1HTOH $ Q A 3/4 S y Amendment No. 36 6- f/
i u-602283 Att. 2 Page 40 of 183 ELECTRICAL POWER SYSTEMS AC SOURCES - OPERATING SURVEILLANCE REQUIREMENTS (Continued)
~
[4.8.1.1.2 (Continued)
- 13. Verifying that the sequence times for loads automatically sequenced by individual timers are within 10% of their design interval for each load block for diesel generators IA and 18.
- 14. Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:
a) Maintenance mode. b) Diesel generator lockout. f. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting all three diesel generators simultaneously, during shutdown, and verifying that all three diesel gener-ators accelerate to,at least 900 18 rpm in less than or equal to 12 second .
- g. At lea'st once per 10 years by:
- 1. Draining each fuel oil storage tank, removing the accumulated sedi-ment and cleaning the tank using a sodium hypochlorite solution or- ~
equivalent, and
~
- 2.
- Performing a pressure test of those portions of.the diesel . fuel oil system designed 'to Section III, subsection ND of the ASME Code in accordance with ASME Code Section 11 Article IWD-5000.
1.1. 3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2, within 30 days.
,,, Reports of diesel generator failures shall include the information recommended '
in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977 If the number of failures in the last 100 valid tests of any diesel { gene. rator is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C 3 b of . . j
@gulatory Guide 1.108, Revision 1, August 1977.
}
'@setr CLINTON -444T_ k'2--
G-91 (g
-3/4 C-,
Amendment No. 49
j u-sone 8 Att. 2 1 TABLE 3.3.7.5-1 (Continued) "*8' *8 oP rea j ACCIDENT MONITORING INSTRUMENTATION ACTION ] ]' j ACTION 80 - a. With the number of OPERABLE accident monitoring instrumenta-
! tion channels less than the Required Number of Channels shown j in Table 3.3.7.5-1, restore the inoperable channel (s) to J
p OPERABLE status within 7 days or be in at least HOT SHUTDOWN i within the next 12 hours. The provisions of Specifica-l tion 3.0.4 are not applicable. I b. With the number of OPERABLE accident monitoring instrumen-i i N g4F tation channels less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1 restore the inoperable
! channel (s)toOPERABLEstatuswithin48hoursorbeinat j ;
least HOT SHUTOOWN within the next 12 hours. 1 { ACTION 81 - With the number of OPERABLE Channels less than required by the ! j Minimum Channels OPERABLE requirement, either restore the i
' inoperable Channel (s) to OPERABLE status within 72 hours, or: ;
l a. Initiate the preplanned alternate method of monitoring thej l
! accrooriate caramatarM. and e- l l b. Prepare and submit a Special Report to the Commission purd l 1 suant to Specification 6.9.2 within 14 days following the !
4 j [A3 ,_ event outlining the action taken bility and the plans and schedule, the cause of the inopera l j to OPERABLE status. f j ___ / c . The provisions of Specification 3.0.4 are not applicable, i ACTION 82 - a. With the number of OPERABLE accident monitoring instrumentation } channels less than the Required Number of Channels shown in l* Table 3.3.7.5-1, verify the valve (s) position by use of alter-nate indication methods; restore the inoperable channel (s) to j OPERABLE status within 30 days
- or be in at least HOT SHUTOOWN l
, within the next 12 hours and in COLD SHUTDOWN within the fol-lowing 24 hours.
j b. With the number of OPERABLE accident monitoring instrumentation ! I channels less than the Minimum Channels OPERABLE requirements i of Table 3.3.7.5-1, verify the valve (s) position by use of i alternate indication methods; restore the inoperable channel (s) i
\ to OPERABLE status within 7 days or be in at least HOT SHUT-DOWN within the next 12 hours and in COLD SHUTDOWN within the j following 24 hours.-
- c. The provisions of Specification 3.0.4 are not applicable.
l i *For valve 1821-F0220, operation may continue until the first reactor shutdown
- after June 3,1991, provided that a. planned alternate method for determining i the post-accident isolation status of the associated containment penetration
; is implemented.
4 O
%WT i CLINTON U:GT i 3/4 0 00 Amendment No. 58 (o-Z [h i
l
7 ADMINISTRATIVE CONTROLS u-no2283 Att. 2 RECORD RETENTION (Continued ;
- e. Records of changes made to the procedures required by ' Specification 6.8.1. I
- f. Records of sealed source and fission detector leak tests and results. I
- g. Records of annual physical inventory of all sealed source material of l record. _
6.10.3 The following records shall be retained for the duration of the unit Operating Licens~e:
- a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.
- b. Records of new and irradiated feel inventory, fuel transfers, and assembly burnup histories.
- c. Records of doses received by all individuals for whom monitoring was required.
,. d. Records of gaseous and liquid radioactive material released to the environs.
' \/ e. Records of transient or operational cycles for those unit components i identified in Table 5.7.1-1. l
- f. Records of reactor tests and experiments.
- g. Records of training and qualification for current members of the unit staff.
l h. Records of inservice inspections performed pursuant to these Technical l Specifications. j
- i. Records of quality assurance activities required by the Operational Quality Assurance Manual 'not listed in Section 6.10.2, which are l classified as permanent reccrds by applicable regulations, codes, anj '
standards. I l j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- k. RecordsofmeetingsoftheFRGandtheNRAG.f
?
6-22 CLINTON - U"IT f Amendment No. 69 ) I
.-. - . .- . . . . . . . . - ~ . . - - - . .. . -. . - . . -.
vu i uvause 4 su v.v i - u-002203 Att. 2 SCClio e O of 156 j ADMINISTRATIVE CONTROLS P* si e 183 hCORORETENTION(Continued 1
- 1. Records of the service lives of all snubbers including the date at which the service life commences and associated installation and maintenance 1 records.
~
,Ng m. Records of analyses required by the radiological environmental monitoring l program that would permit evaluation of the accuracy of the analysis-at a I later date. This should include procedures effective at specified times and QA records showing that these procedures were followed. j
- n. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
- o. Records of radioactive shipments [
f 6.11 RADIATION PROTECTION PROGRAM . 2 6.11.1 Procedures for personnel radiation protection shall be prepared consis-y tent with the requirements of 10 CFR Part 20 and shall be approved, maintained, j 1 V and adhered to for all operations involving personnel radiation exposure. J
.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by para-graph 20.1601 of 10 CFR Part 20, each high radiation area in which the !
intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and I 4 (entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).* Any individual or group of individuals permitted to enter such ' f areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is !
received. Entry into such areas with this monitoring device may be made l Op after the dose rate levels in the area have been established and personnel have been made knowledgeable of them. !
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose. rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Protection Supervisor in the RWP.
f i
- Radiation protection personnel or personnel escorted by radiation protection personnel shall be exempt from the RWP issuance requirement for fields of less than 3000 mrem per hour during the performance of their assigned radiation l
7rotection duties, provided they are otherwise following plant radiation l eprotection procedures for entry into high radiation areas. CLINTON 6 0tHT_ [ 6-23 Amendment No. 69 l
' u-602283 Att. 2 ADMINISTRA.TIVE CONTR0tS _ Page 52 of 153 j ,
! 6.12 HIGH RADIATION AREA (Continued) 6.12.2 in addition to the requirements of Specification 6.12.1, areas accesD sible to individuals with radiation levels such that an individual could J 3 receive in I hour a dose greater than 1000 mrem
- but less than 500 rads t.t one I meter from sources of radioactivity, shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Radiation l
Protection Supervisor. Doors shall remain locked except during periods of access by personnel under an approved RWP.** For areas accessible to individuals with radiation levels such that an individual could receive in I hour a dose in excess of 1000 mrem
- but less than 500 rads at one meter from
- sources of radioactivity, that are located within large areas, such as the l $ containment, where no enclosure exists for purposes of locking, and no l
enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted, and a flashing light shall be activated as a warning device. 6.12.3 In addition to the requirements of Specifications 6.12.1 and 6.12.2, for areas accessib.le to individuals such that an individual could receive in 1 hour a dose in excess of 3000 mrem
- but less than 500 rads at one meter from i
sources of radioactivity, entry shall require an approved RWP which will specify dose rate levels in the immediate work area and require that stay times shall be established. l' In lieu of the stay time specification of the RWP, continuous surveillance, direct or. remote (such as use of closed circuit TV cameras), may be made by personnel qualified in radiation protection procedures to provide positive texposure control over the activitics within the area.
,) -
6.13 PROCESS CONTROL PROGRAM (PCP) Changes to the PCP: TN)SEET CTS l 32-j
- a. Shall be documented and recobs of reviews performed shall be retained as required by Specification 6.10.3.n. This documentation ;
shall contain: I
' l. Sufficient information to support the change together with the 1
\ appropriate analyses or evaluations justifying the change (s), !
and )
- 2. A determination that the change will maintain the overall con-formance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
)
' 1
- b. Shall become effective after review and acceptance by the FRG and the !
approval of the Manager - Clinton Power Station. b
- Measurements made at 30 cm (12 inches) from sources of radioactivity.
,- ** Radiation protection personnel or personnel escorted by radiation protection personnel shall be. exempt from the RWP issuance requirement for fields of less than 3000 mrem per hour during the performance of their assigned radiation pro-
)lJ' tection duties, provided they are otherwise following plant radiation protection procedu.es for entry into high radiation areas.
6-24 Amendment No. 69 CLINTON - UNIT 1
U-802283 Att. 2 page 53 of 183 1 DEFINITIONS > PRIMARY CONTAINMENT INTEGRITY 1.31 PRIMARY CONTAINMENT INTEGRITY shall ex4
- a. All primary containment penetratio- Je closed during accident I conditions are either:
- 1. Capable of being close \\ _C containment automatic isolation system or Closed by at le' W9 valve, blind fl' nge, or deactivated
- 2. 1 a
.ts closed position, except as provided in automatic val-Specificati If l
- b. All primary .spment hatches are closed and seal'ed. I
.c air lock is in compliance 'with the requirements of
- c. Each pri I Speci# .
- d. T' ainment leakage rates are within the limits of Specification l r
4 :ssion pool is in compliance with the requirements of ; eation 3.6.3.1. '
- f. . sealing mechanism associated with each' primary containment penetration, e.g., welds, bellows or 0-rings, is OPERABLE.
- I !
PROC (SS CONTROL PROGRAM (PCP) i I 1.32 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and ,
- t packagina of solid radioactive waste Ubased on aemonstrated crocessino or actuat> f 1 tr si_mularad war eniid wastes ATTl be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71 and Federal and i State regulations, burial ground requirements and other requirements governing the I disposal of the radioactive waste.
PURGE - PURGING . I 1.33 PURGE or the controlled process of discharging air or gas from a i confinement mperature, pressure, humidity, concentration or other operatir- . a manner that replacement air'or gas is required to puri f- l { c, _3;g i
- f. b INERMAL POWER shall'be'a total reactor core heat transfer rate to the coolant of 2894 MWt.
.Z A/SGA T
-1 : Amendment No. AD, 68 CLINT0t@'"!T M c-2y CD
u-602203 Att. 2 ure 2 to ADMINISTRATIVE CONTROLS g j 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM) Changes to the 00CM': UI C U /' 2 1 Shall be documented and records of reviews performed shall be
- a. l retainedcrs-recurred by saec4uhtra n.a This documentation l shall contain:
~
- 1. Sufficient information to support the change together with the l appropriate analyses or evaluations justifying the change (s),
and
- 2. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Pait 190,10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reli ility of effluent, dose, or setpoint calculations. j % /;
- b. Shall become effective after review and acceptance (u U e FZ aM the approval of the 4 tanager - Clinton Power 5tation
- c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire 00CM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the 00CM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.
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l i I l CLINT0ftQtM 6-25 AMENDMENT NO. 76 4 l
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p, ssor[ea OEFINITIONS ti[NiMUM CRITi gd )g CM M 1.25 The MIN exists in the [ 90 .....v
. (MCPR) shall be the smallest CPR which d' 0FFSITE DOSE CALCULATION MANUAL (ODCH) 1.26 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactiv Effluent Controls Gnd Radiological EnvironmentaT) d' oni t o ri nOProgra uired by sie t' i n n A R n and (2) descriptions of the information that hould be incldded in the Annual Radioinnie En"4ronmantal Operatinn and Radioactive'Efflu6nt Release Reports,geluired by SpecificatioD }
d .9.1.6 and 6.9.1.7 1 , , _- OPERABLE - OPERABILITY 1.27 A system, subsystem, train, compo- .: shall be OPERABLE or have OPERABILITY when it is capable o' .ts specified function (s) and when all necessary attendant in' , controls, electrical power, cooling or seal water, lubricatior .iliary equipment that are required for the system, subsyst inponent or device to perform its function (s) are also capable e cheir related support function (s). OPERATIONAL CONDITION - C' l.28 An OPERATIONAL r s.,' CONDITION,-shall'be any one inclusive combination of mode .on and average reactor coolant temperature as specified in Tab 1r hp PHYSICS TESTS 1.29 PHYSI'
)h , be those tests performed to measure the fundamental nuclear ' .s of the reactor core and related instrumentation as 1) descrik 14 of the FSAR, 2) as authorized under the provisions of 10 Cr fj ,) as otherwise approved by the Commission.
P' aRY LEAKAGE sRE BOUNDARY LEAKAGE shall be leakage through a nonisolable fault in coolant system component body, pipe wall or vessel wall. w
.ZAISER T
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CLINTON -:
. AMENOMENT NO. 76 c, - 2 r o)
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U-607263 Att. 2 Enclosu e... . ., ,,, ( A'I"I'ACHMENT 1B ! 1 CTS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES eh me 4 4 I I l l
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l r l U-602283 Att. 2 Page $7 of 103 i DISCUSSION OF CHANGES CTS: 6.1 - RESPONSIBILITY l ADMINISTRATIVE . l A.1 Where possible, plant specific management position titles in the current Technical Specifications are replaced with generic titles as provided in ANSI /ANS 3.1. Personnel who fulfill these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details ; relating to the plant specific management position titles are j identified in licensee controlled documents (such as the USAR) . < The two major specific replacements are the generic " plant ! manager" for the manager level individual responsible for the ! overall safe operation of the plant and the generic descriptive ! use of "the corporate executive responsible for overall plant ; nuclear safety" in place of the Vice President position. The plant specific titles fulfilling the duties of these generic i positions will continue to be defined, established,' documented i and updated in a plant controlled document with specific regulatory review requirements for changes, such as the USAR or Illinois Power Nuclear Program Quality Assurance Manual (QAM). This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation ! of the corporate and unit organization charts to licensee l controlled documents. The intent of the Generic Letter, and of l this proposed change, is to reduce the unnecessary burden on l NRC and licensee resources being used to process changes due solely to personnel titles changes during reorganizations. Since this change does not eliminate any of the qualifications, , l responsibilities or requirements for these personnel or the positions, the change is considered to be a change in presentation only and is therefore administrative. A.2 Lines of authority are also required to be defined and maintained in the USAR by TS 5.2.1.a. Repeating the t organizational responsibilities via an internal management directive only increases the administrative burden on the ; facility with no resulting benefit. Since the actual lines of authority are not affected by this change and since the requirement to define these lines of authority in the USAR will j remain in affect this is considered an administrative change. RELOCATED SPECIFICATIONS { None in this section. i i l 1 CLINTON 1 4/15/94 1
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u.002283 Att. 2 Page 50 of 183 I DISCUSSION OF CHANGES l CTS: 6.1 - RESPONSIBILITY l TECHNICAL CHANGES - MORE RESTRICTIVE None in this section, 1 TECHNICAL CHANGES - LESS RESTRICTIVE j None in this section. l i i I l l l l l I l i i i l CLINTON 2 4/15/94
l U-602283 Att. 2 l Page 59 or 183 l DISCUSSION OF CHANGES l CTS: 6.2 - ORGANIZATION ADMINISTRATIVE A.1 This comment number is not used for this station. A.2 This comment number is not used for this station. A3 This comment number is not used for this station. A.4 Where possible, plant specific. management position titles in the current Technical Specifications are replaced with generic titles as provided in ANSI /ANS 3.1. Personnel who fulfill these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details relating to the plant specific management position titles are identified in licensee controlled documents (such as the USAR) . The two major specific replacements are the generic " plant manager" for the manager level individual responsible for the overall safe operation of the plant and the generic descriptive use of "the corporate executive responsible for overall plant nuclear safety" in place of the Vice President position. The plant specific titles fulfilling the duties of these generic positions will continue to be defined, established, documented and updated in a plant controlled document with specific regulatory review requirements for changes, such as the USAR or QAM. This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation of the corporate and unit organization charts to licensee controlled documents. The intent of the Generic Letter, and of this proposed change, is to reduce the unnecessary burden on NRC and licensee resources being used to process changes due solely to personnel titles changes during reorganizations. Since this change does not eliminate any of the qualifications, responsibilities or requirements for these personnel or the positions, the change is considered to be a change in presentation only and is therefore administrative. A.5 This comment number is not used for this station. i A.6 This comment number is not used for this station. A.7 This comment number is not used for this station. A.8 Redundant details regarding the description of overtime ! constraints have been eliminated. These changes are considered l to be changes in presentation only and are therefore ' administrative. A.9 This comment number is not used for this station. I l CLINTON 3 4/15/94 ) l
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U 602283 Att. 2 page 60 of 103 l l DISCUSSION OF CHANGES I CTS: 6.2 - ORGANIZATION l l A.10 The requirement for an SRO'to be present during fuel handling ; and core alterations is contained in 10 CFR 50.54. Therefore, l there is no need to repeat these requirements in the Technical l Specifications. Since the requirements remain the same, this I change only affects the presentation method and is considered I administrative. 1 RELOCATED SPECIFICATIONS None in this section. TECHNICAL CHANGE - MORE RESTRICTIVE ) l M.1 This comment number is not used for this station. l M.2 This comment number is not used for this station. l 1 M.3 This comment number is not used for this station. i 1 l TECHNICAL CHANGE - LESS RESTRICTIVE I
" Generic" l LA.1 This comment number is not used for this station.
l LA.2 Details of the educational requirements of the Shift Technical Advisor have been relocated to the USAR and procedures. , Additionally, changes to the procedures and the USAR are l controlled in accordance with 10 CFR 50.59. l j LA.3 The staf fing requirements of Table 6.2.2-1 are relocated to the
- USAR and procedures. The requirements of Table 6.2.2-1 are
! removed for the Technical Specifications and will be controlled by the licensee's administrative controls. 10 CFR 50.54 provides the requirements for shift complement regarding licensed operators. Additionally,' the Technical Specifications will continue to specify when a licensed operator must be in the control room. The Table 6.2.2-1 requirements associated , with the auxiliary operators are retained as 5.2.2.a with the 4 associated allowance for unexpected absences retained in 5.2.2.c. Changes to the procedures and the USAR are controlled in accordance with 10 CFR 50.59. LA.4 This comment number is not used for this station. LA.5 The requirements relating to the Independent Safety Engineering Group (ISEG) is relocated to the USAR and procsdures. Changes CLINTON 4 4/15/94 l
_ _ _ _ . __. ._m i U-602283 Act. 2 Page el of 183 DISCUSSION OF CHANGES CTS: 6.2 - ORGANIZATION l to the procedures and the USAR are controlled in accordance ! with 10 CFR 50.59. i LA.6 This comment number is not used for this station. LA.7 These specific fire protection issues are covered by a more ; generic program (i.e. , Fire Protection Program) which provides ! for the definition and implementation of these details. , Therefore, it is not necessary to specifically identify each of ; these fire protection issues in the TSs. Changes to the fire ; protection program are adequately controlled by a license '
-condition which allows changes to the approved' fire protection i program only if they do not adversely affect the ability to l achieve and maintain safe shutdown of the plant. " Specific" L.1 It is' proposed to permit the STA role to be filled by an on-I shift Shift Supervisor or SRO provided that the individual meets the Commission Policy Statement on Engineering Expertise on Shift. Although this has not been a stated requirement in l l the present Technical. Specifications, it has always been the l l stated NRC policy, that the STA function could be fulfilled by !
l an SRO as long as the. SRO had the necessary Engineering l background to perform the tasks. The fundamental requirement remains that someone on shift be qualified technically to evaluate events from an engineering perspective, and make necessary engineering recommendations during the' initial event occurrences. L.2 This comment number is not used for this station. i CLINTON 5 4/15/94
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U-602283 Att. 2 PaQe 62 of 183 l DISCUSSION OF CHANGES CTS: 6.3 - UNIT STAFF QUALIFICATIONS i I l ADMINISTRATIVE ' A.1 This. comment number is not used for this station. RELOCATED SPECIFICATIONS None in this section. TECHNICAL C'HANGES - MORE RESTRICTIVE None in this section. TECHNICAL CHANGES - LESS RESTRICTIVE " Generic" LA.1 This comment number is not used for this station. LA.2 This comment number is not used for this station. " Specific" L.1 This change removes the requirement for the Director-Plant Operations (DPO) to have held an SRO License. However, the Assistant Director-Plant Operations, who reports to the DPO, shall hold an SRO License. Consequently, adequate qualifications are maintained. I l CLINTON 6 4/15/94
] l 4 U-602283 Att. 2 Page e3 or gg3 DISCUSSION OF CHANGES CTS: 6.4 - TRAINING l ADMINISTRATIVE A.1 This comment number is not used for this station. 1 A.2 Plant specific titles provided as per CPS letter (i.e, ; Technical Specification Change Request) U-602139, dated June i 18, 1993. l ! 1 A.3 This comment number is not used for this station. ! RELOCATED SPECIFICATIONS i None in this section. l TECHNICAL CHANGES - MORE RESTRICTIVE None in this section. TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic" l l
LA.1 The requirements relating to the retraining and replacement j program are relocated to the USAR and procedures. Changes to l the procedures and the USAR are controlled in accordance with 10 CFR 50.59.
" Specific" None in this section.
U CLINTON 7 4/15/94
i l U-002283 Att. 2 page 64 of 183 DISCUSSION OF CHANGES CTS: 6.5 - REVIEW AND AUDIT ADMINISTRATIVE A.1 This comment number is not used for this station. l 1 A.2 This comment number is not used for this station. A.3 This comment number is not used for this station. A.4 This comment number is not used for this station. A.5 This comment number is not used for this station. A.6 This comment number is not used for this station. A.7 This comment number is not used for this station. A.8 This comment number is not used for this station. ! A.9 This comment number is not used for this station. l l A.10 This comment number is not used for this station. A.11 This comment number is not used for this station. A.12 This comment number is not used for this station. 1 RELOCATED SPECIFICATIONS None in this section. l TECHNICAL CHANGE - MORE RESTRICTIVE M.1 This comment number is not used for this station. j M.2 This comment number is not used for this station. M.3 This comment number is not used for this station. l M.4 This comment number is not used for this station. M.5 This comment number is not used for this station. l TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 This comment number is not used for this station.
LA.2 This comment number is not used for this station. CLINTON 8 4/15/94 1
l l U-602283 Att. 2 Page 65 of 183-l i DISCUSSION OF CHANGES CTS: 6.5 - REVIEW AND-AUDIT l ' I TECHNICAL CHANGE - LESS RESTRICTIVE 1 (continued) LA.3 This comment number is not used for this station. ) LA.4 Review and Audit requirements are relocated to the USAR and procedures. Changes to the procedures and the USAR are I controlled in accordance with 10 CFR 50.59. i i
" Specific" L.1 This comment number is not used for this station.
L.2 This comment number is not used for this station. L.3 This comment number is not used for this station. L.4 This comment number is not used for this station. l l
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I i 1 CLINTON 9 4/15/94
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l~ l U+602203 Att. 2 Page 66 of 103 DISCUSSION OF CHANGES CTS: 6,6 - REPORTABLE EVENT ACTION t [WMINISTRATIVE A.1 This requirement is contained in Title 10 of the Code of Federal Regulations. Repeating the requirements of the regulations is redundant and unnecessary, and creates an unnecessary burden to revise the Technical Specifications when the regulations change. Therefore, this requirement is not repeated in the proposed Technical Specifications. Since the requirements remain the same, this change only affects the presentation method and is considered administrative. A.2 This comment number is not used for this station. RELOCATED SPECIFICATIONS None in this section. l TECHNICAL CHANGES - MORE RESTRICTIVE None in this section. TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic" l LA.1 Review and Audit requirements are relocated to the USAR and procedures. Changes to the procedures and the USAR are controlled in accordance with 10 CFR 50.59.
" Specific" L.1 This comment number not used for this station.
l CLINTON 10 4/15/94
__ q Ua602283 Att, 2 Page 67 or 183 j DISCUSSION OF CHANGES ; CTS: 6.7 - SAFETY LIMIT VIOLATIONS l MMINISTRATIVS i A.1 The technical content of this requirement is being moved to another chapter of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specification, NUREG-1434. Any technical changes to this requirement will be addressed with the content of the proposed chapter location. I i RELOCATED SPECIFICATIONS I None in this section, i t TECH _NJ_ CAL CHANGES - MORE RESTRICTIVE None in this section. j i i i' IECHNICAL CHANGES - LESS RESTRICTIVE None in this section. i l l l 1 a i i 1 3 CLINTON 11 4/15/94 2
U-602283 Att. 2 Page 88 of 183 DISCUSSION OF CHANGES CTS: 6.8 - PROCEDURES AND PROGRAMS ADMINISTRATIVE A.1 These types of procedures are required by the item immediately preceding which references Regulatory Guide 1.33. Therefore, it is not necessary to specifically identify each type of procedure. Since the requirements remain, this is considered to be . a change inl the method of presentation only, and J therefore, is considered an administrative change. A.2 Procedures to implement the Emergency Plan and the Security Plan are required by 10 CFR 50, Appendix E and 10 CFR 50.54 (p) . Since conformance with 10 CFR Chapter I is a license condition ', and the Emergency Plan and Security Plan are required to be implemented by 10 CFR Chapter I, specific identification of these plans is unnecessary duplication. This is a change in the presentation of the requirements only, and therefore, is considered an administrative change. 1 A.3 These specific programs are covered by a more generic item which requires this activity for all Programs and Manuals. Therefore, it is not necessary to specifically identify each ! program. Since the requirements remain, this is considered to I be a changh in the method of presentation only, and therefore, is considered an administrative change. L A.4 The technical content of several requirements are being moved l from another chapter of the current Technical Specifications l and are' proposed to be identified as Programs in accordance l with the format of the BWR Standard Technical Specification, 1 NUREG-1434. Other Programs currently identified in the Administrative Controls section are consolidated into this section. Any technical changes to the requirements are identified in their respective markups and addressed as l indicated. These Programs include: PSTS CTS LA.1 6.11 Radiation Protection Program LA.1 6.13 & 1.32 Process Control Program 5.5.1 6.14 & 1.26 Offsite Dose Calculation Manual 5.5.2 1 6.8.4.a Primary Coolant Sources Outside Cont . LA.7 6.8.4.b In Plant Radiation Monitoring 5.5.3 6.8.4.c Post Accident Sampling 5.5.4 6.8.4.f Radioactive Effluent Controls Program LA.8 6.8.4.g Radiological Environmental Monitoring 5.5.5 5.7.1 Component Cyclic or Transient Limit LA.9 4.0.5 Inservice Inspection Program LA.9 4.0.5 Inservice Testing Program 5.5.6 4.6.6.3
& 4.'.2 Ventilation Filter Testing Program 5.5.7 3/4.7.8 Explosive Gas and Storage Tank
& 3/4.7.7 Radioactive Monitoring Program 5.5.8 4.8.1.1.2.d Diesel Fuel Oil Testing Program CLINTON 12 4/15/94
i 5 U-602203 Att. 2 Page 69 of 103 DISCUSSION OF CHANGES CTS: 6.8 - PROCEDURES AND PROGRAMS 1 ADMINISTRATIVE ! , (continued) t i 1 LA.10 6.8.4.e Fire Protection Program i 5.5.9 NA Safety Function Determination Program 5.5.10 NA Technical Specification Bases Control ' 5.5.11 6.8.4.d Ultimate Heat Sink Monitoring Program A.5 This comment number is not used for this station. , 1 A.6 This comment number is not used for this station. A.7 An statement of applicability of SR 3.0.2 or SR 3.0.3 is needed to maintain the current allowances for surveillance frequency extensions since these SRs are not normally applied to frequencies identified in the Administrative. Controls section of the Technical Specifications. Since'this change maintains current requirements, it is considered a change of presentation method only. t A.8 This comment number is not used for this station. I A.9 This comment number is not used for this station. A.10 This change was previously proposed in a letter to the NRC, U- I 602243, LS-91-023 dated April 18, 1994. The No Significant Hazards Consideration for the proposed changes is still valid. Therefore, this change is considered administrative for purposes of this submittal. RELOCATED SPECIFICATIONS None in this section. TECHNICAL CHANGES - MORE RESTRICTIVE M.1 This comment number is not used for this station. M.2 Two new programs are included in the proposed Technical Specifications. These programs include: 5.5.9 Safety Function Determination Program 5.5.10 Technical Specification Bases Control The Safety Function Determination Program is included to support implementation of the support system operability characteristics of the Technical Specifications. The Bases Control program is provide to specifically delineate the appropriate methods and reviews necessary for a change to the Technical Specification Bases. CLINTON 13 4/15/94 o__ _ _
f i l U-602283 Att. 2 Page 70 of 163 DISCUSSION OF CHANGES CTS: 6.8 - PROCEDURES AND PROGRAMS l l l TECHNICAL CHANGE - MORE RESTRICTIVE j (continued) M.3 This comment number is not used for this station. l l l TECHNICAL CHANGES - LESS RESTRICTIVE l
" Generic" LA.1 Details of the methods for implementing this Specification are relocated to the USAR and procedures. The guidance documents which dictate the methods are also identified in the USAR .
Additionally,- changes to the procedures and the USAR are controlled in accordance with 10 CFR 50.59. LA.2 Details of the methods for implementing this Specification are I relocated to the USAR, procedures and the proposed Administrative Controls section of the Technical Specifications as a Program. The guidance documents which dictate the methods are also identified in the USAR. Additionally, changes to the procedures and the USAR are controlled in accordance with l 10 CFR 50.59, and the Program requirements are controlled as a ) proposed Technical Specification. LA.3 This comment number is not used for this station. LA.4 Review and Audit requirements are relocated to the USAR and procedures. Changes to the procedures and the USAR are controlled in accordance with 10 CFR 50.59. LA 5 This comment number is not used for'this station. LA.6 Requirements relating to the Process Control Program are l relocated to the USAR and procedures. Changes to the ' procedures and the USAR are controlled in accordance with 10 CFR 50.59. LA.7 Requirements relating to the In-Plant Radiation Monitoring Program are relocated to the USAR and procedures. Changes to j the procedures and.the USAR are controlled in accordance with i 10 CFR 50.59. LA.8 Requirements relating to the Radiological Environmental Monitoritig Program are relocated to the USAR and procedures. Changes 'to the procedures and the USAR are controlled in accordance with 10 CFR 50.59. LA.9 Requirements relating to the inservice inspection and testing are relocated to plant procedures. Changes to the procedures are controlled in accordance with 10 CFR 50.59 and 10 CFR 50.55a. CLIN'I'ON 14 4/15/94
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U-602283 Att. 2 Page 71 of 183 I DISCUSSION OF CHANGES l CTS: 6.8 - PROCEDURES AND PROGRAMS l LA.10 These specific fire protection issues are covered by a more l generic program (i.e., Fire Protection Program) which provides I for the definition and implementation of these details. Therefore, it is not necessary to specifically identify each of . these fire protection issues in the TSs. Changes to the fire l protection program are adequately controlled by a license condition which allows changes to the approved fire protection ] l program only if they do not adversely affect the ability to i achieve and maintain safe shutdown of the plant. I 1 TECHNICAL CHANGES - LESS RESTRICTIVE j 1 ! " Specific" l ' l L.1 This comment number is not used for this stat. ton. ' L.2 The intent of NUREG-0737 was to address " emergency operating procedures", consequently the existing TS wording was revised l appropriately to limit the scope of procedures requiring l consideration of NUREG-0737 j l L.3 This comment number is not used for this station. l l l CLINTON 15 4/15/94
U-602283 Att. 2 Pege 72 of 103 j DISCUSSION OF CHANGES 1 CTS: 6.9 - REPORTING REQUIREMENTS ADMINISTRATIVE A.1 This comment number is not used for this station. A.2 This reporting requirement is unnecessary since it generally included in the LER requirements to report fuel cladding failures that exceed expected values or that are caused by unexpected factors, i.e., being seriously degraded. Since the criteria identified in 10 CFR 50.73 have been identified as the criteria in the area of degraded boundaries that necessitates reporting, any minor differences are negligible with regard to safety. Therefore, the current reporting requirement is a duplication of the 10 CFR 50.73 reporting requirement and can be deleted. A.3 This comment number is not used for this station. A.4 This change provides additional time to obtain calendar year based analyses results which are needed for submittal of this report. Since the report frequency is unchanged from annually, this change is considered administrative. l A.5 This information duplicates the requirements of 10 CFR 50.36a l and is therefore unnecessary. Only a reference to the requirement is provided. A.6 This comment number is not used for this station. l A.7 A listing of Specifications which identify core operating limits has been added. Since this change represents a presentation preference only, it is considered administrative. A.8 This change provides additional clarification of which of the I analytical methods for determining the core operating limits that have been removed from the Technical Specifications. A.9 The initial report requirements for the Annual Report are being deleted. The initial report has been submitted, and deleting the discussion surrounding the initial report is no longer necessary. Therefore, the deletion is purely administrative in i nature. A.10 Recipients for documents sent to the NRC staff are governed by 10 CFR 5 0. 4. Therefore all references in the Specifications to NRC recipients of reports are being deleted. Since 10 CFR 50.4 l is the governing requirement, this deletion is considered administrative. l A.11 This comment number is not used for this station. l A.12 The reporting of challenges to safety and relief valves is ! revised from an Annual report to a monthly report. Since no ! CLINTON 16 4/15/94
I U-602283 Att. 2 Page 73 of 183 DISCUSSION OF CHANGES CTS: 6.9 - REPORTING REQUIREMENTS ADMINISTRATIVE (continued) change in the details of the reporting are required, this is a change in the timing only and is considered an administrative l change.
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A.13 This comment number is not used for this station, l 1 RELOCATED SPECIFICATIONS None in this section. l TECHNICAL CHANGES - MORE RESTRICTIVE M.1 This change details the information to be included in the report. These details are necessary to assure the reports are provided with similar content and format for comparison with j other plants and with prior reports, i M.2 A new report is required in conjunction with the changes ) described in Section 3.4 for the reactor coolant system ' pressure and temperature limits. In addition, requirements are included for methods used to determine such limits and for submitting the report to the NRC. TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic" ;
LA.1 This comment number is not used for this station. l LA.2 Requirements relating to the Process Control Program are relocated to the USAR and procedures. Changes to the procedures and the USAR are controlled in accordance with 10 CFR 50.59. LA.3 Requirements for Special Reports and Startup Reports are relocated to the USAR and procedures. Changes to the l procedures and the USAR are controlled in accordance with 10 CFR 50.59.
" Specific" L.1 This comment number is not used for this station.
l L.2 This comment number is not used for this station. l CLINTON 17 4/15/94
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U-602283 Att. 2 Page 7 11 of 103 DISCUSSION OF CHANGES CTS: 6.10 - RECORD RETENTION ADMINISTRATIVE i A.1 This comment number is not used for this station. A.2 This comment number is not used for this station.- A.3 This comment number is not used for this station. A.4 This comment number is not used for this station. 1 A.5 This comment number is not used for this station. J l A.6 This comment number is not used for this station. I j RELOCATED SPECIFICATIONS None in this section. TECHNICAL CHANGES - MORE RESTRICTIVE l l 1 M.1 This comment number is not used for this station. M.2 This comment number is not used for this station. TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic" LA.1 Record Retention requirements are relocated to the USAR and procedures. Changes to the procedures and the USAR are controlled in accordance with 10 CFR 50.59.
" Specific" None in this section.
l l f I' l CLINTON 18 4/15/94
l l I U-602703 Att. 2 Page 75 of 183 DISCUSSION OF CHANGES ! CTS: 6.11 - RADIATION PROTECTION PROGRAM I ADMINISTRATIVE None in this section. i i RELOCATED SPECIFICATIONS None in this section. l l TECHNICAL CRANGES - MORE RESTRICTIVE None in this section.
)
1 TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic"
! LA.1 Radiation Protection Program requirements are relocated to the USAR and procedures. Changes to the procedures and the USAR are controlled in accordance with 10 CFR 50.59.
" Specific" l
l None in this section. l CLINTON 19 4/15/94
U-602283 Att. 2 Page 76 of 103 DISCUSSION OF CHANGES CTS: 6.12 - HIGH RADIATION AREA ADMINISTRATIVE ; j A.1 This comment number is not used for this station. A.2 This comment number is not used for this station. , RELOCATED SPECIFICATIONS None in this section. TECHNICAL CRANGES - MORE RESTRICTIVE None in this section. TECHNICAL CHANGES - LESS RESTRICTIVE " Generic" LA.1 This comment number is not used for this station. LA.2 High Radiation Area control requirements are relocated to the USAR and procedures. Changes to these controls can only be made with pre approval of the NRC consistent with paragraph 20.203 of 10 CFR 20. " Specific" L.1 This comment number is not used for this station. I I i i l l i CLINTON 20 4/15/94 J
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l U-002283 Att. 2 Page 77 or pg3 I DISCUSSION OF CHANGES CTS: 6.13 - PROCESS CONTROL PROGRAM ADMINISTRATIVE A.1 This comment number is not used for this station. I A.2 This comment number is not used for this station. i A.3 This comment number is not used for this station. l l A.4 This comment number is not used for this-station. RELOCATED SPECIFICATIONS I None in this section. TECHNICAL CHANGES - MORE RESTRICTIVE None in this section. TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic" l
LA.1 Process Control Program requirements are relocated to the USAR and procedures. Changes to the procedures and the USAR are l controlled in accordance with 10 CFR 50.59.
" Specific" None in this section.
l l I 1 l I CLINTON 21 4/15/94
f U-6022B3 Att. 2 l Page 70 or 183 DISCUSSION OF CHANGES ; CTS: 6.14 - OFFSITE DOSE CALCULATION MANUAL ! ADMINISTRATIVE A.1 The applicable Specifications provide the requirements without these additional cross references. Therefore, the reference 1 to the Specifications serve no functional purpose, and their l removal is purely an administrative dif ference in presentation. I A.2 Where possible, plant specific management position titles in the current Technical Specifications are replaced with generic i titles as provided in ANSI /ANS 3.1. Personnel who fulfill
- these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details relating to the plant specific management position titles are identified in licensee controlled documents (such as the USAR) .
l The two major specific replacements are the generic " Plant Manager" for the manager level individual responsible for the overall safe operation of the plant and the generic descriptive use of "the corporate executive responsible for overall plant nuclear safety" in place of the Vice President position. The plant specific titles fulfilling the duties of these generic positions will continue to be defined, established, documented and updated in a plant controlled document with specific regulatory review requirements for changes, such as the USAR or QAM. This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation of the corporate and unit organization charts to licensee controlled documents. The intent of the Generic Letter, and of this proposed change, is to reduce the unnecessary burden on NRC and licensee resources being used to process changes due solely to personnel titles changes during reorganizations. Since this change does not eliminate any of the qualifications, responsibilities or requirements for these personnel or the positions, the change is considered to be a change in presentation only and is therefore administrative. A.3 This comment number is not used for this station. A.4 This comment number is not used for this station. A.5 This comment number is not used for this station. RELOCATED SPECIFICATIONS l None in this section. TECHNICAL CHANGES - MORE RESTRICTIVE None in this section. CLINTON 22 4/15/94
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1 ' I j U-602203 Att. 2 Page 70 or 103 . 'I I 1 DISCUSSION OF CHANGES I CTS: 6.14 - OFFSITE DOSE CALCULATION MANUAL I TECHNICAL CHANGES - LESS RESTRICTIVE
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" Generic" LA.1 Review and Audit requirements are relocated to the USAR and l j procedures. Changes to the procedures and the USAR are l controlled in accordance with'10 CFR 50.59.
LA.2 Requirements relating to the Radiological Environmental i l Monitoring Program and the Process Control Program are l } relocated to the USAR and procedures. Changes to the j procedures and the USAR are controlled in accordance with l l 10 CFR 50.59. l i
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LA.3 This comment number is not used for this station. f " Specific" i None in this section. i i j 1 l 1 J a l } CLINTON 23 4/15/94
i C"l'/";,' hs"M) - l ! l i t ATTACHMENT 1C ! } .
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4 l 1 CTS - PSTS t ! COMPARISON DOCUMENT NO SIGNIFICANT HAZARDS CONSIDERATIONS l i l l - . I O g B
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I U-602283 Att. 2 , j Page B1 of 103 l 4 l ! NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 6.2 - ORGANIZATION 4 "L1" CHANGE ] Illinois Power Company has evaluated this proposed Technical l Specification change and has determined that it involves no i significant hazards consideration. This determination has been i performed in accordance with the criteria set forth in 10 CFR 50.92. 1 The following evaluation is provided for the three categories of the significant hazards consideration standards: 1
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The addition of the note would permit an SRO to perform the
- functions of an STA as long as he had the knowledge and i training to fulfill this position. This is in keeping with the orl.ginal intent of the requirement for an STA. The original ;
intent was to have a temporary engineering person on shift ' until the licensed shift. personnel could acquire the needed training in event evaluation. The STA is not an initiating condition for any previously evaluated accident, nor is this person required to reduce the consequences of any accident, } since the required knowledge will still be required to be i possessed by someone present on shift. Therefore, there is no increase in the probability or consequences of an accident. ) 2. Does the change create the possibility of a new or different
- kind of accident from any accident previously evaluated?
1 The proposed change introduces no new mode of plant operation nor does it require physical modification to the plant, Therefore, the change does not create the possibility of a new
- or different kind of ' accident from any accident previously
. evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
j This proposed change does not ef fect the margin of safety. The ~ purpose of the STA will still be fulfilled by a qualified individual. Therefore, the change does not involve a significant reduction in a margin of safety. I 4 CLINTON 1 4/15/94
U-6022as agg, y Page og or gg3 1 i NO SIGNIFICANT HAZARDS CONSIDERATION CTS: 6.3 - UNIT STAFF QUALIFICATIONS "L1" CHANGE Illinois Power Company has evaluated this proposed Technical Specification change and has determined that it involves no significant hazards consideration. This determination has been j performed in accordance with the criteria set forth in 10 CFR 50.92. i The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the i
probability or consequences of an accident previously evaluated? The proposed change would remove the requirement for the Director-Plant Operations (DPO) to have held an SRO License. l However, since the Assistant Director-Plant Operations must I still hold an SRO License, and he is a direct report to the DPO, there are adequate qualifications being maintained on staff. These SRO License requirements are not considered as initiators for any previously evaluated accident and are not required for the mitigation of any evaluated accident. Therefore, the proposed change will not increase the probability or consequences of any accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change introduces no new mode of plant operation and it does not involve physical modification to the plant. l Therefore it does not create the possibility of a new or different kind of accident from any accident previously evaluated. l l 3. Does this change involve a significant reduction in a margin of safety? This change does not involve a significant reduction in a margin of safety since the proposed change will continue to provide for adequate staff with the required qualifications. l CLINTON 2 4/15/94 l , _ - . _ . _ . . - . - . , . . - - _ . . . , . - - - - - - = - - - - - - -
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l U-602283 Att. 2 j Page 83 of 103 ) i NO SIGNIFICANT HAZARDS CONSIDERATIONS ) CTS: 6.8 - PROCEDURES AND PROGRAMS "L2" CHANGE Illinois Power Company has evaluated this proposed Technical , i Specification change and has determined that it involves no ' significant hazards consideration. This determination has been l performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the i significant hazards consideration standards: l l
- 1. Does the change involve a significant increase in the l probability or consequences of an accident previously evaluated?
i
- The proposed change would limit the scope of procedures i requiring consideration of NUREG-0737 to emergency operating {
procedures. Even though NUREG-0737 also covers shif t turnover, use of overtime, control room access, etc., this is considered reasonable since the intent of NUREG-0737 was to address
" emergency operating procedures". These procedures are not considered as initiators for any previously evaluated accident i and although they may be required for the mitigation of an evaluated accident, this change is not proposing revisions to these procedures. Therefore, the proposed change will not increase the probability or consequences of any accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change introduces no new mode of plant operation and it does not involve physical modification to the plant. Therefore it does not create the possibility of a new or different kind of accident from any accident previously , evaluated. ! 1
- 3. ~ Does this change involve a significant reduction in a margin of l safety?
This change does not involve a significant reduction in a , margin of safety since the proposed change will continue to l provide for adequate emergency operating procedures. 1 l h CLINTON 3 4/15/94
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u-co22e3 Att. 2 . y , p. ' , . 96 Page et of 103 3 ATTACHMENT 2 1 ITS - PSTS COMPARISON DOCUMENT 1 2A: MARKUP OF ITS 2B: DISCUSSION OF CHANGES i i 9 k .) 1 i
a - ge so os g 1 ATTACHMENT 2A l l ITS - PSTS COMPARISON DOCUMENT MARKUP OF ITS l @ j l l l
u-602283 Att. 2 Page 80 or 183 Responsibility S.1 5.0 ADMINISTRATIVE C0llTROLS 5.1 Responsibility 5.1.1 The b ant S p N uen[hallberesponsibleforoverallunit operation and shall delegate in writing the succession to this responsibility during his absence. The flant Superintend t], or his esignee, i P33 " an oved ad nistrati e procedur , shall ap ove,accordancewith] prior /
' plement ion, eac proposed t st or exper' ment and pro osed changes nd modiff ations'to it systems r equipment hat affect nuclea safety.j 5.1.2 h ThhShift Supervisor (SS) ball be responsible for the control goom comand function.O ..an.pms d dirativ; = :m: eff= t Ol w sigace ey tne :nignest !=1 ef c per:te e- et-ce-ent]W,/ *Ocn J hell be issued enngolly t: 011 ~;t0 tion scr40nne4,fTuring any h aDsence MODE 1, of 2, theat5Sffrom the control or 3, an individual with -: vali room while the unit is in GM
" Senio _ on Operator me the control
; h room comma nd function.(SRO)During anylicense absence of shall th SSMromthebe designate control room while the unit is in MODE 4 or 5, an individual with Ch a -; valid SR0 license or Reactor Operator license shall be designated to assume the control room command function.
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l I U-602283 Att. 2 , page of of 183 aniza on 5.2 1 I 5.0 ADMIT 11STRATIVE C0f4TROLS 5.2 Organization 5.2.1 Onsite and Offsite Oroanizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities i affecting safety of the nuclear power plant. j
- a. Lines of authority, responsibility, and communication shall I be defined and established throughout highest management levels, intennediate levels, and all operating organization positions. These relationships shall be documented and
- updated, as appropriate, in organization charts, functional ,
descriptions of departmental responsibilities and I I relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These l
@1 requirements shall be documented in the h b. Th6611 ant Mr dcat]Nhall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and ,
maintenance of the plant; l g c. specified corporate executive positiaiif hall have
,.urporate responsibility for overall plant nuclear safety i
and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical sucoort to the plant to ensure nuclear x.ciQed docurnCor OPIO .The, safetyand < h a ll engogcde, executNe. i n V+.e,U SAR g ,N d. The individuals who train the operating statt, carry out glg% 3ealth physicM or perfonn qu li report to tne appropriate ons$ ty assurance ite manager; however,functions these may pro %on individuals shall have sufficient organizational freedom to ensure their independence from ( perating pressures. 5.2.2 Unit Staff p'inclVdd 'the @lM ng The unit staff organization shall 3 be r fe!bs:
- a. bch cii-duty shif t shcil bc cc.;.pmd vi us icon se min.mu ch"t = = pritier th r " nb : :.:.: r. A non-licered w ope =*er &ll be. on M, when he.
re e- on andste anNhilt, addiHond
+hc. Unit an-hcense ht is in +for) is ,#ctarw shall .
ggggg (continued) BWR/6 STS 5.0-2 Rev. O, 09/28/92 l l
U-602233 Att. 2
- Organization.
Page se ,, ,,, 5.2 5.2 Organization 5.2.2 Unit Staff (continued)
- b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In
--IM addition, while the unit is in MODE 1, 2, or 3, at least one 3b m licensed Senior Reactor Operator (SRO) shall be present in 4 the control room.
gj AY [:eelth I'hpic echnicianbhallbeonsitewhenfuelis
. don in the reactor, e position may be vacant for not more protte'@ than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required h'; - position.
-d . Either : licc.3cu 3RG vr licenseu 3RG iimited Lu fuci-q" .. h a dlia; -Se Sn nc ::::;. rent ,capensit;;;isca duiing this-epereuen shall bc prescat during fuci handlin3 ond-shall-di.ocLi, supe..;sc all CORE M.T" RATIONS.
- e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform l safety related functions (e.g., licensed SR0s, licensed R0s, l
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--hecith personnel;.h)"riai-+r,
. f@diah auxiliary prowcRon operators.
technic.ionQ and key maintenance i Adequate shift coverage shall be maintained without routine i 31 heavy use of overtime. -The ;bj::tivc shell b; to hav e ! operating-pencnnel '!cr' 29,[8 er -12] hour day, acmia21 2 . f-
- 40 houi week, while the Unit is ep; rating.* However, in the
/ event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of I
shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines l shall be followed: l 1. An individual should not be permitted to work more than l 16 hours straight, excluding shift turnover time; l 2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours I in any 48 hour period, nor more than 72 hours in any 7 day period, all excluding shif t turnover time; l (continued) l BWR/6 STS 5.0-3 Rev. O, 09/28/92
i U-602283 Att. 2 Page 80 or 183 INSERT 3A
- c. Shift crew composition may be one less than the minimum requirements of 10 CFR 50.54 (m) (2) (i) and Specification 5.2.2.a for a' period of time not to exceed 2 hours to accommodate unexpected absence of on-duty shift crew members, provided )
immediate action is taken to restore the shif t crew composition within the minimum requirements. 1 l l I l l l ! l i I I 1 1 I l I j INSERT CLINTON 5.0-3 4/15/94
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["(([,;[ Organizat 5.2 Organization 5.2.2 Unit Staff (continued)
- 3. A break of at least 8 hours should be allowed between work periods, including shift turnover time;
- 4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff ag@a shift.
marmAe 8l Any deviation from the a ve guidelines shall be authorized ' f i c&cnce by theAlant SupcrintendentFor his designee, in j accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. rmni ( Controls shall be included in the procedures such that l individual overtime shall be reviewed monthly by the Superintendent}"or his designee to ensure that nveexces$lant hours have not been assigned. Routine deviatior from the i above guidelines is not authorized.
--The cmount of cycrtiac workcd-- by unit :tcf f incmben. ~ g i y pcrfs-. m:ng sofety rcicted functions 3l.all be 1-;mited cad l g mttePred ;u accordartec-with --thtHfRC Fuiicy 5tatement un u
working hour; (Ccacric -Letter 32-12)T _) g;pont \ f. The -[0pcretions Macgcr cr 1--%t2nt Occrat'nn " naccr]+ ~ ygt . Ys all hold an SR0 license (b Qinton Powet %'an] Plafd g. The Shift Technical Advisor (STA) shall provide advisory k g on5 technical support to the Shift Supervisor -(SS) in the areas of thennal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.L
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Q I'n 3 cx&O; Won, % sTA %\\ eneet de quah%Rons spedGeb b M. Commisdon Tbi;ce %wernent on angineey nj expe<usej. (on snf+. BWR/6 STS 5.0-4 Rev. O, 09/28/92
U-602203 Att. 2 Page 01 of 183 gg l , N_ - W Table 5.2.2-1 (page 1 of 1) Minimum Shift Crew Composition ( ) [ Single Unit Facility]
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PO ON(b) MINIMUM CREW NUMBER UNIT IN MODE 1, 2, OR 3 UNIT IN E 4 OR 5 SS 1 1 SRO 1 None R0 2 1 2 1 A0 STA (c) 1 None (a) The shift crew composition may b e less than the minimum requirements of Table 5.2.2-1 for not more an ours to accommodate unexpected absences of on-duty shift cr members rovided immediate action is taken to restore the shift rew composit n to within the minimum requirements of Table 5. 2-1. This provi 'on does not permit any shift crew position to be un nned upon shift chan due to an oncoming shift i crewman being late o absent. (b) Table Notation: SS - [Shif Supervisor] with a Senior Reactor Operato license; { SR0 - In 'vidual with a Senior Reactor Operator license; i R0 - dividual with a Reactor Operator license; A0 - Auxiliary Operator; ST - Shift Technical Advisor. , (c) he STA position may be filled by an on-shift SS or SR0 provided th , individual meets the Commission Policy Statement on Engineering i Expertise on Shift. I 1 i l l BWR/6 STS 5.0-5 Rev. O, 09/28/92 l l
U-602283 Att. 2 b Page 92 of 103
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f.- _ Table 5.2.2-1 (page 1 of 1) 1 { Minimum Shift Crew Composition (4) (Two Units With a Comon Control Room) (Totals for Both Units) A P 10N(b) HIN! HUM CREW NUMBER ONE UNIT IN H00E 1, 2, OR 3. AND ONE EACH UNIT IN H00E UNIT IN H0DE 4 OR 5 p H UNIT IN H00E 1, 2. OR 3 OR DEFUELED OR S OR DEFUELED SS 1 1 1 SRO 1 1 None RO 3 3 2 3 3 A0(C) STA 1 None (a) The shift crew composition may be one les than the minimum requirements of Table 5.2.2-1 for not more than 2 hours o comodate unexpected absence of on-duty shift crew members, provided imediat action s taken to restore the shift crew composition to within the minimum r irements Table 5.2.2-1. This provision does not pennit any shif t crew positio o be unmanne pon shift change due to an oncoming
- shift crewman being late or abse .
(b) Table Notation: i SS -[ShiftSupervisor) ith a Senior Reactor Operator 1 ense for each unit whose reactor contains uel. SRO - Individual wit a Senior Reactor Operator license for each unit whose reactor contains fue . Otherwise, provide an individual for each unit who holds a Senior Rea or Operator license for the unit assigned. Du ng CORE ALTERATIONS on eithe unit at least one licensed SRO or licensed SR0 lim ed to fuel handli , who has no other concurrent responsibilities, must bg present. RO - Indi ' dual with a Reactor Operator license or a Senior Reactor 0 erator license for unit assigned. At least one RO shall be assigned to each uni hose reactor e tains fuel and one RO shall be assigned as relief operator for u t(s) in 00E 1, 2, or 3. Individuals acting as relief operators shall hold a license for both units. Otherwise, for each unit, provide a relief operator w holds a license for the unit assigned. 0- At least one auxiliary operator shall be assigned to each unit whose reac r contains fuel. l STA - Shift Technical Advisor. (c) The STA position may be filled by an on-shif t SS or SRO provided the individual meets the Conraission policy Statement on Engineering Expertise on Shif t. (/- - I BWh/6 STS 5.0-6 Rev. O, 09/28/92 1
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U-602283 Att. 2 Page 93 oP 183 ii 9 usi i (.d L i on M f (- Table 5.2.2-1 (page 1 of 1) MinimumShiftCrewComposition(a) [Two Units With Two Control Rooms] (Numbers for Each Unit) POS N(b) MINIMUM CREW NUMBER UNIT IN HODE UNIT IN MODE 1, 2, OR 3; / NIT IN HODE 1, 2, OR 3; UNIT IN HODE OTHER UNIT 4 OR 5; OTHER OTHER UNIT IN 4 OR 5; OTHER IN HODE UNIT IN MODE MODE UNIT IN HODE 4 OR 5 4 OR S OR 1, 2, OR 3 1, 2, OR 3 DEFUE D DEFUELED SS (d) 1(d) 1(d) 1(d) SR0 1 None 1 None R0 2 1 2 1 2 'l(e) 2 A0 STA (c) 2(d) 1 None 1 None
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(a) The shift crew composition-may be ne less than the minimum recuirements of Table 5.2.2-1 for not more th hours in order to accommocate unexpected absence of on-duty ift ew members provided immediate action is taken to restore t shift ew composition to within the minimum requirements of Tab e 5.2.2-1. This )rovision does not pennit any shift crew position t be unmanned u n s11ft change due to an oncoming shift crewman ing late or absen (b) Table Notation: SS - [ Shift Sup visor] with a Senior Reactor 0 rator license; SR0 - Individu with a Senior Reactor Operator 11 ense; RO - Indivi al with a Reactor Operator license; AO - Auxil' ry Operator; STA - Shi Technical Advisor. (c) The S position may be filled by an on-shift SS or SR0 pr ided the ind' idual meets the Commission Policy Statement on Enginee 'ng. Ex ertise on Shift. (d) ndividual may fill the same position on the other unit if licen d for both. e) One of the two required individuals may fill the same position on the other unit. BWR/6 STS 5.0-7 Rev. O, 09/28/92
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U 002283 Att. 2 Page Qat or 103 Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications Y _N Reviewer's N >: Minimum quali" cations for mem s of the unit st shall be specif* by use of an ov all qualificati statement refere ing an ANSI Stand acceptable to th iRC staff or b pecifying indivi l position j qua 'fications. Gener y, the first od is preferabl however, the cond method is ad able to those it staffs requir' special i ments because unique organizat' nal structures, j (,~ qualification st _ j l 1 l 5.3.1 Each member ofunit tip[5Nsrl AgE3.lmeeM78)t or exceed the minimum staii snais qualifications ofy[ egulato y Guide 1. , Revision , 1987, r mo ! recent revi ions, r ANSI tandard ac ptable t the NRC aff] ! [The aff t cov red by Regulatory Guide 1.8 shall me t or l texc d th mini m qual ications [Regula ans, Re taff]. , gInatory , p JGuides, or ANS Standards acceptab e to NRC add ion, the hif Techni 1 Advipbr shalVmeet tye qualifi a; ions ! s cifi by th Comis/ ion Polfty Sta)4 ment on gineer' g xpert e on ift.f J l h 4 l BWR/6 STS 5.0-8 Rev. O, 09/28/92
U 602283 Att. 2 Page 95 of 183 , _, _ u....3
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5.0 ADMINISTRATIVE CONTROLS b5.4 Train 5.4.1 A retr ning an replacement aining program r the unit staff / shal be maint ined under t direction of t [positiontitle. and shall me or exceed e requirements d recommendatio of S tion [ of [an ANSI tandard accept e to the NRC s ff] and 0 CFR 55 and, for ap opriate design ed positions, s 1 ; include miliarizat' n with relevan industry operat' nal experie ce. L i i i t
!1 BWR/6 STS 5.0-9 Rev. O, 09/28/92
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u-602283 Att. 2 Page 96 of 183
- RC V i T.!: ' " d A u d i ( ', _
EE 5.0 ADMINISTRATIVE CONTROLS [ 5.5 Reviews end Audits f
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l eviewer's Hote: The licensee shall describe the method (s) establi. ed to c duct independent reviews and audits. The methods may take a ra e of forms ac table to the NRC. These methods'may include creating an or nizational unit r a standing or ad hoc committee, or assigning individual capable of conduc ing these reviews and audits. When an individual perf s a review functio a cross disciplinary review determination is neces ry. If deemed necessary such reviews shall be performed by the review p sonnel of the appropriat discipline. Individual reviewers shall not r view their own work. l Regardless o the method used, the licensee shall speci the functions, I organizationa arrangement, responsibilities, appropri te ANSI /ANS 3.1-1981 qualifications, nd reporting requirements of each f ctional element or unit that contributes these processes. l Reviews and audits o activities affecting plan safety have two distinct elements. The first e ment is the reviews p fonned by plant staff personnel to ensure that day to da. activities are con cted in a safe manner. These reviews are described ina ction 5.5.1. T second element, described in Section 5.5.2, is the '[ offs'te] reviews a d audits of unit activities and programs affecting nuclear s ety that e perfonned independent of the )lant staff. The [offsite) reviews d audi s should provide integration of tie reviews and audits into a cohesi e p gram that provides senior level utility management with an assessment of ility operation and recommends actions to improve nuclear safety and plant r iability. It should include an assessment _of the effectiveness of reviews nd ted according to Section 5.5.1. _ 5.5.1 Plant Reviews l Reviewer's N e: The licensee s 11 describe provisions for plant reviews (or anization, reporting, cords)andtheappropriate _ ANSI /ANS andard for personnel qual' ication. _ 5.5.1.1 Functi s i The plant review method specified in Speci ' cation 5.5.1] shall, a a minimum, incorporate functions that:
- a. Advise the [ Plant Superintendent] on all mat ers related to nuclear safety; l
l (contin d) BWR/6 STS , 5.0-10 Rev. O, 09/28/92
U-002283 Att. 2 4 h Edvidn5 anu Auuj ( - ,
)
\5.5 Reviews and Audits l 5.5 .1 Functions (continued) _.
! b. Recommend to the [ Plant Superintendent] approval disapproval of items considered under Specificat'ans !
1 5.5.1.2.a through 5.5.1.2.e prior to their impi mentation, except as provided in Specification 5.7.1.3;
! c. Determine whether each item considered und Specifications l 5.5.1.2.a through 5.5.1.2.d constitutes unreviewed safety ;
uestion as defined in 10 CFR 50.59; an ! I d. No ify the [Vice President-Nuclear perations] of any j saf y significant disagreement b ween the [ review ] edinSpecification5.5.1] organ and th zation or individual sp]eciithin 24 hours. However, [ Plant Superintendent the [Pla t Superintendent] s 11 have responsibility for resolutio of such disagree ents pursuant to Specificati n 5.1.1. t 5.5.1.2 Responsibilities
- The [ plant review met d pecified in, Specification 5.5.1] shall l
- be used to conduct, as minimum, reviews of the following
4 a. All proposed pr cedu s required by Specification 5.7.1.1
. and changes t reto;
. b. All propos programs re ired by Specification 5.7.2 and changes t ereto; i c. All pr osed changes and modi ications to unit systems or equi ent that affect nuclear fety; a
- d. Al proposed tests and experiment. that affect nuclear fety; and
- e. All proposed changes to these Technica Specifications (TS),
their Bases, and the Operating License. 1 4 , (continued)j BWR/6 STS 5.0-11 Rev. O, 09/28/92 1
U-602283 Att. 2 Page 98 of 183 - Dguig..g 3 g ag -;;;
-G-G f 5.5 Reviews and Audits (continued) 5 2 [0ffsitel Review and Audit -
Reviewer's Note: The licensee shall describe the prov'sions for reviews and audits independent of the plant's staff rganization, reporting, and records) and the appropriate ANSI /AN Standards for ersonnel qualifications. These individuals may e located onsite offsite provided organizational independence rom plant staff is aintained. The [ technical] review respons' ilities, Spec'fication 5.5.2.4, shall include several ndividuals located onsit 5.5.2.1 Function The [offsit review and audit provisi s specified in Specification 5.5.2] shall, as a min'.um, incorporate the following func 'ons that:
- a. Advise the ice President (uclear Operations] on all matters relat to nuclear safety;
- b. Advise the manag ent o the audited organization, and [its Corporate Manageme t d Vice President-Nuclear Operations],
of the audit result as they relate to nuclear safety;
- c. Recommend to the nag ent of the audited organization, and its management, ny corr ctive action to improve nuclear safety and pla operatio and
- d. Notify the ice President- clear Operations] of any safety significan disagreement betwe n the [ review organization or individua specified in Specifi tion 5.5.2] and the
[organi tion or function being r viewed] within 24 hours. 5.5.2.2 [0ffsite] eview Responsibilities The [re iew method specified in Specificatio 5.5.2] shall be respo ible for the review of:
- a. The safety evaluations for changes to proce res, equipment, or systems, and tests or experiments complett under the provisions of 10 CFR 50.59, to verify that suc actions do not constitute an unreviewed safety question as efined in 10 CFR 50.59; (cont'nued)
BWR/6 STS 5.0-12 Rev. O, 09/28/92
U-002283 Att. 2 Page gg or gg3 _Dneing 'A laj;; M 5.5 Reviews and Audits /
\
r . fh . 2 [0ffsite] Review Responsibilities (continued) _
- b. Proposed changes to procedures, equipment, or syfs ems that involve an unreviewed safety question as define in 10 CFR 50.59;
. Proposed tests or experiments that involve n unreviewed safety question as defined in 10 CFR 50.59, j
- d. roposed changes to TS and the Operatin License;
- e. Vi ations of codes, regulations, or rs, license requ ements, and internal procedur s or instructions having nuclea safety significance;
- f. All Licen e Event Reports requ' ed by 10 CFR 50.73;
- g. Plant staff rformance;
- h. Indications of anticipat deficiencies in any aspect of design or operati of st uctures, systems, or components that could affect cle safety; .
- i. Significant accidenta unplanned, or uncontrolled radioactive release , i cluding corrective action to prevent recurrence; I
- j. Significant ope ating abno alities or deviations from nonnal and ex cted performa ce of equipment that affect nuclear safe y; and
- k. The perf ance of the correctiv action system. I Reports or ecords of these reviews shal be forwarded to the
[Vice Pr ident-Nuclear Operations] with 30 days following complet'on of the review. I l 5.5.2.3 Aud" Responsibilities ( e audit responsibilities shall encompass:
- a. The conformance of unit operation to provision contained within the TS and applicable license conditions,
- b. The training and qualifications of the unit staff; (conQnued)'
L . BWR/6 STS 5.0-13 Rev. O, 09/28/92
l Att. 2 g,- Page 100 of 183
'N'i'na
,_J unu
,,J, _)
rm u i s .)
--G,5- {
, , s :
y.5 Reviews and Audits / 3 5.5. 3 Audit Responsibilities (continued) _ I
- c. The implementation of all programs required by Specification 5.7.2; Actions taken to correct deficiencies occurring equipment, structures, systems, components, or ethod of operation that affect nuclear safety; and
- e. Ot r activities and documents as requested y the [Vice l Pres' dent-Nuclear Operations] .
, Reports or r ords of these audits shall be orwarded to the [Vice l President-Hu ear Operations] within 30 d s following completion l of the review. ; 5.5.2.4 [ Technical] Review esponsibilities The- [ technical] review esponsibili es shall encompass:
- a. Plant operating char teris 'cs, NRC issuances, industry 1 advisories, Licensee ent eports, and other sources that I may indicate areas for roving plant. safety;
- b. Plant operations, modi ica ions, maintenance, and ;
surveillance to veri inde ndently that these activities ' are perfonned safe and corr tly and that human errors are reduced as much a practical;
- c. Internal and e ernal operational xperience infonnation that may ind ate areas for improvi plant safety; and
- d. Making de iled recomendations throu the [Vice Presiden -Nuclear Operations] for revi ing ^ procedures, equipm t modifications, or other'means improving nuclear safe and plant reliability.
5.5.3 Recor ' Wr'tten records of reviews and audits shc11 be maintal ed. As a nimum these records shall include:
- a. Results of the activities conducted under the provis ons of Section 5.5; (conti E <
ted)) BWR/6 STS 5.0-14 Rev. O, 09/28/92
U-602283 Att, 2
" * " '*' *f 'e3 I C2.0 -
ne v i c n ., umm ,mm, m 4-r '
]
'5 eviews and Audits 5.5.3 _(ords (continued) _.
- b. Rec endations to the management the organization being audite
- c. An assessment the s ty significance of the review or audit findings;
- d. Recommende pproval or dis royal of items considered j under ci fications 5.5.1.2.a ugh 5.5.1.2.e; and j
- e. termination whether each item conside under Specifications 5.5.1.2.a through 5.5.1.2.d c titutes an unreviewed safety question as defined in 10 CFR 59. i
, l
^
^
J i i l I a BWR/6 STS 5.0-15 Rev. O, 09/28/92
l . I U-602203 Att. 2 1 Psoe 102 o9 183 -TS E2;c; S; ,t74 i cl (move 70 5,g,go) F*jvoWadhg:.c 5 fo3GfT15 aniM411vd$) 5.0 ADMINISTRATIVE _ CONTROLS (a
^
A+ fr (Technical Specifications (TS) Bases control Pecorah s,s,lo i - j 3 S.C.1- h Changes to the Bases of the TJ Qall m e u der appropriate j ' ' trative_contro Specificatier T.S.E. p d cc,;cocd au m dir; t: l
- W h Licensees may make changes to Bases without prior NRC approval
! provided the changes do not involve either of the following: A change in the TS incorporate in the license; or USAR
- 2. )'. A change to th ::pd:ted I';AR r Bases that involves an j unreviewed safet es o as defined in 10 CFR 50.59.
j M .0.1 hC. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with th AR.
"'?
i Pet i$cdon T.S. lD h p- 5.5.4 Proposed changes that meet tne criteria off} cr N.above shall a be reviewed and approved by the NRC prior to impiementation.
- Changes to the Bases implemented without prior NRC approval shall j' f be provided he NRC on a frequency consistent with-i 10 CFR 50.7 (e),
i D I This program changes % %provJes Bases erf %es ct TechnicoA rneans & ptocess:n3 ( 6peciWCa.No AL , j i .l. 5 i i i i l BWR/6 STS 5.0-16 Rev. O, 09/28/92 s k
g Proceduresj programs, and "anual3-7- Uf 5.0 ADMINISTRATIVE CONTROLS U-602283 Att. 2 h>rocedure(Fiv3om, end"snuals] h7.1 Proccdur [ (17.1.1 Aupe [ (5.4.)) Written procedures shall be established, implemented, and maintained covering the following activities:
- a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;
- b. The emergency operating'procedur s required to implement he requirements of NUREG-0737 and REG-0737, Supplement lj a steicd .n [Guneric Lcttcr 02 33h C2.
-c . Security plcn %ple=ntetien;
- d. Dr.crgcncy plen imn'cm:ntation: .
s C p'. Quality assurance for effluent and environmental monitoring- Qnd f. Fire Protection Progrem w imple .mntetien, en"A ' ~~ 3 h , ,,g'. All programs specified in Specificatio 5.7.2.55 5.7.1.2 eview and Approval Each proce re of Specification .7.1.1, and chang thereto, shall be eviewed in accordan with Specificati 5.5.1, approved , by the lant Superintendent or his designee i accordance with approv administrative p.r edures prior to i lementation and , revi ed periodically as et forth in a.dmin' trative procedures. d .7.1.3 T porary Changes Temporary changes o procedures of S cification 5.7.1 may be ade provided:
- a. The inte of the existing rocedure is not altere ,
- b. The c ange is approved y two members of the p1 t man ement staff, at ast one of whom holds Senior R ctor Operator lic nse on the unit affect ; and j f
centinued) BWR/6 STS 5.0-17 Rev. O, 09/28/92
^^ ^ ^ ~
~ {Procedere+7 Programsp and Manuals-5 5 5-+
1
; Programs /_and Manuals ese o o ea 5.7.1.3 Temporary Chan s (continued) _
- c. The ch ge is documente and reviewed in cordance with l
Speci ication 5.5.1 a approved by the lant l Sup intendent] or h' designee in acc dance with app oved I ad nistrative proc dures within 14 da s of implemen tion, j 1 1 4.7.2 Fro m mA Manuale >
-Q The following programs sha e esta lished, implemented, and maintained. A i r5.7.2.1 i Radiation Protection Program 1 /
Y l Proc dures for per onnel r iation protec ion shall be epared co istent with e requi ments of 10 R 20 and sh be a roved, main ined, an adhered to f r all operat ns invol ing personnel rad'ation exp sure. _ C5.7.2.2
~
P ocess control Pr ram (PCP) f l l The PCP shall c ain the current ulas, sampling, analy es, < i tests, and de rminations to be mp to ensure that proce ing and packaging o solid radioactive ydstes will be accomplis d to ensure co liance with 10 CFR 0, 10 CFR 61, and 10 C 71; state regulat' ns; burial ground quirements; and other quirements gover ng the disposal of olid radioactive waste L' ensee initiated ch ges to the PCP:
- a. Shall be do ented and records of r iews perfomed shall be retain . This documentation s 11 contain:
- 1. ficient infomation to upport the change (s.) a appropriate analyses or valuations justifying e j change (s),and
- 2. a determination th the change (s) maintai the overall ,
conformance of t solidified waste pro ct to the existing requi ments of Federal, Stat , or other applicable r ulations. (continued) BWR/6 STS 5.0-18 Rev. O, 09/28/97.
_.~ . . i i i C fProcedum,ProgramsfandManuals
- l. L 55 t9 l
l u-eo22es Att. 2 l
-S.7 P. cccdu, a Programs [and Manuals e.g. m r tea 5.7.2.' # Proce Control rogram (PC ) (continue _
C3) b. Shall be effective fter review d acceptance y the 5 ( [revie' method of pecification 5.5.1] and th approval ofj the ant Super}.tendent). > 5.7.2.3 Offsite Dose Calculation Manual (0DCM) 55,I
- a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation ;
of gaseous and liquid effluent monitoring alarm and trip __. setpoints,andintheconductoftheydiological Qvironmental $nitoring gogram; and l i
- b. The 00CM shall also contain the ladioactive \ffluent
', s _ . ntr s nd iolonicallovi ronmental i
~-'
-re"ui rcd b- 5:ca fication 5' 7.2, land des
$nitoring criptions of theprograms i in onna ion fiE~sficuId d be included in the Annual
, Radiological Environmental ratingfandRadi_o ive s
n Releasefr ts . cguired Ey geciTiciti...
,_- [5.3.1.0; and spic;Iscei.wo3 5 0.1.4 .
- - -3 l Licensee initiated changes to the ODCM: j
- a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:
- 1. fi 'e 'f ation to support the change (s 0g -tegcthcr gh tb appropriate analyses or evaluations Justi ying th_e change (s), and
- 2. a determination that the change (s) maintain the levels.
of radioactive effluent control required by I 10CFR20.)106,40CFR190,10CFR50.36a,and I M- 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint ! calculations;
- b. Shall become. effective after review and accsptance by-the 4=
theicw mcthodercf Spcc;ficj; dint; tion
;gjg;g and 5.5.0 and thyapproval of
@ plant-5 (continued)
BWR/6 STS 5.0-19 Rev. O, 09/28/92 l
- - -Proccdurc;3 ProgramsfandManuals T.5T l U-502203 Att. 2 G.5 Page 106 of 183 h-7 "rc:cdurc;,ProgramsfandManual5 4 s -
5.5. l SJ.Z.3j Offsite Dose Calculation Manual (0DCM) (con +.i nued) _ 1
- c. Shall be submitted to the NRC in the form of a complete, l legible copy of the entire ODCM as a part of, or concurrent l with, the Radioactive Effluent Release Report for the period j of the report in which any change in the ODCM was made.
Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month h, and year) the change was implemented. 5.7.2.4 Primary Coolant Sources Outside Containment 1 5 5 2. This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to ( levels as low as practicable. The systems include.v[the t-ow !
@] Pre 52ui c Cerc Spray, "igh ar ;;urc Cerc , ray, Residuc' !! cat-Remevel, Reau;.cr Cere--T M ation Coolir.g, iydiv3cn s c uumb i n e r, l 3n9p t 9FGGc33- 3Gmpling, and Standby G : T reatmer.tl u The program shall l include the following:
qpA 71
- a. Preventive maintenance and periodic visual inspection )
l requirements; and
- b. Integrated leak test requirements for each system at ;
refueling cycle intervals or less. l (5.7.2.5 In Plant Rad' tion Monitorin f * ! D This pro am provides co rois to ensure he capability to y detennine t airborne iodi concentration in tal accura ! areas nder accident ondi ti ons '. Thi program shall inc de the fol wing: C,E3 . Training personnel;
- b. Proce res for monitorin ; and
- c. P visior.s for maint ance of samplin and analysis quipment. .
(continued) BWR/6 STS 5.0-20 Rev. O, 09/28/92
U-602283 Att. 2 Page 107 of 183 INSERT 20A
- a. LPCS System;
, b. HPCS System;
- c. RHR System;
- d. RCIC System;
- e. Suppression Pool Makeup System;
- f. Combustible Gas Control System;
- g. Containment Monitoring System; and
- h. Post-accident Sampling System.
I i i INSERT CLINTON 5.0-20 4/15/94
- recedures, Programsfand Manuals g s.s -s .7 r f u-no22ea att. 2 T .5 ,
Page 108 of 183 ( -5.7 hoccduren Programsfand Manuals GG
; b,h2-f Programs and Manuals (continued) _
-5 . 7 . 2 . G~ Post Accident Sampling 553 j This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:
- a. Training of personnel;
- b. Procedures for sampling and analysis; and ;
- c. Provisions for maintenance of sampling and analysis equipment.
6.S.4
, 4.7.2.7- , Radioactive Effluent Controls Program h This program confonns to 10 CFR 50.36a for the control of radioactive effluents and for maintaining' the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM; 79
- J b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricte reas,
.hten Sm ]estfc conforming to/10 CFR 20, Appendix B, Table Column 2; (ContenW#0n VMu.e6 i n f
- c. Monitoring, sampling, and analysis of radioacti@v liquid and gaseous effluents in accordance with 10 CFR 2 .Mfr nd with the methodology and parameters in the ODCM; (302.
- d. Limitations on the annual and quarterly doses or dose j commitment to a member of the public from radioactive materials in liquid effluents released from ear.h-unit to unrestricted areas, conforming to 10 CFR 50, Appendix I; M
(continued) BWR/6 STS 5.0-21 Rev. O, 09/28/92 l l
, - - ~ h rocedurecy Programsf and Manuals O s t.s++j ;
7
- 6. 6 u-earres Act. 2 5.7 Procedures # Programs 7and Manuals n.a. 109 or ies 5 5,Y Radioactive Effluent Controls Program (continued) _
(-M.Z.7J
- e. Detennination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that i appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I; Limitations on the dose rate resulting from radioactive i
,g. material released _in caseous effluent nd the l OP26 's liiegnformirig7tc fhc d~csc @mosuted "ith !
h ~2.2 TC ' I"N b I"" I' -
- h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from Ceat-h +hp unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; )
- i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from +e unit to areas beyond the site boundary, conforming to 10 CF 50, endix I; and h P7
- j. Limitations on the annual dose or dose commt ment to any member of the public due to releases of radioactivity and to radiation from uranium . fuel cycle sources,. confonning to 40 CFR 190. f 5.7.2 Radiolog' al Environ ental Monito ing Program This ogram is f monitoring he radiation d radionucli s in ,
the virons of e plant. T e program shal provide Q2. rep 'sentative easurements f radioactivit in the highe t po ntial exp oure pathway and verificati n of the acc acy of t e effluent onitoring pr gram and model ng of enviro ental exposure pathways. The p ogram shall be contained in heODC,j (continued) BWR/6 STS 5.0-22 Rev. O, 09/28/92
- -,,-, s _ - , + , we --
. - . . - . . - . -- - .~_ - - . . . . . _ . .. . .
[ U-002283 Att. 2 Page 110 of 183 INSERT 22A
- 1. For noble gases
- Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin, and i 2. For Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days: less than or equal to 1500 mrem /yr to any organ; ,
l i ? i e b 1 e 4 E l 1 f 1 s 9 i l l l 4 q 1 l 1 i i 1 i i \ INSERT CLINTON 5.0-22 4/15/94
g --' hrecedurar ; Programsfand Manuals
< L 5 5 5,7
$$ U-602283 Att. 2 5.' "eced"rer Programsrand Manual s Page til of 183 o
5.7.2.8 Radiological nvironmental onitoring Pro, am (continue )_ shall conf rm to the gui nce of 10 CFR 0, Appendix , and shall include e following:
- a. onitoring, sarr ing, analysis, and reportin of radiation and radionucl' es in the envi nment in acc rdance with the methodology nd parameters i' the ODCH;
- b. A Land U Census to ens e that chang in the use of are s at and eyond the site undary are i ntified and that modif' ations to the nitoring pro am are made if req ired by e results of t s census; and
- c. P rticipation in n Interlabora ry Comparison Pror am to nsure that in pendent check on the precision a accuracy of the measur ents of radio ctive materials in environmenta sample matri s are performed as art of the /
quality ass rance program or environmental m itoring. } g Snh2 Component Cyclic or Transient Limit
@ This program provides controls to track th "cction [ '
cyclic and transient occurrences to ens t e I maintained within the design limits. @ m c w m getis . I F- o i 5.7.2.1 Pre-St essed Concrete Containment Tendon Surveillance Pr gram Thi program prov' es controls or monito ng any tend n de adation in p e-stressed c crete con ainments, i luding of ectiveness its corrosi n protecti n medium, t ensure ntainment s uctural int . rity. Th program sh 1 include aseline mea rements pri to initi operation . The Tendon Surveillanc Program, in ection fr uencies, a d acceptance criteria all be in ac ordance wi [Regulat y Guide 1.35, Revision ,1989]. [ The pr isions of S 3.0.2 and R 3.0.3 ar applicable t the Tendo Surveillance Program inspection frequencies. 4 _ ( ---. Inservi e Inspectic Program / 5.7.2.1/ This rogram prov' des controls for nservice inspec on of ASME
/ Cod Class 1, 2, and 3 components including appli able suppor Th program sh< 1 include the fo owing:
2. e > (continued) BWR/6 STS 5.0-23 Rev. O, 09/28/92
m . ~ _ . _ .. _ __ . - - h dure:f ProgramsfandManuals
- ,@ , , V6*
U-002283 Att 2
. P.ccedur^c, Programs 7 and Man *o* II2 e sa
~ ._ a
- 5.7.2.11 Inservice In pection Program ontinued) _
- a. Pro sions that inser cc inspection ASME Code Class 1, 4
2 and 3 components all be perfo .d in accordance ith SME Boiler and Pr ssure Vessel C e and Addenda, Section XI, as r uired by 10 CF 50.55a; e frequencies 1
. The provisio of SR 3.0.2 e' applicable to ng inservice i spection activit' s;
{ for perfo
- c. An ins vice inspectio program for pip' g identified in f j Gener'c Letter 88-01 n accordance wi the NRC staff
- pos' ions on sched e, methods, per nnel, and sample
! e ansion includ in Generic Let r 88-01, or in a ordance ith alternate easures approve y the NRC staff and d Nothing in e ASME Boiler an Pressure Vesse Code shall be h construed' o supersede the requirements of any-TS. __) I5.7.2.12 Inservice Te ting Program ~ l This pro am provides ntrols for inse ice testing of E Code Class , 2, and 3 co onents includin applicable supp ts. The ,
- prog m shall incl e the following-1 a Provisions hat inservice sting of ASME Co Class 1, 2, 4 CM and 3 pu s, valves, and ubbers shall be erfonned in accorda e with Sectio I of the ASME Bo* er and Pressure Vesse Code and appli ble Addenda as r uired by 10 50.55a;
- b. sting frequenc' s specified in S tion XI of the ME oiler and Pre ure Vessel Code a d applicable A enda as follows: .,
I (continued) l sg BWR/6 STS 5.0-24 Rev. O, 09/28/92 l 4
,,,,_,,,.-,,,,____,,,___,,y. . . , , , . _ . , ,_ ,,,,, .,
l l l
'~~~'(-Precedures;ProgramsfandManuals f
< < s.s % I 65 u-no22e3 Act. 2
-S . 7 Preeedures7 Programs 7and Manual s ) Pao. 113 or ies 5.7.2.1 Inservice /stingProgram continued)
A< E Boiler and Pr ssure l essel Code and applicable Add da i terminology f r Required frequen es I inservice t ting for performing nservice ) activitiec testina activ' ies Weekly At least per 7 days At leas once per 31 days g Mont y Qua erly or every I i months At le st once per 92 d fs ! emiannually or every 6 mon s A least once per 1 days ; Every 9 mon s t least once per 6 days j Yearly or nnually At least once pe 366 days i 1 Biennial or every 2 ye s At least once er 731 days , i
- c. The ovisions of SR 3 .2 are applicabl to the above ,
I i rep ired Frequencies or performing ins vice testing ( a tivities; l 1
- d. The provisions o SR 3.0.3 are app cable to inservi l I
testing activi es; and
- e. Nothing in e ASME Boiler an ressure Vessel ,Cmde shall be construed o supersede the r uirements of any/TS.
1 cg 5@c2?lt Ventilation Filter Testing Program (VFTP) 6.5. G A program shall be established to implement the following require,d testing of Engineered Safety Feature (ESF) filter ventilation systems _ at the frequencies snecified in8Regu14tery Guide ], r-and in accordance .,ith--tRegulatory Guide 1.52, Revision 2; f,S",E-g N510 1030; andf.CIK l (Continued) BWR/6 STS 5.0-25 Rev. O, 09/28/92
l
'(frecedurxjProgramsfandManuals f #$.& E.7 l
I 1 U-602283 Att. 2 4 5.7 I'ccccdur S j Programsf'and Manuals e.g. iiii or tes I i
.a -5.7.2.13 Ventilation Filter Testing Program (VFTP) (continued) ~
l (5.54 J
- a. Demonstrate for each of the ESF systems that an inplace test i of the high efficiency particulate air QH PA) filters .shows ;
a penetration and system bypass <g[0.05f4 when tested in accordance withMRegulatory Guide 1.52, Revision 2, and -ASME ! O81 Aid N510-1 gat the system flowrate specified below) 1047 l ESF Ventilation System Flowrate i S6 76 M ch 7 OBl y contro i Remn p wnw'l.# y ,O2Dc87g
. i rrwV,eup ;N - -
,a
{ 20> .
- b. Demonstrate for each of the ESF systems that an inplace test i of the charcoal adsorber shows a penetration and system j r-\ c. M % 3 spedRed bypas?s: [0.5]*.% hen tested in accordance with,dRegulatory {
below Guide 1.52, Revision 2, and N510-19 t the system j g flowrate specified below) 10% : Aus i ESF Ventilation System Flowrate U aM Bjp255
$GTS t/a%) c4 OiOS% 1 l 4~ DNdO*'
- YeddWon C,h*P-EccocGn
-M,ad c.%
1 O, es7e 2.'%
)
- c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when btained as W
_j bE;
\/
described inJRegulatory Guide 1.52, Revision 2 methyl iodide penetration less than the value specifici. shows the g lessi nl below when tested in ac l ofe ~ temperatureMf : .[00"C]grdanceand greaterwithg[ than orASTM equal to D38034989 the Fat a ! ' N6 iRc8 relative humidity specified below: i A i Tepc . ESF Ventilation System Penetration RH eat 5 o,l'75 9. p 7o#/oj 80*c_
~
j Conhol n*cu Rw$w%n+kkiy/ , 6 33*c 1 1
- o. )?5% 769o
- E rc.) , Son g ie r 7107d Sot (continued)
BWR/6 STS 5.0-26 Rev. O, 09/28/92
1 I
~'
~
Pradrc:, Programs /'andManuals) ( 5 5 5.7 l i fee "* U-602283 At '
-6 .'7 Fraeedtwes,ProgramsfandManuals Pass iis or ie3 e
- s. 5 c.
5.7.2.13 Ventilation Filter Testing Program (VFTP) N Revi wer's N/te: Alowablepdnetratio/=[100% methyl io ide (continued) _ [m - 7 j ef ciency or cha oal cre 'ited in aff safet evaluati ]/ / afety f tor) . Safety actor - [5] for systems ith heater . ,
- [7] f systems without he ters.
_ j
- d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters /-thc prcfilters, and h the charcoal adsorbers is less than the value specified below when tested in accordance witQRegulatory Guide 1.52, ;
Revision 2, and at the system flowrate <
@ specifiedbelopi10%:
510-198 A 60 ESF Ventilation System Delta P Flowrate 0 SaTS Ljoccc(m # OS\ g Conn o rnarevpt Room r, gjeg Ven+;hk5s'(c,,,0 o"yg wc,-Q goco eg
- 1 Demonstrate that the heaters for each of the ESF systems J' '
- e. _
g - dissipate the value specified below [1 10'.When tested in OBI - accordance withjASME- Mist N510-195> & - ESF Ventilation System Wattage
$6 76 /8ORW~# ;
4 ConWo\ _, i mvcopRcom % %%{ 4 N.V xW The provisions .of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. 5.5,1
-5.7.2.14- Explosive Gas and Storage Tank Radioactivity Monitoring Program
' This program mixtures provides contained controls in the-{ for potentially Waste-Gas exp]losive
-licidup Syste gas
, [the quanH+y
'-of-radioaetivity cantained in gas staregc tank cr fed into-the --
fr #. -woffgas treatment system, and the quantity of radioactivity Cord'M contained in unprotected outdoor liquid storage tanks-f' -Me-ity quartitie P ' ' bc de t e" aed %' ' ~ . ng t h: --
@1a,tecurr2diencti (continued)
BWR/6 STS 5.0-27 Rev. O, 09/28/92
l O
~
h acedarne ProgramsfandManuals] I b l l l Q* & U-602283 Att. 2 ; 5.7 h vetuure,y Programs [and Manuals Page M6 of 183 5' 5,9 5.7.2.14 Explosive Gas and Storage Tank Radioactivity Monitoring Program i (continued) l r i method ogy in Branc Techn' al Pos' ion (BT ETSB l-5, y l
.- "Pos lated dioac ve Rel ase du to Wast Gas Sys em Leak jfr l Oc2g Fai ure"] . The li uid ra aste antitier shall b determi d in I a cordan with tandar Revie Plan, ction 1 .7.3, "Po tulated (adica ive Rel ase du to Tan Failur "] .
j The program shall include: g --frgojn condenser
\prrgas treshmen+ ggtern
- a. The limits for concentrations of hydrogen and-exygeft-in the# 1 g
s ['..'astc Ce3 Meldup Systei@ and a surveillance program to' ' g ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether i or not the s m is designed to withstand a hydrogen i explosion); T. b survei lance pro ram to ensure that the cantity 'f l radioa ivity con ined in [ fch nd fed l into e offgas reatment s stem]isgas le stor ge t. tank than amount l tha would res t in a who e body expo ure of a .5 rem l s an individua in an unr tricted ar , in the event of an g I
\ u controlle release of he tanks' ntents]; and -
l l A surveillance program to ensure that the quantity of l radioactivity contained in all outdoor liquid radwaste tanks I that are not surrounded by liners, dikes, or walls, capable l of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the g .g{biquid Radwaste %reatment bstemMs less thanphe = Ma.t would racul+ 4 a-concentratier les: than t..e ' M tr of 40 CF" 20-Appendix B, Tabic II, Celumn 2, at the-nearest- I pa+eMe ecter apply-and-thcarcst surfdcc water supply in. an. unrestricted-crec, in--the event of an uncontrc11ed - relce.sc cf the tM:r' centents. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. (continued) BWR/6 STS 5.0-28 Rev. O, 09/28/92
'~'~'(orece& ren Programsf and Manuals N E5 i -5'-+-
~ r 5.5 5.7 "recchrnj Programs /'and Manuals N u-no22e3 4ti. 2 Page 117 of 103 Programs and Manuals (continued) 55,6 (5.'.2.15- Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling-and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:
- a. Acceptability of new fuel oil for use prior to addition to storage tanks by detennining that the fuel oil has:
- 1. an API gravity or an absolute specific gravity within limits, y
- 2. a(f4=P peint =0,kinem tic viscosity within limits for
- h. h-" 3.
ASTM 20 fuel oil,Cand) atw are <4rnwr conent wWn Iirrnts oQ a clear and bright appearance witn proper color; ,, dh @
- b. Other propertie)(for ASTM 2D fuel oillare within limits within ays .v1 wing ==9 5; - '-addition to storage >
h . , tanks; an ,:3g gngg%4wg
- c. Totalparticulateconcentrationofthefueloil[iss10m I when tested every 31 days in accordance with ASTM D-227 -86,
(&cadA4crA p d h .7. 16 ire Pro ctionPrgrad [ This- ogram pr ides ntrols o ensur. that.ap ropriat .fi re maint ned to rotect e plant from fi e prot tion me ures a and o ensur the ca ability o achi e and m ntain s fe shut own C,3 i the even of a re is m intaine . BWR/6 STS 5.0-29 Rev. O, 09/28/92 t I i
,, - r , . - ,
o3fa m S cit!77b ntfoJs , 15.s e 6'I Trgrams ca nd /hnvods 50 ACL ET'" U"E CONm0Lt m . - . u-co22ea Au. 2 si$ Pfo<forn$ oQfodnuds (CoAh0Ved) Page 118 or 183
. C Safety Function Determination Program (SFDP) ;
W ( @ntry into LCO 3.0 6-GA- This program ensures loss of safety function is detected and 5.5,C}j ' . appropriate actions taken. Upon faihre te meet twe -cr ;c r LCos at-thc so;; time, an evaluation shall be made to detennine if loss h of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LC03.0.6.f h 5.0.2 The SFDP shall contain the following:
- a. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
- b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
- c. Provisions to ensure that an inoperable supported system's )
Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
- d. Other appropriate limitations and remedial or compensatory actions.
5.S.3 A loss of safety function exists when, assuming no concurrent OG) single failure, a safety function assumed in the accident analysis cannot be perfonned. .For the purpose of this prdgram, a loss of safety function may exist when a support system is inoperable, and:
- a. A required system redundant to system (s) supported bphe inoperable support system is also inoperable (C=: A,, or 6 b. A required system redundant to system (s) in turn supported by the ino
-(C;;; D){ perable or supported system is also inoperable (continued)
BWR/6 STS 5.0-30 Rev. O, 09/28/92 4
- - . . - - . - . . . . _ . ~ . _ _ . . . . , . _ - . . _ . . . , , . . . , _ . . , _ , . . ..._,__ . ,-.r _ y . _ . , . . . , , . , _ , , , . . . , _ . , _ . , . . . , , .
iI
'~~' rarns anf /T)agcg ts s ++
5,5 P<os<ow ad Nnmis ; u.,y ,,, ,,,, , j Page 11g of 103 M 2 j 55,3, 6-6-s continued) A required system redundant to support system (s) for the ! c. supported s
' ' ' (Casc C) ? ystems (a) and (b) above is also inoperable
Gene ic Example-Divisi A ivision B Sys mi System i M ase C 4 !
4 Syste ! ystem ii Support System ii 4 Inoperable) 4 ; System ii' Sy em iii da A i 4 4 l System v ystem iv 'ase B I a i W O h The SFDP identifies where a loss of safety function exists. If loss of safety function is determined to exist by this program, a the appropriate Conditions and Required Actions of the LC0 in i which the loss of safety function exists are required to be l entered. l f t%CPJ BlA l born O
.0-31 l t M ItFART 5,5,lo A- @
% r , u -ic 5.0-31 Rev. O, 09/28/92 BWR/6 STS
U-602283 Att. 2 3 Page 120 of 183 JESERT 31A 5.5.11 Ultimate Heat Sink (UHS) Erosion, Sediment Monitoring and Dredging Program A program to provide maintenance on the UHS in the event inspections of the UHS dam, its abutments or the UHS shoreline indicate erosion or local instability. This 1 program shall ensure that the UHS is maintained in such
; a way as to achieve the following as objectives:
4 a. During normal operation, there will be a volume of water in the UHS below elevation 675 sufficient to 1 receive the sediment load from a once-in-25-year flood event; and
- b. Still be adequate to maintain the plant in a safe-shutdown condition for 30 days under meterological conditions of the severity suggested by Regulatory Guide 1.27.
i 1 i 4 4 h i i I' i a 1 4 2 INSERT CLINTON 5,0-31 4/15/94
Reporting Requirements ]
. ,QiM@
.3 5.0 ADMINISTRATIVE CONTROLS m u 22e3 An. 2 i Page f21 or 133
, (6,6 -Ed) Reporting Requi rements @(Lg,1
%gg n- g The following reports shall be submitted in accordance with 1 10 CFR 50.4.
5.9[1 tartupRe[rt [
' A sum report plant start p and pow escalatio testing shall e submit d following:
Receip of an Operati g License-
- b. Ame ment to the cense in lying a pl ned increase in p er level;
- c. Installation f fuel t t has a di erent design or as been manufactur by a di erent fuel upplier; and
@,,- d. Modifi tions th may have 'gnificantly alt ed the nucl , thema , or hydra c perfomance the unit. f The in' ial Sta p Report s 11 address each f the startup te ts iden fied in F R, Chapter 14),andshall nelude a descrip 'on of e measur values of he operating c ditions or aracteris cs obtaine during the tes program and a co arison of these ues with sign predictio and specificatio s. Any correcti e actions t t were require to obtain satisf ctory .
operat' n shall al be described. Any additional s ecific l deta' s required n license cond . ions based on at r comitments sh 1 be inclu d in this repor . Subsequent St tup Reports s all. address startup tests t t are necessary o. demonstrate the. cceptabili of changes an modifications. Startup eports shall be ubmitted withi 90 days followin compl ion of the Star p Test Program- 0 days followin resu tion or comen ent of comer 'al power operatio , or 9 nths following itial critical y, whichever is riiest. If e Startup Repor does not cover 11 three events .e., initial criticality, co letion of Star p Test Program, a d resumption comencement o- comercial op ation),supplemen ry reports st 1 be submitted t least every months until all hree events h e b y een compi ed. (Continued) BWR/6 STS 5.0-32 Rev. O, 09/28/92
Reporting Requirements (
,Q -(5'4M) I 5.6 u-ear 283 a n. 2 4
(-5-t Reporting Requirements Conh'nVed) , Page 122 of 183 h.0.: *u t ' n-e g:;ntiau dQ _ l
,pr u ' 5pertr$ - ~~' P
{C.0.1.2 I--~__ ___.... /-------- --------N E------ ------------ ---------. i single mittal rr y be made or a mul pie unit st< ion. The j ubmitta should c .bine sect' ns commo to all unit at the I_ _ _.....____ _________ --. ...__ ___________ ---_______ ____
~4
[ 1 NL W @o r g# "DgAnnual Rcport+1 w for the covering previousthe activities calendar of the year shall beunit submitted-:: de:Cribed by 7 ; l
, T@ f ,
March 31 of each year. -FThe init;el repor-tuhal' be reb-4 +ted by f} l ihc ycar f01hi".g 4"iti21 cr4+4"lity-][
, . - A di 31 vi _
s,, Repeet+-required en an annual b::i includ; Q r r w horri
'h
$,(, j -er Occupatiunal Radiation Exposure Report rk nibr W WG6 YequireA ,
- l A tabulation on an annual basis of the number of station, I utility, and other personnel (including contractors) I receiving exposures > 100 mrem /yr and their associated man I rem exposure according to work and job functions (e.g., *
, l B reactor operations and surveillance, inser,.g'ce lnspection, ' i
') routine maintenance, special maintenance ![ describe maintenanc 6 waste processing, and refueling). This, tabulation supplements the requirements of 10 CFR 20 The dose assignments to various duty functions may be@lG a estimated based on pocket dosimeter, thennoluminescent ,
dosimeter (TLD), or film badge measurements. Small l exposures totalling < 20% of the individual total dose need I not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sou ces should be assigned to specific major work functionwd , @
-[b . Any othcr unit unique reports requi-r-ed on--an annual bask.] *
(continued) BWR/6 STS 5.0-33 Rev. O, 09/28/92
1 l Reporting Requirements 1 [ 6 (o 5 3 l
% C37 u-e u2ea m . 2 5.3 Reporting Requirement (cogged Paos 123 or ie3
- 5 . 3 .-l (
~"
3
" - t ' .a- " g a , a
..ccrt n _
4 s _7
-5.9.1.3 Annual Radiological Environmental Operating Report y
$4,7-; r -
- ----- ------- ------- ------- OTE--
sin e submi tal may be mad for a ultiple u t statio . The /
-- submi tal sh Id com ne sec ons c . on to al units at the '
/
sta on.
-- ------- ------ ------- ----- --- ...-- --------- --------- - \
" The Annual Radiological Environmental Operating Report covering , the operation of the unit during the previous calendar year shall ! be submitted by May 15 of each year. The report shall include l sumaries, interpretations, and analyses of trends of the results i of the Radiological Environmental Monitoring Program for the I reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Oose Calculation Manual (00CH), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological en/ironmental samples and of all environmental radiation measurement!. taken j during the period pursuant to the locations specified in the table and figures in the 00CM, as we]1 as sumarized and tabulated
-I results of these analyses and measurementsdin the format of the
- table in the Radiological 3sessment Branch Technical Position,
$ -~ k Revision 1, November 1979f. JThe report shall identify the TLD results that represent collocated dosimeters in relation to the s NRC TLD program and the exposure period associated with each
'l result.P available In for theinclusion event that some with the individual report, theresults report areshallnot be submitted noting and explaining the reasons for the missing results. The missing data shall be subihitted in a supplementary report as soon as possible.
(continued) BWR/6 STS 5.0-34 Rev. O, 09/28/92
I Reporting Requirements Cl lo &
, 3 u-co22e3 Att. 2 S' 5,G Page 1216 of 183 U -5 M Reporting Requirements ntinVed
\
C E1 -(=b ".1 Routine "^r^rts- (certin"edh . i l
~ 5.v.Y I
'[$0.1.4 Radioactive Effluent Release Report D.- l
~. --------- --------- -------- NOTE---- ---------- -----------/--
single bmittal y be ma > for a , ltiple uni station. pie submitt should mbine s tions co on to all nits at the/
/ statio , however for uni a with se arate radwa te systems the /
O$ submi tal shall specify le releae s of radio tive mater' 1 from I eac unit. Iddry th The Radioactive Effluent Release Report covering the operation of ccrd=c with 10 CI",50.35s. The
' feNiov6 the unit,l.shall be submittedg =f the quantities of radioactive Cd\tdet report shati include a sumary o liquid and gaseous effluents and solid waste released from the Cao L pr i
unit. The material provided shall be consistent with the p Tg objectives outlined in the 00CH and Process Control Program and in confonnance with 10 CFR 50.36a and 10 CFR 50, Appendix I, o eo, p Section IV.B.1. C.0.1.5 Honthly Operating Reports 5.0,9 y
/ Routine reports of operating statistics and shut wn experiencar,
/ g includin documentation of all challenges to the safety / relief valves, shall be submitted on a monthly basis no later than the OC3'1 15th of each month following the calendar month covered by the
' report.
i T.0,5 t E . 0.1. E CORE OPERATING LIMITS REPORT (COLR)
- a. . Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be' documented.in the COLR for the
~DYERT356
-j C following:/
~
r, e individu specifica ons tha address co e operati g imits must e referenc here.
. The anal tical metho a used to etermine e core o erating limits hall be th e previou ly reviewe and appr ved by the N , speci fic ly those escribed i the fol wing documents: ~
(continued) BWR/6 STS 5.0-35 Rev. O, 09/28/92
. - - . . . . .-. . .~ _ . =-
U-002283 Att. 2 , Pege 125 of 183 j INSERT 35A 1 l
- 1) LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR),
- 2) LCO 3.2.2, Minimum Critical Power Ratio (MCPR),
1
~
- 3) LCO 3.2.3, Linear Heat Generation Rate (LHGR) , and l
- 4) LCO 3.3.1.1, RPS Instrumentation (SR 3.3.1.1.14);
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in General Electric Standard Application for Reactor Fuel (GESTAR), NEDE-24011-P-A, (latest approved version).
l l l INSERT CLINTON 5.0-35 4/11'94 1
i Reporting Requirements
.- ~ & & b* )
6.4 u-602283 Att. 2 P Reporting Requirements e.g. ire or 183 , 54.s I g ,, 5.3.1.G CORE OPERATING LIMITS REPORT (COLR) (continued) _ p
- r_ ___ Q
- dentif the T ical Repo (s) by nu er, tit , date, d NRC s ff app oval docu nt, or id tify the staff Sa ty /
g / Eva ation port for plant spe ific met dology b NRC _le er and date. /
- c. The core operating limits shall be determined such that all
- applicable limits (e.g., fuel thermal mechanical limits, l
core thennal hydraulic limits, Emergency Core Cooling i Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety - analysis are met, f ! d. The COLR, including any midcycle revisions or supplements,
- shall be provided upon issuance for each reload cycle to the 1 NRC.
~
l 5'. (c . 6 l C6'I -
'5. 9 .1. P '
Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 3.40 h -- Q Cgr P/L,im.tk".q'yCS Presou ond TempWaMe v
,. - G. he _RCS pr_ essure and tempera ure iimits, includin heatup Qene
~":9. r:fei? criticality, and hydrostatic and eak test limits, a0 ' lie'estHblished and documented in the PTLR, [The 'ndi.id;;1-s 1
s Spccificatiens that addics; thc rc; cts. ve55ul yressur; and te per:turc liaits and um heeiup and cuviuawa ra m my be-l %_ referenced.' 1The analytical methods used to detennine the pressure and temperature limits including the heetup and ::alded de (shall be those + 3 ubio9 g*g +in [T:pic;l "eyect(:)previously , n;;t:r,reviewed title, date,and andapproved M C ct:f# by =ppthe NRC d::==nt, c. aam*fety =aluation =p=t f= ; plant :pecifie ' i 1.9 ) Renon'2- methodology by NRC ictter and d:td.P/The reactor vessel pressure d QM 10 g g, and temperature limits, including ::::: f:r Se: tup :nd cre! der"
- Appeg8ce6 -r
- t::7 shall be determined so that all applicabl~e limits (e.g, i g heatup li-its, cerld:=- li=itt, nd %:e'uie= 1siak and hudenc+a+ic lQ crd R, t tin: li-it ) of the analysis are met. 4The PTLR, including j
/ revisions or supplements thereto, shall r v' e upon issuance 3
for each reactor vessel fluency perio f, fo /J , 1 4 D (continued) BWR/6 STS 5.0-36 Rev. O, 09/28/92 t y ,- r.-. --r--+r. - - - . . . - , , . , , . ,, - .
. - , - - . w - . , , . - - - w-
Reporting Requirements O T,G
. . Q) 9.Wu u-oo22aa Att. 2 9-t Reporting Requirements (continued) p.a. 127 or ies W Y 5.9 Spec /al Reports -
S etial Re rts may be equired cov ring inspec on, test, an aintenan e activitie . These sp cial report are determin on
.- an indiv dual basis or each un' , and their reparation a submit 1 are desi ated in th Technical ecifications.
C3 l l /_Spe al Reports all be su ,itted in a ordance with CFR 50 wi iin the tim period sp ified for e h report. ( ( The followin Special Reports shall be submitted: C. 4 the ev nt an ECCS 's actuated d injects wat r into the CS in M DE 1, 2, o 3, a Specia Report shall e prepared and sub itted with'n 90 days d cribing the c'rcumstances f i the a uation an the total a umulated acti tion cycles o date The curr t value of e usage fact for each zzle shall b provided in i l aff cted safet injection Special Repor whenever its valun avennric n.70. his ] y S
'esel genera r (EDG)
-l
- b. If an
- dividual mergency exper'ences fou or more v id failure in the 1 t 25 mands, t se failur and any no alid fai res ex erienced that ED in that tim period sh 1 be
/ r ported w' hin 30 da Reports n EDG fail res shall C .
/
nelude 3, Revisio t eRegula informa ry 'on recome Positio edorine Regu C.5, sting toryGuidj1.9,/ Regplatory Guide .108 repor ing require / _ nt. [__ r c. en a Sp ial Repor is requ' ed by Condi on B or G 1 C0 3.3. .1],"Po Acciden Monitoring AM) Instru ntation," a report hall be sub tted with' the folio ng 14 da . The r port shall tline the eplanne alte nate met d of mon' oring, the ause of th in erabilit , and the plans and s edule for estoring e i trumenta on chan is of the F nction to ERABLE s tus.] (continued) BWR/6 STS 5.0-37 Rev. O, 09/28/92
Reporting Requirements Cl h @ d u'-so22e3 At 2 +'r Reporting Requirements e.g. 12e or 103 5.0.2 Lu al Revvi L3 (cuni,,.ued) # _ __ f r _
- d. ny abnormal egradation of 'e containment s ucture detected dur ng the tests r uired by the Pr -Stressed Concrete C tainment Tend Surveillance P gram shall be reported o the NRC with' 30 days. The eport shall l include description the tendon co ition, the cond' ion of th concrete (esp ially at tendon nchorages),the insp tion procedu s, the toleranc on cracking, a the cor ective actio taken. __j BWR/6 STS 5.0-38 Rev. O, 09/28/92 1
. -- -- - - - -. _.= . - .
U-602283 Att. 2 - PQQs 120 of 183 RecordRetention) 5.10
\
ADMINIST ATIVE CONTROLS 5.10 Reco d Retention j 5.10. The following ecords shall be retained r at least 3 years:
- a. All L' ense Event Reports requi d by 10 CFR 50.73;
- b. R ords of changes made to e procedures required ecification 5.7.1.1; an
- c. Records of radioactiv shipreents .
f5.10.2 The following record shall be retained for a least 5 years:
- a. Records and ogs of unit operation c ering time intervals at each p er level;
, b. Recor and logs of principal intenance activities-insp tions, repair, and rep cement of principal items of e pment related to nucle safety;
- c. Records of surveillanc activities, inspections, and calibrations require y the Technical Specifications 7' (TS) [and the Fire rotection Program];
/
/
- d. Records of sea d source and fission detector leak tests and results; and /
- e. Records o annual physical inventory of a)'l sealed source materi of record.
l 5.10.3 The fo owing records shall be retaine for the duration of the unit perating License: a Records and drawing changes eflecting unit design modifications made to sys ms and equipment described in the FSAR;
- b. Records of new and i radiated fuel inventory, fuel 4 transfers, and ass bly burnup histories; l c. Records of rad' tion exposure for all individuals tering radiation co rol areas; (continued) 3
/ j BWR/6 STS 5.0-39 Rev. O, 09/28/92 t._... _ _ . ~ , _ _ _ , . _ _ . __ _
P Record Reten 1 I.a. ao r $8
^ -
, e 5.10 Record Re ntion j ] l l 5.10.3 (c tinued) ,,
- d. Records of ga ous and liquid radica ive material released to the envir s;
/ e. Records o transient or operati nal cycles for those un't componen s identified in [FSA , Section X];
j
/
- f. Recor s of reactor tests a experiments;
- g. Re rds of training and ualification for membe s of the it staff;
- h. Records of inservi inspections performed ursuant to the TS;
- i. Records of qua ty assurance activiti required by the Operational ality Assurance (QA) nual [not listed in Specificati 5.10.1 and which are classified as pennanent records by applicable regulation , codes, and standards];
- j. Records of reviews performed for changes made to procedur ,
equip nt, or reviews of tee s and experiments pursuant o l 10 C 50.59; !
~
)
l
- k. R ords of the reviews d audits required by Speci cation
.5.1 and Specificati 5.5.2;
- 1. Records of the serv ce lives of all hydraulic d mechanical snubbers required y [ document where snubber equirements relocated to], i cluding the date at which e service life commences, and ssociated installation an maintenance records;
- m. Records of nalyses required by the diological Environme tal Monitoring Program th would pennit l
evaluat' n of the accuracy of the nalysis at a later date 1 (these ecords should include pr edures effective at speci ied times and QA records owing that these procedu es wer followed);
- n. cords of reviews perfonn for changes made to the ffsite ose Calculation Manual a d the Process Control Pr ram; and l
o Records of pre-stresse concrete containment te on surveillances. BWR/6 STS 5.0-40 Rev. O, 09/28/92
.~ _
~.- J.- -
tg u-co2n3 At t. 2 h{gjnf padiation Area] Page 131 of 103 [5.1 ] b l 5.0 ADMINISTRAT VE CONTROLS [5.11 High R iationArea) 5.11.1 Pursuant to 10 C 20, paragraph .203(c)(5), in lie of the l requirements of 10 CFR 20.203(c) each high radiatio area, as l defined in 10 FR 20, in which he intensity of ra ation is
> 100 mrem /h but < 1000 mrem r, shall be barric ed and l conspicuou y posted as a h' h radiation area an entrance thereto shall be ontrolled by re iring issuance of a adiation Work Permit WP). Individua qualified in radiacnon protection proce res (e.g., [ Heal h Physics Technician . ) or personnel cont' uously escorted y such individuals , y be exempt from the RWP issuance require ent during the perfo .ance of their assigned d ies in high rad' tion areas with expo ure rates 5 1000 mrem /hr, rovided they are therwise following - ant radiation protection procedures for e try into such high r diation areas.
Any individua or group of individ is permitted to enter such ,- areas shall e provided with or a companied by one or more of thb following: / vice that continuously indica'tes
/
- a. A diation monitoring e radiation dose ra in the area.
l b. A radiation monitor ng device that continuousi integrates the radiation dos rate in the area and alann hen a preset integrated dose s received. Entry into sup areas with this monitori device may be made after the dose rate levels in th area have been establishe /and personnel are aware of t m.
- c. An ind' idual qualified in radiatip protection procedures with radiation dose rate monitofing device, who is l
res onsible for providing posi (vc control over the ! ctivities within the area a shall perform periodic adiation surveillance at e frequency specified by the [ Radiation Protection Ma ger] in the RWP. 5.11.2 In addition to the requir_ments of Specification 5.11.1 reas with radiation levels a 000 mrem /hr shall be provided ith locked or continuously guard doors to prevent unauthorize entry and the keys shall be ma ntained under the administrati e control of the Shift Foreman n duty or health physics super 1sion. Doors I shall remain loc ed except during periods of acc ss by personnel _ (continued) BWR/6 STS 5.0-41 Rev. O, 09/28/92
i 1 U-502283 Att. 2 O(j '7 Pao = 132 or 183 [High Radiation Ar a'
.11
[5.11 High Radiation Area], ,, 5 .11. (continued)
~
under a approved RWP that s all specify the dose ate leve1s in' } the i ediate work areas a the maximum allowa e stay times for indiv duals in those are 2. In lieu of the s y time
/ spe fication of the RW , direct or remote uch as closed circuit / TV cameras) continuou surveillance may be ade by personnel alified in radiat' n protection proced es to provide positiv exposure control o er the activities be'ng performed within t -
area. 5.11.3 For individua high radiation area with radiation level of 1 f > 1000 mrem / r, accessible to per onnel, that are loca d within I large area such as reactor con inment, where no en osure exists l for purp es of locking, or t t cannot be continuo sly guarded, and whe e no enclosure can b reasonably construc d around the ' indiv' ual area, that indi dual area shall be tricaded and l cons cuously posted, and a flashing light sh be activated as war ing device. l J i I I l l i BWR/6 STS 5.0-42 Rev. O, 09/28/92
i U-002283 Att, 2 Page 133 or les i 4 ATTACHMENT 2B ITS - PSTS COMPARISON DOCUMENT DISCUSSION OF CHANGES 1 i I
U-802283 Att. 2 Page 134 o9 183 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS BRACKETED ADMINISTRATIVE CHOICE l B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant specific requirements. PLA_NT SPECIFIC PIFFERENCE P.1 This comment number is not used for this station. P.2 Plant specific staff qualifications have been previously reviewed and approved as identified in NUREG-0853; as supplemented. P.3 This comment number is not used for this station. P.4 This comment number is not used for this station. P.5 The safety analysis report for this station is identified as the Updated Safety Analysis Report and is correctly referred to as the USAR. f P.6 Plant specific organization titles are used where appropriate, P.7 The references to "each" unit and "on the unit affected" has I been revised to reflect that only one unit'is located at this ! site. l l ! P.8 This requirement is only applicable to PWRs. l P.9 The current plant specific requirement details for diesel I generator fuel oil testing are retained in the proposed program. P.10 Where possible, plant specific management position titles in the proposed Technical Specifications are replaced with generic titles as provided in ANSI /ANS 3.1. Personnel who fulfill i l these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details relating to the plant specific managemen oosit3*n titles are identified in licensee controlled documeras (such as the USAR) . The two major specific replacements are the generic " plant manager" for the manager level individual responsible for.the overall safe operation of the plant and the generic descriptive
- use of "the corporate executive responsible for overall plant j nuclear safety" in place of the Vice President position. The l plant specific titles fulfilling the duties of these generic i positions will continue to be defined, established, documented and updated in a plant controlled document with specific regulatory review requirements for changes, such as the USAR or CLINTON 1 4/15/94
_ ._ _ _ . _ _ _ _ _ _ _ . . . _ . _ _. _ ~ _ _ . . _ . l U+602203 Att. 2 Page 135 of 103 l DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER S - ADMINISTRATIVE CONTROLS 4 PLANT SPECIFIC DIFFERENCE (continued) QA Manual. This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation of the corporate and unit organization charts to licensee controlled documents. The intent of the Generic Letter, and of this proposed change, is to reduce the unnecessary burden on NRC and licensee resources being used to process changes due solely to personnel titles changes during reorganizations. Since this change does not eliminate any of the qualifications, responsibilities or requirements for these personnel or the positions, the change is considered to be a change in presentation only and is therefore administrative. P.11 This plant has an additional program that was added to the proposed TSs . The name of this additional program is the,
" Ultimate Heat Sink Erosion, Sediment Monitoring and Dredging Program". .
P.12 This comment number is not used for this station. P.13 This comment number is not used for this station. P.14 This comment number is not used for this station. P.15 This comment number is not used for this station. P.16 This comment number is not used for this station. P.17 This comment number is not. used for this station. P.18 The ref erence to Generic Letter (GL)- 82-33 has been deleted since this GL does not cover all requirements of NUREG - 0737 as presently reflected in the current Technical Specifications. P.19 Appropriate clarification has been provided for sampling of diesel fuel oil. P.20 This comment number is not used for this station. P.21 The discussion has been modified to more accurately describe the safety / relief valves being discussed. P.22 This comment number is not used for this station. P.23 This comment number is not used for this station. P.24 This change has been made to clarify that the limitations on , the concentrations of radioactive material released in l effluents to unrestricted areas, conform to ten times the l- concentration values in 10 CFR 20, Appendix B. ; CLINTON 2 4/15/94
U-002283 Att. 2 Page 136 o r 18 3 i DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS PLANT SPECIFIC DIFFERENCE (continued) P.25 Plant specific criteria has been provided for limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary. P.26 Plant specific criteria has been provided for the Ventilation Filter Testing Program. P.27 This change has been made to provide plant specific limits of 10 Curies for this surveillance program. P.28 This comment number is not used for this station. P.29 This comment number is not used for this station. P.30 This comment number is not used for this station. ! P.31 The plant specific requirements for the component cyclic or transient limit program are identified. P 32 The PTLR discussion is modified to reflect the retention of l requirements in the associated LCO and editorial changes are made to reduce the repetition of information. As part of the , retention of specific limits in ITS SRs 3.4.11.8 and 3.4.11.9, I the acceptance criteria for these SRs were changed to include the limit (e.g., less than or eaual to 100 F/50 F) . This was an administrative change in presentation only to provide consistency in the limits identified and resulted in no real change to the limits. I P.33 The requirement for the Plant Superintendent, or his designee, l to approve proposed tests, experiments, and modifications to l unit equipment that affect nuclear safety has been deleted. l This requirement does not exist in the current CPS Technical Specifications and is adequately controlled by plant procedures. CHANGE / IMPROVEMENT TO NUREG STS C.1 The sections related to Technical Specification Bases Control, Procedures, Programs and Manuals, and the Safety Function Determination Program have been reformatted to include all programs in the Programs and Manuals section. Additionally, the Procedures requirements have been separated into an individual section apart from the Programs and Manuals. Appropriate renumbering of these sections and the ones that follow them are also incorporated, along with revisions to references to these sections. CLINTON 3 4/15/94
i U-602283 Att, 2 Page 137 of 103 DISCUSSION OF CHANGES TO NUREG-1434 i CHAPTER S - ADMINISTRATIVE CONTROLS l CHANGE / IMPROVEMENT TO NUREG STS (continued) , C.2 '6nis change reflects an editorial correction to the wording of ! NUREG-1434. C.3 This comment number is not used for this station. C.4 Procedures to implement the Emergency Plan and the Security Plan are required by 10 CFR 50, Appendix E and 10 CFR 50.54 (p) . Therefore, there is no need to repeat the requirements. C.3 This comment number is not used for this station. C.6 This change is to provide consistency of requirements for the items specified as Manuals. This change prevents potential confusion and misinterpretation. C.7 This change provides editorially rewords the requirement. The change avoids unnecessary confusion that may develop in attempting to determine the intent of different wording. l C.8 This comment number is not used for this station. C.9 10 CFR 20, Appendix B, has been revised such that the correct reference is Table 2 rather than Table II. C.10 This change provides consistency of the wording in the l requirements. The change avoids unnecessary confusion that may develop in attempting to determine the intent of different wording. C.11 This change is made to provide consistency with the generic surveillance frequencies. C.12 The description of the entry conditions into the SFDP are l clarified and generalized to assure that they include all l possible required entry conditions. l C.13 This comment number is not used for this station. C.14 This comment number is not used for this station. C.15 This comment number is not used for this station. l C.16 This comment number is not used for this station. C.17 This comment number is not used for this station. l C.18 This comment number is not used for this station. C.19 These changes provide for consistency with the new 10 CFR 20. l CLINTON 4 4/15/94 t . - - -. .
U-602283 Att. 2 Page 130 of 103 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER S - ADMINISTRATIVE CONTROLS l l CHANGE / IMPROVEMENT TO NUREG STS (continued) C.20 This comment number is not used for this station. C.21 This change removes a requirement that is only applicable to Pressurized Water Reactors (PWRs). C.22 This comment number is not used for this station. l C.23 These changes reflect editorial corrections to the wording of j NUREG-1434. These changes conform with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the i Chairpersons of the Owners' Groups Technical Specifications l Committees dated October 25, 1993. C.24 The requirements of Table 5.2.2-1 are removed for the Technical Specifications and will be controlled by the licensee's administrative controls.10 CFR 50.54 provides the requirements for shift complement regarding licensed operators. Additionally, Section 5.2.2.b specifies when a licensed operator must be in the control room. The Table 5.2.2-1 requirements associated with the auxiliary (non-licensed) operators are retained as 5.2.2.a and 5.2.2.c retains the allowance for unexpected absences. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993 with editorial rewording of 5.2.2.a to be consistent with the wording of 5.2.2.b, the addition of the allowance for unexpected absences of an auxiliary operator which was allowed on Table 5.2.2-1, and the addition of the allowance for an on shif t SS or SRO to perform the STA function as was allowed by Table 5.2.2-1. C.25 The requirements for SRO presence during fuel handling and core alterations are contained in 10 CFR 50.54. Therefore, there is no need to repeat the requirements. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.26 The requirements for the STA to meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shif t is moved f rom 5. 3.1 to 5. 2. 2.g. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.27 The retraining and replacement training program requirements for the unit staff are relocated. Details of this program can CLINTON 5 4/15/94
U-602203 Att. 2 Page 130 of 183 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS j (continued) ! be adequately controlled by the licensee's administrative controls. Technical Specification Section 5.3, Unic Staff Qualifications, provides adequate requirements to assure an acceptable, competent unit staf f . In accordance with Technical Section 5.3 and CPS commitments, members of the unit staf f meet or exceed the minimum qualifications of the specific ANSI Standards for their positions. Additionally, Technical Specification Section 5.2, Organization, details unit staff requirements. Sections 5.2.2.a and 5.2.2.b and 10 CFR 50.54 describe the minimum shift crew composition and delineates those positions which require an RO or SRO license. Training and qualification for these positions are specified in 10 CFR 55. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.28 The review and audit requirements can be adequately controlled by the licensee's administrative controls. The control of the these requirements would rely on the Quality Assurance Manual implementing 10 CFR 50.54 and 10 CFR 50, Appendix B to control the requirements in conjunction with the USAR and procedures. Such an approach would result in an equivalent level of regulatory authority while providing for a more appropriate change control process. The net affect of the change is the level of safety of f acility operation is unaf fected and NRC and utility resources associated with processing license amendments to this Administrative Control are optimized. The onsite review function, composition, alternate membership, meeting frequency, quorum, responsibilities, authority and records are all covered in equivalent detail in ANSI N18.7-1976 and the current Technical Specification requirements will be contained in the USAR. The offsite review group is also addressed, although with less detail, in ANSI N18.7-1976. The USAR will include the requirements for the offsite review group. Audit requirements are specified in the QA Program to satisfy 10 CFR 50, Appendix B, Criterion XVIII. These audit requirements will be augmented by the inclusion in the USAR or procedures those audit requirements currently contained in the Technical Specifications . In addition, audits are also covered by ANSI N18.7, ANSI N45.2, 10 CFR 50. 54 (t) , 10 CFR 50.54 (p) , and 10 CFR 73. Therefore, duplication of the requirements contained in the above documents by the Administrative Controls CLINTON 6 4/15/94
l U-002283 Att. 2 I Page 140 or les DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER S - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO 11UREG STS (continued) Section of the Technical Specifications does not enhance the level of nuclear safety for the unit. Therefore, the l provisions relating to audits are not necessary to assure safe operation of the facility. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated i October 25, 1993. References to these controls are ; corresponding moved to the license's administrative controls. C.29 The details of review and approval requirements for procedures and programs can be adequately controlled by the licensee's administrative controls. These requirements will be included 1 in the USAR. This proposal is based on the existence of the l following requirements which duplicate 10 CFR 50.36 in these I areas and which assure operation of the f acility in a safe manner. The requirement for procedures is mandated by 10 CFR 50, Appendix B, Criterion II (Second sentence) and Criterion V. l ANSI N18.7-1976, which is an NRC Staff-endorsed document used in the development of many licensee QA plans, also contains specific requirements related to procedures. ANSI N18.7-1976, Section 5.2.2 discusses Procedure Adherence. This section clearly states that procedures shall be followed, and the requirements for use of procedures shall be prescribed , in writing. ANSI N18.7-1976 also discusses temporary changes l to procedures, and requires review and approval of procedures i to be defined. 1 ANSI N18.7-1976, Section 5.2.15 describes the review, approval and control of procedures. The section describes the requirements for the Licensee's Quality Assurance program to provide measures to control and coordinate the approval and changes thereto, which issuance of documents, including ! prescribe all activities affecting quality. The Section l further states that each procedure shall be reviewed and approved prior to initial use. The reviews required are also described. Licensees can continue to implement the requirements of 10 CFR 50, Appendix B, regarding procedures without duplicating the l necessity of procedure requirements in the facility Technical Specifications. Safe operation of the plant will continued to be maintained, and therefore, the requirements for procedures and their control should not be re-addressed in Technical Specifications. Duplication of the provisions related to procedures is not necessary to assure safe operation of .the facility. CLINTON 7 4/15/94
~ . -. - . - -. -- - -- - . - . __ = -
l l U-802283 Att. 2 Page 18e1 oP 183 i DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER S - ADMINISTRATIVE CONTROLS CHANGE /IMPRQVEMENT TO NUREG STS (continued) Procedure control requirements are already addressed in the Operational Quality Assurance Manual and those requirements will be supplemented by the addition of the current Technical Specification requirements into the USAR. This change would result in an equivalent level of regulatory authority while providing for a more appropriate change control process. The l net affect of the change is the level of safety of facility operation is unaffected and NRC and utility resources associated with processing license amendments to this Administrative Control are optimized. l This change conforms with the changes suggested by the NRC in i the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.30 The requirement for procedures to implement the personnel radiation protection requirements of 10 CFR 20 can be adequately controlled by the licensee's administrative controls. These procedures are developed to ensure nuclear plant personnel safety and have no impact on nuclear safety. Additionally, nuclear plant personnel are not " members of the public." Thus, the principal operative standard in Section 182a of the Atomic Energy Act; " health and safety Jf the public" does not apply. Based on these consideratians, the Radiation Protection Program administrative contral is not necessary to assure operation of the facility in a safe manner and can be deleted from Technical Specifications. The requirement to have procedures to implement Part 20 is also contained within 10 CFR 20.1101(b). Periodic review of these procedures is addressed under 10 CFR 20.1101(c). This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.31 The requirement for a Process Control Program (PCP) and its review and approval requirements can be adequately controlled by the licensee's administrative controls. The PCP can be adequately described in another controlled documents, e.g. , the ODCM and the USAR. Control of changes is preserved by 10 CFR 50.54a. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, and 10 CFR 71. Relocating the description of the PCP does not affect the safe operation of the facility. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. I CLINTON 8 4/15/94
._ ~ _- - - -- - -~. .. - ._ - ..- U 002283.Att, 2 Page 142 of 103 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) C.32 The radiological environmental monitoring program is required l to be part of the ODCM, the additional details of the program l can be adequately controlled by the licensee's administrative i controls. The ODCM requirements in conjunction with the USAR will continue to control this program. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups i Technical Specifications Committees dated October 25, 1993. l ! C,33 The requirement for an in plant radiation monitoring program and the associated details of the program can be adequately controlled by the licensee's administrative controls. This program provides controls to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program was developed to minimize radiation exposure to plant personnel post-accident and has no impact on nuclear safety. Additionally, nuclear plant personnel are not " members of the public. " Thus, the ! principal operative standard in Section 182a of the Atomic l Energy Act; " health and safety of the public" does not apply, i Based on these considerations, the In Plant Radiation Monitoring Program administrative control is not necessary to assure operation of the facility in a safe manner and can be l removed from Technical Specitications. This program will continue to be controlled by the licensee's administrative controls. This change conforms with the changes suggested by
- the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated l October 25, 1993.
C.34 The requirement for an Inservice Inspection Program and the Inservice Testing Program can be adequately controlled by the licensee's administrative controls and the requirements of 10 CFR 50.55a. Removal of the Inservice Inspection Program and the details of the Inservice Testing Program conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. l C.35 The requirement for a Fire Protection Progra:n and the associated details of the program can be adequately controlled by the licensee's administrative controls. The Fire Protection Program provides controls to ensure that appropriate fire protection measures are maintained to protect the plant from fire and to ensure the capability to achieve and maintain safe shutdown in the event of a fire. The administrative control provides assurance that the capability to provide for alternate / dedicated safe shutdown in accordance with 10 CFR 50, Appendix R. As such, it does not directly assure nuclear CLINTON 9 4/15/94
( U-002283 Att. 2 Page 143 of 183 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) safety, but rather allows for the ability to place the unit in a more stable condition in the event of a fire. The deletion of this administrative control from Technical Specifications is also consistent with the guidance in NRC Generic Letter 86-10 " Implementation of Fire Protection Requirements. " In that letter, the NRC concluded the provisions of 10 CFR 50.59 should apply directly.to changes the licensee desired to make in the fire protection program so long as those changes did not adversely affect the ability to achieve and maintain safe shutdown. The standard license condition, included within 86-10, stated that changes which adversely affected the ability to achieve and maintain safe shutdown in the event of a fire required prior approval of the Commission. Thus, the license condition established as part of the NRC Generic Letter 86-10 implementation also makes this administrative control unnecessary. Based on these considerations, the Fire Protection Program administrative control is not necessary to assure operation of the f acility in a safe manner and can be removed from Technical Specifications. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.36 The requirement for a Startup Report and the associated details of the report and when it should be submitted can be adequately controlled by the licensee's administrative controls. The report was a summary of plant startup and power escalation testing following receipt of the Operating License, increase in licensed power level, installation . of nuclear fuel with a different design or manufacturer than the current fuel, and modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. The report provided a mechanism for NRC to review the appropriateness of , licensee activities after-the-fact, but provided no regulatory authority once the report was submitted (i.e., no requirement i for Commission approval). The approved 10 CFR 50, Appendix B, Quality Assurance Plan and Startup Test Program provide , assurance the listed activities are adequately performed and that appropriate corrective actions, if required, are taken. i Given that the report was required to be provided to the Commission no sooner than 90 days following completion of the respective milestone, it was clearly not necessary to assure operation of the facility in a safe manner for the interval CLINTON 10 4/15/94
U-602283 Att. 2 Page 199 or 183 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) between completion of the startup testing and submittal of the report. Additionally, given there is no requirement for the Commission to approve the report, then the Startup Report is not necessary to assure operation of the facility in a safe manner. Based on these considerations, the Start-up Report may be removed from TS and relocated to a licensee controlled document. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.37 Since no report descriptions remain in the Special Report section, this section heading is removed, the corresponding Routine Report Section designation is removed, and the items are renumbered to reflect these changes. C.38 The requirement for a Special Report following an ECCS injection and the associated details of the report and when it should be submitted can be adequately controlled by the licensee's administrative controls. Title 10, Part 50, Section 73 already provides the requirement for the license to submit a Licensee Event Report in the event of an ECCS actuation. The report is required to be submitted within 30 days'and will contain the same type of information as the special report. Removing a duplicate requirement from TS has no impact on assuring safe operation of the facility since the requirement to submit a report to Commission still exists in 10 CFR 50.73. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.39 The requirement for a Special Report following EDG failures, the associated details of the report, and when it should be submitted can be adequately controlled by the licensee's administrative controls. The report provided a mechanism for NRC to review the appropriateness of licensee activities ! after-the-fact, but provided no regulatory authority once the l report was submitted (i.e., no requirement for Commission l approval). Given that the report was required to be provided l to the Commission no sooner than 30 days following the EDG l failure, it was clearly not necessary to assure operation of the facility in a safe manner for the interval between the EDG f ailure and submittal of the report . Additionally, given there CLINTON 11 4/15/94 l 1 L - - .,- -- , , -. , , , - , . - - . -, - . . -
- _ . _ . . _ ~ . . _ _ . - .
U-602283 Att. 2 Page 1 f4 5 o f 183 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) is no requirement for the Commission to approve the report, this Special Report is not necessary to assure operation of the facility in a safe manner. Based on these considerations, this Special Report may be removed from TS and relocated to a licensee controlled document. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.40 The requirement for a Special Report following extended Post Accident Monitoring instrumentation inoperability and the associated details of the report and when it should be submitted can be adequately' controlled by the licensee's administrative controls. The report provided a mechanism for NRC to review the appropriateness of licensee activities after-the-fact, but provided no regulatory authority once the report was submitted (i.e., no requirement for Commission approval). Given that the report was required to be provided to the Commission no sooner than 30 days following the instrumentation f ailure, it was clearly not necessary to assure operation of the facility in a safe manner for the interval between failure of the instrumentation and submittal of the report. Additionally, given there is no requirement for the Commission to approve the report, this Special' Report is not necessary to assure operation of the f acility in a safe manner. Based on these considerations, this Special Report may be removed from TS and relocated to a licensee controlled document. This information has been added to the Bases for the l LCO Required Actions which required the Special Report to be ! written. This change conforms with the changes suggested by l the NRC in the letter W. T. Russell (NRC) to the Chairpersons l of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.41 Record retention requirements and the associated details of the report can be adequately controlled by the licensee's administrative controls. The requirements on record retention may be removed from Technical Specifications on the basis that they are adequately addressed by the QA Plan (10 CFR 50, Appendix B, Criteria XVII)and because provisions relating to record keeping do not assure operation of the facility in a safe manner. l CLINTON 12 4/15/94
l U*002283 Att. 2 Page 1:40 of 103 4 DISCUSSION OF CHANGES TO NUREG-1434 i CHAPTER 5 - ADMINISTRATIVE CONTROLS l l CHANGE / IMPROVEMENT TO NUREG STS (continued) l Facility operations are performed in accordance with. approved 1 written procedures. Areas include normal startup, operation l and shutdown, abnormal conditions and emergencies, refueling, 1 safety related maintenance, surveillance and testing, and I radiation control. Facility records document appropriate j station operations and activities. Retention of these records 1 provides documentation retrievability for review of compliance l with requirements and regulations. Post compliance. review of I records does not assure operation of the facility in a safe manner as activities described in these documents have already been performed. Numerous other regulations such as 10 CFR 20, Subpart L, 10 CFR 50.71, etc. also require the retention of certain records related to operation of the nuclear plant. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.42 The accepted alternate methods of complying with 10 CFR 20.203(c) for high radiation areas can only be changed with prior NRC approval and can be adequately controlled by the i licensee's administrative controls. These controls are developed to ensure nuclear plant personnel safety and have no i impact on nuclear safety-. Additionally, nuclear plant : personnel are not " members of the public. " Thus, the principal operative standard in Section 182a. of the Atomic Energy Act;
' health and safety of the public' does not apply. Based on these considerations, the Radiation Protection Program l administrative control is not necessary to assure operation of the f acility in a safe manner and can be deleted f rom Technical Specifications.
This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.43 In general the format of the NUREG-1434 does not include the l use of cross references. These unneeded cross references are ! deleted. C.44 The TS need not require an administrative letter be issued to station personnel on an annual basis describing responsibility to the Shift Supervisor. Organizational responsibilities are adequately described by the station's internal administrative controls, Repeating the organizational responsibilities via an internal management directive only increases the administrative CLINTON 13 4/15/94
U-002283 Att. 2 Page 147 of 183 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE /_JMPROVEMENT TO NUREG STS , (continued) burden on the facility with no resulting benefit, Plant safety is not compromised by this proposed change. C.45 This comment number is not used for this station. C.46 The generic example is removed. i i l CLINTON 14 4/15/94 1
U-607203 Att. 2 n... i, e i .3 y CLINTON POWER STATION IMPROVED TECHNICAL SPECIFICATIONS I CONVERSION PACKAGE i l SECTION 1 , USE AND APPLICATION l l
! U-002283 Att. 2 2 Page 140 of 103 gg f 7 1.1 Definitions l CORE ALTERATION monitors, traversin ! (continued) movable detectorsincluding (g incoreundervessel probes, or special 2 replacement) is not considered a CORE ALTERATION. j . In addition, control rod movement with other than { the normal control rod drive is not considered a 4 CORE ALTERATION provided there are no fuel ! assemblies in the associated core cell, i Suspension of CORE ALTERATIONS shall not preclude { completion of movement of a component to a safe
/ position.
l CORE OPERATING LIMITS The COLR is the unit specific document that.
! REPORT (COLR) provides cycle specific parameter limits for the i current reload cycle. These cycle specific limits i shall be determined for each reload cycle in I accordance with Specification (5MM Plant
) operationwithintheselimitsis)addressedin. individual Specifications, g;;, g 1 DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would i produce the same thyroid dose as the quantity and , isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose i conversion factors used for this calculation shall
- be those listed in;{ Table III of TID-14844, AEC,1962, "CalcuTation of Distance Factors for Power and Test Reactor Sites" dr the e 1 M ed #"
l [i HMe r7 ^# " ; & t m Ca i 1.100. "^" uce , 1 o771 __ 1
- 8-AVERAGE $ shall be the average (weighted in proportion i
DISINTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the j sum of the avera L j 1 j fI disintegration (ge beta and gamma energies perin_ MeV).for iodines, with half lives >M15F86inutes, making up at least 95% of the total noniodine activity in _ the coolant. ,_ 1 EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS j TIME initiation setpoint at the channel sensor until i' the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their
; (continued)
- CL WTcuJ .
j - 0= /0 ST b 1.1-3 Rev. O, 00/00/01 l 4 2
u-602283 Att. 2 I" " " "' ' 8 8 Definitions l 1.1 Definitions (continued) MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio class o(CPR) that f fuel}'.' The CPR exists is that in power the core in the 9for each 6j assembly that is calculated by application of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. MODE l A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with C3 el in the reactor vess_el .r] OPERABLE-OPERABILITY A system, subsyste & kA R_ortRABIt.tTf , 5 \ shall be OPERA 8LE en ,iscomponen,t capable o.f ot devtgr_ i, l performin i its specified safety function (s) and when all necessary attendant instrumentation, controls, gg I normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are readred r for the system. subsystem, 4w444, component, or device to perfona its specified safety function (s) are also capable of performing their related support function (s). PRSIEfMESTS s PHYSICSTbTSshanbethosetestsperformedto*
^
measure the reactor the , core and related instrumentationf3f These 'ps'fs are: s ,. /
/ /
' a. escribed in Chapter '[14,' initial Test pl i Program) of the.FSAR
\ Authorized under'j ;'
\ , b. the provisions of I
s /' 10 CFR 50.59; or
/ -
s c. OtherWfse ap' proved by the Nuclear Regulatory Comission. PRESSURE AND The PTLR is the unit specific document that TEMPERATURE LIMITS provides the reactor vessel pressure and REPORT (PTLR) temperature limitst' .C ..., '. A M -- ' t - nt::, for the current reactor vessel fluence period. These pressure and temperature limits (continued) C LirJTo r)
- CWR/6 5 b -
1.1-6 Rev. O, ^9/Z5/E -
U-602203 Att. 2 Pa0s 151 u9 183 96 Definitions Enc 7,y I Pa f71 1.1 Definitions [ h g C.!.7 d ,(o,(p,
~,
PRESSURE AND shall be determined for each fluence period in TEMPERATURE LIMITS accordance with Specification m.0.1 73 Plant REPORT (PTLR operation within these operating limits is (continued addressed in LC0 3.4.11. "RCS Pressure and Temperature (P/T) Limits." RATED THERMAL POWER RTP shall be a total reactor core; heat transfer (RTP) rate to the reactor coolant of 3833 MWt. REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval l SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TlHE trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or l total steps so that the entire response time is measured. j SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical i assuming that:
- a. The reactor is xenon free;
- b. The moderator temperature is 68'F; and i
- c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
r-H With control rods not capable of being fully
@ inserted, the reactivity worth of these control rods must be accounted for in the determination of
$>]
i STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems,
, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function, q g (continued)
- CWR/0 STS 1.1-7 Rev. O, GG/Zo/ K -
l l
i
- D U 602283 Att. 2 -
Page 152 of 183 l 1
- CLINTON POWER STATION l
1
- IMPROVED -TECHNICAL SPECIFICATIONS
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CONVERSION PACKAGE i
! SECTION 2 SAFETY LIMITS l
4 4 4 4
SLs i i p, . 33 er $ea M
- 2.0 SAFETY LIMITS (SLs) !
i l i 2.1 SLs ' i , 2.1.1 Reactor Core SLs 2.1.1.1 With 'the reactor steam dome _ pressure < 785 psig or core ! j flow < 10% rated core flow: l ' \ THERMAL POWER shall be s 25% RTP. j 4 2.1.1.2 With the reactor steam dome pressure m 785 psig and core i flow a 10% rated core flow:
$1 v MCPR shall be 1.07 operation or a 1.08h forrsingle two looJrecirculation/h oo recirculationg operation. V
, 2.1.1.3 Reactor vessel water level shall be greater than the top 4 of active irradiated fuel. l l 2.1.2 Reactor Coolant System Pressure SL
, 4-Reactor steam dome pressure shall be 95ahtsi.ied]s 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed: 2.2.1 Within I hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.2 Within 2 hours: 2.2.2.1 Restore compliance with all SLs; and i 2.2.2.2 Insert all insertable control rods. ib , 2.2.3 Within 24 hours, notify the.C:n;rdbnager ant
, Nica Pre;4d=6 =u_itan cra, m.@ arTd thej[Odub cff:ite ren erer O% .', cpccif;cd i;- Specifi;;tnn.5.5.2, "[^TTaiki ":^.ic ...d Aud. ;[r A l jike Eorporche cecah've respondb, ,M /"e $
d i for overd dawh nuclear {cke 'gg2 [ l SW"/E STS - 2.0-1 Rev. O, 00/20/02 C LINTo s] l All PMe5
t . u-e02283 Att. 2 go gg 2Q Page 154 of 183 Sect' & age 33 l 2.0 SL: 2.2 SL Violations (continued) hN, , n u__ g, ~ , 2.2.4 Within 30 days, a Licensee Event Report (LER) shall be prepared The LER shall be submitted m to the NRC, pursuant y,, to 10 CFR tku v i 2 . a, re.:1 = : 50.73.-:;
.,. 4.
icui ii a u i' < 1 1 i t a t i oii a . s . t. N.J the
. 4Genemp4 M anaaer4~"uc icaD@d Orr "-r-i rr - pi^2 %
.%' " - - - - -" " ~ %e co r p ore.+e eged wk' vore es pc-n s M n ,. , , a . . ;, ,
r _ ' ave'ralFpht nulear Me+H . 2.2.5 Operation of the unit shall not be resumed until authorized by the NRC. l 1 i l l l l
-BWR/5 STS- 2.0-2 Rev. O,-09/2S/92
I - Reactor Core SLs O'": ""; ' BASES W] .8 2.1.1 J SAFETY LIMIT VIOLATIONS 2.2.3 h!- Q3 l-2# ^"^ I^'# (continued) If any SL is viorafeT, thej#ppropriat: :enior management-ofML, [the--nuc4eaFo!:nt and the bLil'Lyfshall be notified within : 24 hours. ~The 24 hour period provides time for plant ! operators and staff to take the appropriate innediate action I and assess the condition of the unit before reporting to the l senior management.
'Jag ,E ..oec$ d de \/hredde t- Ableah-If any SL is violated, a Licensee Event Report shall be prepared and submitted within 30 days, to the NRCfpe-sento&
Emanagement Q resident n;.gl._Lae.nuc! ec!earloceratien: ear .p! ant, u ,P the utiilty Vib (4-
~
TM: neuhe in accordance with 10FCFR 50.73 (Ref A6)[. TA 6 Y6 kb6 bl% d ', gi L?af! If any SL is violated, restart of the unit shall not h comence until authorized by the NRC. This requirement ' ensures the NRC that all necessary reviews, analyses, and-actions are completed before the unit begins its restart to normal operation. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.
-2. NEDE-24011-P-A, (latest approved revision).
( i
- 4. An-nf ac* (n) , L ' ' - "-i-'"" '
~
3% 10 CFR 50.72. O gg gjyn gwc,-J
$N k, \ 10 CFR 100. App llce.+lm $wt keae40r j
Tj,h 10 CFR 50.73. ' l l -BWR/C STS; B 2.0-7 ~ Rev. 0,-99M8/M-- l
l u-eo2283 4e RCS Pressure SL l PeDe 15 0 o r e3 8 2.1.2 BASES l SAFETY LIMIT 2.2.2 VIOLATIONS (continued) Exceeding the RCS pressure SL may cause imediate RCS failure and create a potential for radioactive releases in o excess of 10 CFR 100, " Reactor Site Criteria," limits 5~7eC4). e Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours. The 2 hour Completion-Jig.,gnsurg.sJhr C l,4 - o reo p, , ear-ompt' remedi al,gt 4 nd 160 en6ures 4 1-p oF an o.ce;cien o i focHrg . . 2. W ~5 ~. t n \ .T n n gO } If any SL is_v repristedenen e naoYnt-et-the-nudear--plant and-the-ut4MtyJshalI De notifica witnf 24 hours. ThT241ottr pu~fW1rovides . time for plant operators and staff to take the appropriate im,ediate action and assess the condition of the unit before reporting to the senior-management ._m..-~ t A F-olin ei7esiden (No r W /g .@ j If any SL is violated, a Licensee Even Report shall be t' prepared and submitted within 30 da f{xer4management-of-the nuctear he-utility Vfee pient@ysLtJLthe dent-Nuc4ea4 Rest 4_ons@regtnxemen '* accordance with 10 CFR 50.73 (_Ref. (@). g6 c#t/>c. hEfW[1750 6 e submi itu ' ,^ Q 4 . 2.2.5 If any SL is violated, restart of the unit shall not comence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to 1 normal operation. REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28. I 2. ASME, Boiler and Pressure Vessel Code, Section III, P4 (RicWNB;70Mi3 g- /l y t a. u su , says. a. i. s . ) 1 1 (continued) l 4WR/6 STS - B 2.0-10 Rev. O,-C9/28/92 1 1
U-602283 Att. 2 1 Pags 157 or 183 21 CLINTON POWER STATION l 1 IMPROVED TECHNICAL SPECIFIC.ATIONS CONVERSION PACKAGE SECTION 3.0 l LCO AND SR APPLICABILITY
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LCO Applicability u-602283 Att. 2 3.0 Page 158 of 103 3.0 LCO APPLICABILITY LCO .0.4 specified conditions in the Applicability that are _ require (continued) to comply with ACTIO ge pa.cf o f a 4 74, o Exceptions te this Specification are stated in3N individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perfonn the te-ting required to demonstrate OPERABILITY. LCO 3.0.6 When a supported system LC0 is not met solely due to a ' support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to f> , LCO 3.0.2. for the supported system. In this event, additional evaluations and limitations may be required in accordaEe with specification (53R " Safety Function Detennination Program (SFDP)." If a loss of safety function i is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2. (continued) CWR/E STS 3.0-2 Rev. O,-99/00-/92-
LCO Applicability B 3.0 s e [ b LC0 3.0.6 declared inoperable or direct entry into Conditions and (continued) Required Actions for the supported system. This may occur imediately or after some specified delay to perfom some ' other Required Action. Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared I inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2. g ., a _ g Specification , " Safety Function Detemination Program" l (SFDP), ensures loss of safety function is detected and appropriate actions are taken. Upon failure to meet two or l more LCOs concurrently, an evaluation shall be made to detemine if loss of safety function exists. Additionally, ! other limitations, remedial actions, or compensatory actions l may be identified as a result of the support system ' inoperability and corresponning exception to entering supported system Conditions and Required Actions. The SF0p implements the requirements of LCO 3.0.6. 4 Cross division checks to identify a loss'of safety function for those support r" stems that support safety systems are required. The cross division check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained. ) If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. i l LCO 3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit. These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perfom , special evolutions. Special Operations LCOs in Section 3.10 allow specified TS requirements to be changed to pemit performances of these special tests and operations, which otherwise could not be performed if required to comply with the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will i (continued)
-BWRM STS 8 3.0-8 Rev. O, 19/2S/S:
U 6C;283 Att. 2 Page 100 or 183 2 CLINTON POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION PACKAGE SECTION 3.3 INSTRUMENTATION l l j l l 1 l l l l
U-002283 Att. 2 Page 101 of 103 j Enc 02 96 Sectic age 6 of 687 PAM Instrumentation
- " 3.3.3.1 4^
3.3 INSTRUMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation i } LCO 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1-1 shall be OPERABLE. i l APPLICABILITY: MODES 1 and 2. I i i ACTIONS 1
-------------------------------------NOTES------------------------------------
\
i 1. LCO 3.0.4 is not applicable. t i l 2. Separate Condition entry is allowed for each Function. \ l 1 CONDITION REQUIRED ACTION COMPLETION TIME i i A. One or more Functions A.1 Restore. required 30 days a with one required channel to OPERABLE j channel inoperable. status. { . ^ 2 B. Required Action and B.1 Initiate action k Imediately associated Completion Fn s . J...-- - , G ' Time of Condition A 5 4 x -: f : . . ..r not met. W. " . ,;, - kpMp& GJvm xctTC.,008]l e s-w w*.J I C. ------- 'NO C.1 Restore one required
-- 7 days j Not icab to channel to OPERABLE '
t rogen nitor status.
- t[,hannels 4
_________________f One or more Functions
- with two required
,. channels inoperable. I j (continued) j 4WR/6=STP 3.3-18 Rev. O, 09/24/S2_.
\
i d
. - - - . . - - . . . - - . . . ~ .- - . - - . . - . - - . . .- . . - . - . . _ - . - .- -
U-002283 Att. 2 - Page 102 or 183 l 6 s 7 PAM Instrume a '] ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME Two equired hyd M b D.1 re one [Pequired 72 hours ' onitor lichannel tto'lemonitor] N annels N hydro ! i irtonerable. PERABLE N 'N N, N -'s tatus.
-~ , y" (f'
! . Required Action and ' (.1 Enter the Condition Imediately associated Completion referenced in Time of Condition C- Q4 Table 3.3.3.1-1 for g not met. ' the channel. ' [.Asrequiredby D .1 Be in MODE 3. 12 hours Required Action Il and referenced in g 9q Table 3.3.3.1-1. \ F / P - D As required by Initiate action 4* Imediately Required Action f.1 .aesa-da,,- . m i and referenced n % 4 'i m+ # - M
/ b\ Y Table 3.3.3.1-1. ni Y W 2-
! # n. -
. ww I '
i i b q an* c~ L QW%90
-BWR/G STS 3.3-19 Rev. O,19/48/92m r
l
_ ~ . . _-_ - . - , . . . - - - . U-002 83 Att. 2 Pegs te3 or 133 Enc - 2 . 6 PAM Instrumentation secti age . r687 B.3.3.3.1 BASES ACTIONS function of the instruments, the operator's ability to (continued) diagnose an accident using alternate instruments and methods, and the low probability of an event requiring these l instruments. I 1 A Note has also been provided to modify the ACTIONS related I to PAM instrumentation channels. Section 1.3, Completion egg Times, specifies that once a Condition has been entered, l 5j subsequenM', subsystems, components, or variables 4 expressed in the Condition, discovered to be inoperable or ) not' within limits, will not result in separate entry into h the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each . additional failure, with Completion Times based on initial l entry into the Condition. However, the Required Actions for I inoperable PAM instrumentation channels provide appropriate compensatory measures for separate inoperable functions. As such, a Note has been provided that allows separate I Condition entiy for each inoperable PAM Function, j A .1 '
^
When one or more Functions have one required channel that is I inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion l Time is based on operating experience and takes into account ) , the remaining OPERABLE channel, (or in the case of. a Function . ! that has only one required channel, other non-Regulator 1 Guide 1.97 instrument channels to monitor the Function)y the , passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. l 7 Yt p@. E_d {S u b o. m N h,{XAT B, dnot beenAction If a channel has restored to OPERABLE status in e
%8' -
J 30 days, this Required actions ,' r^ 2 .... __."..'_ :E :. __ _"__' 's ..?_- u _u -__ _ . _ specifies init ation of
- " - - ^ ^
.;. m2 +
}} ' r 'W_ . ~ ~ w; ;m; : - t'r ' " ' " " '
f W' ' .c ' ' ',,-.,....'m to the NRC. This 7 . . . . . ' . . . . _ _ ... . . i. .z mm .
% 00\-9 report discusses the results of the root cause evaluation of the in w rability and identifies proposed restorative l actionsf This Action is aDDropriate in lieu of a shutdown M hied - b h h $g
) (Wb' O(51f,9, [ lD IN dau<heNM 4 fu '
OfdidC#bM b NONN [ - g ntinued) L M sh'Oyg B.) --- T:C/0 m.> B 3.3 Rev. O, 49/28492 -
U-602283 Att. 2 Page 109 of 183 Enc 2t - 6 PAM Instrumentation Sect' age f687 8 3.3.3.1 BASES ACTIONS (continued) Condition referenced in the Table is Function dependent. Ocq r Each time an inoperabl nnel has not me jty Required !s
/ Action of Condition C r Oft eygi nabidJand th A associated Completion m6hasexpired,Conditionf/is
-$ l l
entered for that channel and provides for transfer to the i appropriate subsequent Condition. l I , ! N , l
-6M '
l For the majority of Functions in Table 3.3.3.1-1, if any Required Acti a d associated Completion Time of l l [PI . Condition C ee-B is not met, the plant must be placed in a MODE in whic e LCD does not apply. This is done by l l placing the plant in at least MODE 3 within ,12 hours. { The allowed Completion Times are reasonable, based on operating experience, to reach the required plant condition [ from full power conditions in an orderly manner and .without challenging plant systems. i
\ -
rp r Since alternate means of monitoring Eacter emel -eleF l *, M'pjg M }q h 7_,AevelpDprimary containment 7 area r fation have been N yhg a j q- 'q$ / develooed and tested. the Required Action is not to shut l down the plant but rather tokf:ll::. tt: fir::ti:.x i ! i QpT S 0:_;';. ;;_.. ;.L^... These alternate means may be
# D, V-l Q temporarily installed if the nomal PAM channel cannot be restored to OPERABLE status within the allotted time. The
%%N l
l
- report provided to the NRC should discuss the alternate ,
- means used, describe the degree to which the . alternate means i l
l l , are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels. 2 l 1 i ; l l d ' SURVEILLANCE REQUIREMEN15 The following SRs apply to each PAM instrumentation Function in Table 3.3.3.1-1. s ac h*e ${lo Cfe 50. vs,sn N days, MM m, Gd'h F, , (c ntinued) _
~BWV6-STS- B 3.3-01 Rev. O , -09/2&/92-
_ _ _ . _ _ _ . , . - v . ..m- -+-wr- .u-- we=. 7yr- y-*-i e- +-we
__ _ .. . --_.~__. _ _ _ .. - -- . U-602283 Att. 2 Page 165 or 183 DISCUSSION OF CHANGES TO NUREG-1434 TS: 3.3.3.1 - POST ACCIDENT MONITORING INSTRUMENTATION BRACKETED ADMINISTRATIVE CHOICE B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant specific requirements. PLANT SPECIFIC DIFFERENCE P.1 This comment number is not used for this station. P.2 The details required for the Special Report are moved here for consistency with revisions proposed in Section 5. P.3 The plant specific Type A and Category 1 PAM instruments are
-identified in accordance with the NUREG Reviewer's Note.
P.4 The current required allowed out-of-service time and CHANNEL CALIBRATION frequency for the Hydrogen Analyzers is retained. P.5 The current plant specific number of channels of suppression pool water level and Primary Containment pressure monitoring required is retained. CHANGE / IMPROVEMENT TO NUREG STS C.1 Revisions are proposed to prevent misapplication of the ACTIONS for the PCIV Position Function. With position indication for both PCIVs on a penetration inoperable, Condition C was intended to be applied. However, with only one channel required per valve, two required channels could never be inoperable. Proposed Table Note b and the proposed Bases changes would clarify that only one channel is required if only one PCIV is provided with position indication in the control room. C.2 The Bases are corrected. The reference Specification in the Administrative Controls Section does not require the Special i Reports to be approved by any specific individual or committee. { C.3 Condition F is not applied to reactor vessel water level in the LCO. Therefore, the Bases are revised to match the LCO. C.4 These changes are proposed to revise specific terminology to ) that which is generically preferred for application to the 1 BWR/6 plants. The BWR LCOs do not use the term " train", however, " division" is used in several places. l CLINTON 11 10/1/93 l
U-602283 Att. 2 Page 166 of 183
?m0 CLINTON POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION PACKAGE SECTION 3.4 REACTOR COOLANT SYSTEM
l M U-602283 Att. 2 Page 167 or 103 RCS P/T Limits 3.4.11
-m \
3.4 REACTOR C00LAflT SYSTEM (RCS) l- ; 3.4.11 RCS Pressure and Temperature (P/T) Limits "
) COP LCO 3.4.11 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculationpy =; ste tin + temperatur r- m t shallbemaintainedwithinglimit(ipcc-;f;ei
,, J , m i'!L R. ,
! .,,~~, - C3 APPLICABILITY: At all times. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l A. ---------NOTE--------- A.1 Restore parameter (s) 30 minutes l Required Action A.2 to within limits. shall be completed if this Condition is AND entered. A.2 Determine RCS is 72 hours acceptable for Requirements of the continued operation. l LCO not met in MODES 1, 2, and 3. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met. B.2 Be in MODE 4. 36 hours (continued) l i 1 I l BWR/6 STS 3.4-25 Rev. O, 09/28/92 l
U-602283 Att. 2 Pop 1H of I" RCS P/T Limits 3.4.11 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME i C. ---------NOTE--------- C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) - shall be completed if to within limits. this Condition is entered. AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LC0 not met in other operation. or 3 than MODES 1, 2, and 3. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 ----------
--------NOTE--------------------
Only required to be performed dur S heatyp and cooldawn operatian e , and RCS
~
~ ,
I. II .'!$ ...
, Verify RCS pre 22qre,_RCE tc.7;cr;turc, er.d 30 minutes s s hoa un ancFcoordown rates a...L.
thi l;m;U specified E thE . _ P ,- 6 IcodFy , i
\n on 1 kos t. _
y .- SR 3.4.11.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes j the PTLR. prior to : control rod { withdrawal for l the purpose of l achieving i criticality (continued) BWR/6 STS 3.4-26 Rev. O, 09/28/92 1 I , _
u-sorres Au, 2 Page 169 oP 103 RCS P/T Limits 3.4.11
'q SURVEILLANCE REQUIREMENTS (continued) 3 SURVEILLANCE FREQUENCY SR 3.4.11.3 -------------------NOTE--------------------
Only re onFS 1. 2< 3 {w.-.. . - vi ai. cam uvmu ei m e. . - B) -i. 25 93ig]. ^- Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minute:: pre r moerature prior to each c3 i withni the l im i i.s apuuiiieu in mu FTLL startup of a
-- 4 lgdF' -
recirculation pump SR 3.4.11.4 -------------------NOTE-------------------- Only required to be met in MODES 1, 2, 3, and 4. - Verify the difference between the reactor once within coolant temperature in the recirculation 15 minutes loop to be started and the RPV ol prior to each 7 ,, ' A t g ithin thflirit; pccif;cd startup of a in the PTuh 6 50% recirculation pump SR 3.4.11.5 -------------------NOTE-------------------- l Only required to be performed when. tensioning the reactor vessel head bolting l i studs. l Verify reactor vessel flanan and b 30 minutes nye.__tgnparatures arefiithin' thc 1iinit3 C3 , apsi ficd in the PILR. 'p~ ~JOoF. (continued) 1 l BWR/6 STS 3.4-27 Rev. O, 09/28/92 ,
- -_.. . . . . -- - _ - .- - - - . . _ . - . _- ~. . . .. -
l U 602283 Att. 2 i Page 170 of 183 RCS P/T Limits 3.4.11 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.4.11.6 -------------------NOTE-------------------- Not required to be performed until 30 minutes after RCS temperature s 80*F in MODE 4. i l Verify reactor vessel flan a e 30 minutes l
- n s r wgMn the limits
'g ' '
l specified in t;ie PTLR. pf 70% i 1 - l l SR 3.4.11.7 -------------------NOTE-------------------- Not required to be perfor ntil 12 hours after RCS temperature s in MODE 4. 90'F - W Verify reactor vessel flan e 12 hours ange-.tempecatu ar- w; thin th; lim;t:; O~' spccificd ,n t!.e TTLR. ') 70dF, I INSEPT M A -' g t e D e BWR/6 STS 3.4-28 Rev. O,09/28/92 s.. .. ... - . _ - , - . - . _ - _ , _ . _ _ . . m- _ . _ - . . - . , . ., .-. , . , _ . . . . . , _ . . - _ . , , ,-s,--. -
U+002283 Att. 2 Pega 17A of 183 INSERT 28A 4 SR 3.4.11.8 --------------NOTE------------- Only required to be met in single loop operation with THERMAL POWER s 30% of RTP or recirculation loop flow in the operating loop s 30% of rated flow. Verify the difference between Once within the bottom head coolant 15 minutes temperature and t M V coolant prior to an temperature is(Is ' 100 F. increase in THERMAL POWER or an increase in loop flow SR 3.4.11.9 --------------NOTE------------- Only required to be met in single loop operation with THERMAL POWER s 30% of RTP or recirculation loop flow in the operating loop s 30% of rated flow, and with the idle recirculation loop not isolated from the RPV. Verify the difference between Once within the reactor coolant temperature 15 minutes in the recirculation loop not in prior to an l operation and th coolant increase in l temperature is s 50 F. THERMAL l
, POWER or an c.3 ' increase in loop flow l
1 l l l l l INSERT CLINTON 3.4-28 10/1/93
-- -. - - . . - - , . . - . _ . . ~ . , ~. ..
U-002283 Att. 2
.Page 172 of 103 INSERT 28A (continued)
SR 3.4.11.10 --------------NOTE-------------- Only required to be performed during RCS inservice leak and hydrostatic testing. Verify RCS pressure and 30 minutes temperature are within the limits of the PTLR. INSERT CLINTON 3.4-28 10/1/93
U-002283 Att. 2 Page 173 o9 183 RCS P/T Limits B 3.4.11 WEnc o 2 to U.40et96 ge 2 of 2tl B 3.4 REACTOR COOLANT SYSTEM (RCS) (s o b , 8 3.4.11 RCS Pressure and Temperature (P/T) Limits i BASES l l BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design i assumptions and the stress limits for cyclic operation. The PTLR contains P/T limit. curves forghec oldown, and OC3 dstatic testingend-det; for t.._
- m. . m vi wann v.
u:o m . y .u.. - he heatup curve provides limits for both eatup an criticality. Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to detemine that operation is within the allowable region. The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCP8). The vessel is the component most subject to brittle failure. Therefore, the LC0 limits apply mainly to the vessel. 10 CFR 50, Appendix G (Ref.1), requires the establishment of P/T limits for material fracture toughness requirement s of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during nomal operation, . anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code, Section III, Appendix G (Ref.2). The actual shift in the RTm of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and 10 CFR 50, Appendix H (Ref. 4). The operating P/T limit curves will be adjusted, j y (continued) 1
.JWR/6 RTS B 3.4-51 Rev. O,-09/28792- j
1 U-602283 Att. 2 Page 17'4 of 183 RCS P/T Limits B 3.4.11 ! Enc o ' 602T96 t BASES Se 6 - age 18 'I APPLICABLE are acceptance limits themselves since they preclude SAFETY ANALYSES operation in an unanalyzed condition. (continued) RCS P/T limits satisfy criterion 2 of the NRC Policy Statement. LCO The elements of this LCO are:
,M ,
- a. RCS pressure, temperat atu, nr Idown rata
.[ '
p
-^ -
evy s4/tCS faAwifAinthelimi ea}yp , coo u ou s e 8 M e'n'wc"c'leQ[L9ydes k terb
- b. The tdmperature difference bWween the reactor vessel bottom head coolant and the reacto f (RPV)coolantiswithinthelimithoreunrevessel._C2 i^fLR Yuring MC/'cArci 8
{l N.re circula.tje Rs t a r t upfd'JTcfu M f c. The temperature differencETeEEt eJ renfNNn*' TUttwA L. PoHER. w mfTcfo% nf o/c. rikW f* A l in the respective recirculation 1 and_in the f g eA4* reactor vessel meets the limit startup - during ump Nv 1 cr t (m , nc iu,t:f .,~.-. t
. .. - - :h ac N h
- e. The reactor vessel flange and the head flange -
temperatures are within the limit of A l'!LR when reactor vessel head bolting studs,eee +=ne'anad e 4ps G ' ,g v') ' g These limits define allopable operating regions and permit a i large number of operating cycles while also providing a wide ! margin to nonductile . failure. l The rate of change of temperature limits control the thermal l gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and inservice leak and hydrostatic testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves. Violati'on of the limits places the reactor vessel outside of the bounds'of the stress analyses and can increase stresses l s in other RCS components. The consequences depend on several ! factors, as follows: 1 \ (continued) l l BWR/44T&-- , 8 3.4-53 Rev. O,-00/20/42 l
U-602283 Att. 2 . j Page 179 of 103 q RCS P/T Limits 8 3.4.11 2 to 2t9 ' Section age ' f211 BASES t ACTIONS C.1 and C.2 i (continued) Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Recuired Action must be initiated without delay and continuec until the limits are restored. Besides restoring the P/T limit parameters to within limits, an evaluation is required to detennine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 200*F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline. SURVEILLANCE SR 3.4.11.1 REQUIREMENTS
-Verification that operation is withi 7Pftft limits is required every 30 minutes when RCS pr ., e and temperature conditions are undergoing planned changes. This Frequency
-is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limit.s are specified.in hourly increments, 30 minutes pennits assessment and correction of minor deviations.
and O _n t v 1 ce or heatup cooldow i
- o. m.m isn ia may be discon given in he relevant plant procedure for ending the u.
e
;nsci v i3c icokes c en--
en tne criteria activity are satisfied. This SR has been modified by a Note that requires this Surveillance to be pe rm d o c[u in stem he u and' cooldown operah and =_rikc .k;k:g: an_ hy_ rett 2t:, s (continued) JWR/6 STP B 3.4-56 Rev. 0,-0944/M-
U-602283 Att. 2 Page 176 of 103 RCS P/T Limits B 3.4.11 Enc 3 r t 96 ) BASES 8** "8* SURVEILLANCE SR 3.4.11.2 - REQUIREMENTS (continued) ?. separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor i cri tical . - Perfonning the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality l provides adequate assurance that the limits will not be i exceeded between the time of the Surveillance and the time 1 of the control rod withdrawal. ' 1 SR 3.4.11.3 and SR 3.4.11.4
- 1 Differential temperatures within the applicabl 4%R limits OCB- ensure that thermal stresses resulting from the tup of
{ ( an idle recirculation pump will not exceed design i 1 allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup l i of an idle recirculation loop (Ref. 8) are satisfied. ! Performing the Surveillance within 15 minutes before starting the idle recirculation puma provides adequate assurance that the limits will not ae exceeded between the time of the Surveillance and the time of the idle pump start. An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.11.4 is to compare the temperatures of the operating recirculation loop and the idle loop. SR 3.4.11.3 has been modified by a Note that requires the Surveillance to be met only in MODES 1, 2, 3, and 4%%-
-reaet+r-steamWome= pressure =E 5 psi @ In H0DE 5 the ,
, overall stress on limiting components is lower; therefore, AT limits are not required.
SR 3.4.11.5. SR 3.4.11.6. and SR 3.4.11.7 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits (continued) 9H9/S STS- B 3.4-57 Rev. 0, 49/28f92 t
. , - , . -,- . - - . - - , . . , , , .e-. -- - - - .
U-602283 Att. 2 Page 177 of 103 RCS P/T Limits B 3.4.11 BASES #" SURVEILLANCE SR 3.4.11.5. SR 3.4.11.6. and SR 3.4.11.7 (continued) REQUIREMENTS during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less'than or equal to certain specified values require assurance that these temperatures meet the LCO limits. The flange temperatures must be verified to.be above the limits 30 minutes before and while tensioning the vessel head bolting studs.to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature s 80*F, 30 minute checks of the flange pq temperatures are required because of the reduced margin to , the limits. When in MODE 4 with RCS temperature s "F, monitoring of'the flange temperature is require very , 90 1 ho t n ura the temperatures are withi 4he- imits., cpec:fi n in bie FTLR) The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time
, that the temperature limits could be exceeded. The 12 hour gg Frequency is reasonable based on the rate of temperature Chan9e Possible at these temperatures.
B BBA REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code, Section III, )
Appendix G. 6
- 3. ASTM E 185-82,(My 1702 A- "Skbed frdice for-furci#mec
- 4. 10 CFR 50, Appendix H. fcds G&d,g & je coote s g yb[4t-MJer doe vessets,,,. n pyg
- 5. Regulatory Guide 1.99, Revision 2, May 988.
- 6. ASME, Boile'r d Pressure _ Vessel Code. Jg.ctip XI Appendix E. 7 "rmsieakfaescere Mscs N/ceflu FWturc ToupWss 12epme-Jh ,
- 7. E00-21778-A, December 1978. /o- EN4.5,
- 8. SAR, Section -[15.1.20] . 6' Y' Y 4
I
-BWR-/6-STS- B 3.4-58 Rev. O, 00/20/02 - l l
\
l < _ ~ _ . _ . _ . _ . . _ _ _ . ~ . . _ _ _ _ _ . . _ _ . _ . _ . _ . _ _ _ . . . _ . _ _ _ . _ _ _ . .
l U-602283 Att. 2 l Page 178 or 183 i l
- INSERT B58A SR 3.4.11.8 and SR 3.4.11.9 C3 -
Differential temperatures within the applicabl limits ensure that thermal stresses resulting from increases in THERMAL POWER or recirculation loop flow during single recirculation loop operation will not exceed design allowances. Performing the Surveillance within 15 minutes before beginning such an increase in power or flow rate provides adequate assurance that l the limits will not be exceeded between the time of the Surveillance and the time of the change in operation. An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.11.9 is to compare the temperatures of the operating recirculation loop and the idle loop. Plant specific test data has determined that the bottom head is ; not subject to temperature stratification with natural l circulation at power levels as low as 30% of RTP and with any I single loop flow rate greater than or equal to 30% of rated loop I l flow. Therefore, SR 3.4.11. 8 and SR 3. 4.11. 9 have been modified l by a Note that requires the Surveillance to be met only when l these conditions are not met. The Note for SR 3.4.11.9 further ! limits the requirement for this Surveillance to exclude l comparison of the idle loop temperature if the idle loop is I ( isolated from the RPV since the water in the loop can not be introduced into the remainder of the reactor coolant system. SR 3.4.11.10 Verification that operation is within PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indications available to monitor RCS status. Also, since temperature rate of change O " limits are specified in hourly increments, 30 minutes permits i assessment and cprrection of minor deviations. l l Surveillance for inservice leakage and hydrostatic testing may l be discontinued when the criteria given in the relevant plant ' procedure for ending the activity have been satisfied. l l This SR has been modified by a Note that requires this i Surveillance to be performed only dur4ng inservice leakage and hydrostatic testing. INSERT CLINTON B 3.4-58 10/1/93 l l
- -- - ~ - . .. .. . . . -. ..
U-002283 Att. 2 page 179 of 183 4 s DISCUSSION OF CHANGES TO NUREG-1434 i TS 3.4.11 - RCS PRESSURE AND TEMPERATURE LIMITS i f BRACKETED ADMINISTRATIVE CHOICE l B.1 Brackets removed and optional wording preferences revised to l reflect appropriate plant specific requirements. I PLANT SPECIFIC DIFFERENCE i i 1 P.1 The safety analysis report for this station is identified as I ! the Updated Safety Analysis Report and is correctly referred to l ! as the USAR. I 1
- P.2 Additional Surveillance Requirements are included to l l incorporate current plant specific verifications of single loop J j operation limits.
j ! P.3 Additional information is included in the references for ease J of identification. This additional information may include the title, revision number and/or date.
- P.4 Plant specific design information provided for this Surveillance Requirement.
i i l CHANGE / IMPROVEMENT TO NUREG STS 4 C.1 These changes made to the LCO elements to include when each element applies. i C.2 Editorial information is deleted since it is not required i information and it is not consistent with the format and i content of other proposed Bases. 1 1 ] C.3 Changes made to be consistent with changes proposed in Section a 5. j 4 i i 1 e e d i 4 CLINTON 16 10/1/93
- ll" : ^";.:
{ W i) CLINTON POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION PACKAGE l SECTION 3.8 ELECTRICAL POWER SYSTEMS l I _ _
__ _ _ _ _ _ - ' ' ~ Kw.e p:p :swu3 : Diesel Fuel Oil, Lube Oil, and Starting r Strbsyst-em Ci7 8383 U 602283 Att. 2 BASES e.g. ici er tes SURVEILLANCE SR 3.8.3.2 (continued) REQUIREMENTS the level reaching the manufacturer's recomended minimum level. A 31 day Frequency is adequate to ensure that a sufficient lube oil supply is onsite, since DG starts and run times are closely monitored by the plant staff. SR 3.8.3. 7
~ ~
%Q cil parrQ4lIhA-f*Y"% A Al The t'estsv ustedEhMre a means of determining whether f new fuel oil is of the appropriate grade and has not been C@ contaminated with substances that would have an immediate detrimental impact on diesel engine combustion and operation. If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of
, , ,0Qpj fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage
~
CNNSPWd'% tank (s), but in no case is. the time betweenh i of new Ca6uld ) fuel andx=""5 _^he tests- to exceed 31 days. ptThe -testw_., limits, a cdh%~ O # g e__ ASTM Standardspare as fol_lg pg g & S6ne a. Sampie h 4esfS IiSWMunsd MOCesfri Aomw & uei oiF in acconraneeMttr-AS-TM MM h (Ref. 6); g g,.4 W5k SI D2.10 - m s 30 e
- b. Verify in accordance with the tests specified in ASTM D975-J 8e) K(Ref. 6) that the ample has an absolute specific gravity at 60/60*F of a 0.83"and 5 0. #(or 8
- a n d 5 49 * ) , a ""
q an API gravity at 60*F of akinematic viscosity at 40*C of a 1. 5 4.1 centistokesf =h' ; '!ssh wint etuad$; and
- c. Verify that the new fuel oil has a clear and bright appearance with pro er color whert t_ested _ irt ucordar/C '
't ASTM 04 g P (Ref. 6)far a de dMh+# Y
% 4mM ongccu-k_ wldAsim-o VI%-82 yW Failure to meet any of Ene aDove-iTmTts is cause7or rejecting the new fuel oil, but does not represent a failure to meet the LCO since the fuel oil is not added to the storage tanks.
OC43 =+iithir 31 deys Nilowing the' initial new fuel oil sample, the fuel oil is analyzed to establish that the other , (continued) {. 4WRM-SM- B 3.8-47 Rev. 0;-09f28/92
v- 7 r U-002283 Att. 2 Enclos o mis)' Sec% Page ' of 25[ Pegs 102 of 183
~
INSERT B48A additional analyses are required by Specification g,$,g-5.7.14 Diesel Fuel Oil Testing Program, to be performed in 31 days following sampling.and addition. This 31 days is intended to assure: 1) that the sample taken is not more than 31 days old at the time of adding the fuel oil'to the storage tank, and 2) that the results of.a new fuel oil sample (sample obtained prior to addition but not more than 31 days prior to) are obtained within 31 days after addition. INSERT M NTON B 3.8-48 10/1/93
Distribution Systems-Operating Enclosure . ?t96 8 3.8.9 i Sectio age 23 '7 ( U-6022e3 Att. 2 BASES Paoe 183 or ie3 ( ACTIONS AM (continued) control centers, and distribution panels must be restored to OPERABLE status within 8 hours. The Condition A worst scenario is one division without AC power (i.e., no offsite power to the division and the associated DG inoperable). In this Condition, the unit is more vulnerable to a complete loss of AC power. It is, therefore, imperative that the unit operators' attention be focused on minimizing the potential for loss of power to the remaining division by stabilizing the unit, and on restoring power to the affected division. The 8 hour time limit before requiring a unit shutdown in this Condition is acceptable because:
- a. Inere is potential for decreased safety if the unit operators' attention is diverted from the evaluations and actions necessary to restore power to the affected division to the actions associated with taking the unit to shutdown within this time limit,
- b. The potential for an event in conjunction with a
, single failure of a redundant component in the
( division with AC power. (The redundant component is ) i verified OPERABLE ir accordance with ~ Specification " afet tion Determination i Program (SFDP)Q.") 7.! ' 5 , 5 .cl The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition A is entered while, for instance, a DC bus is inoperable and subsequently returned OPERABLE, the LCO may already have been not met for up to 2 hours. This situation could lead to a total duration of 10 hours, since initial failure of the LCO, to - restore the AC distribution system. At this time, a DC circuit could again become inoperable, and AC distribution could be restored ' OPERABLE. This could continue indefinitely. This Completion Time allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
This results in establishing the " time zero" at the time the (continued) nuore ere B 3.8-82 Rev. O, Mf28/92--
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