ML19323C763
| ML19323C763 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/10/1975 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323C736 | List: |
| References | |
| FOIA-80-162 NUDOCS 8005190149 | |
| Download: ML19323C763 (34) | |
Text
- - - - - - - - -
8005190/49 a
4 V:
CO' Sit'EPS PCFFF. CEPA?;Y POCKET ';0. 50-155 i
BIG P.DCK P01T PUM NIEND:L',T TG FACILITY CPliR.\\TII;G LICE';5E Anend. rent !;o. 9 Licenso tio. DP:( 6 i
1.
The ;.uclear.(c ulatory Cotmission (the Con.aission) has found that; j
o A.
The application for ar.:endcent by Consumers Tower Conpany (the licensee) dated January 13, 1975, coaplies with the standards and requirements of the Ator.ic Energy Act of 1
i 1954, as arcended (the Act), and the Co::.nission's rules and regulations set forth in 10 CFR Chapter I; l
11 onstruction of the Big Rock Point Plant (the facility) has been substantially conpleted in confornity with Construction l
Pernit t;o. CPPR-0 and the application, as anended, the provisions of the Act and the rules and regulations of the Co=.ission:
1 C.
'Ihe facili.ty will operate in conformity with the application.
as arended, the previsions of the Act, and the ruit. and j
I regulations of the Cor:nission:
I l
D.
- hcre is reasonabic assurance-(i) that the activitics nuthorized j
by this operating license can be conducted without endangering, the health and safety of the public, and (ii) that such activities will be conducted in corrpliance with the rules and regulations of the Cox.ission;
.E.
The licensee is technically and financially qualified to enga.c l
in the activities authori cd by this operatin,., license in acccrd-l ance with the rules and rei.ulations of the Co:::aission, F.
'ine licensec has satisfied tne applicable provisions of 10 CFR l
Part 140, Financial Protection Gequircaer.ts and Inde: nity Ai;ree-ments, of the Co.z.ission's rec,ulations ;
i l
I i
l 1
, orrec e >
evenain s h cats k h
Form AEC-)l3 (Rev. 9 53) AECM 0240 W u. s. nove==wswr pasarino orraces ser.eae.ses
{
3 e
/, /
.s G.
T.:e issu::nce of this operatin,; license will not be ini..:ical y
to the cct.r.on defense and security or to the health nnd safety
/
of the public, and H.
'!he receipt, possession. end use of source, hyproduct and specini nucicar natorini ns authorized by ti.is license vill be in accordance with the Comission's regulations in 10 i
CFF. Parts 30, 40. and 70, including 10 CT'l Sections 30.33,
- 40. 32, and 70.23 and 70.31.
2.
Facility Operating 1.icense No. !;PR-6, issued to the Consuncrs Tower Cox.pnny, is hereby amended in its entirety to read as follois:
A.
Titis license applies to the ' Jig noch Point Plant, a boiling water reactor and associate i equip =cnt (thc facility), ouned by* located in Charlevoi:(the Constraers Power Coupany (the licensee). The facility is Cotuity. :iichigan, and is described in the licensce's application dated January-14, 1960, and ti.c Final liazards Stur.2:ry Report, as suppleic.cnted and auended by subsequent filings by the licensee.
{
J.
Subj ect to the conditicas and requirer.ents incorporated hercin, i
the Coraission hereby licenses Constusers Power Coupany:
(1)
Pursuant to Section 104b of tihe Act and 10 CFR Part 50,
" Licensing of Production and Utilization Facilitics,"
to possess, use, and operate the facility at the designated location in Charicyoix County,141chigan,
~
in accordance with the procedures and limitations set forth in this license; l
(2)
Pursuant to the Act and 10 CFR Part 70, Special Nuclear
, 'aterial," to receive, possess, and use at any one tino up to (n) 1200 kilograms of contained uranitn, 235 os fuel,
./
(b) 10.32 grams of urenlum 235 as contained in fission counters, (c) 150 kilograns of pitttonitr, contained in Pu0.,-1*0,,
~
fuel rods. and (d) 5 curies of plutoniua encapsulated as 5 plutonitr.-berylliun neutron source, all in connection with orcration of the facility:
6 i
orrec a p I
SUAM&Ma k D&Ts b Form Mc.Jle (Rev, p.33) MCM 0240
- u. a.oovanmusar paintene orreces son.sae.ses
i
. (3)
Pursuant to the Act and 10 CFR Part 30, ' Tules of General N
Applicability to the Licensing of typroduct ::sterial," to
\\
receive, possess and use at any one tire up to 7000 curies of anti: ony-beryllitn in the form of neutron sources:
(4)
Pursuant to the Act and 10 CFR Part 40, "!.icensing of Source "aterial," to receive, possess and use at any one tine up to 500 kilogra-s of depicted uraniun dioxide contained in the facility's fuel assemblics:
(5)
Pursuant to the Act and 10 CFP. Parts 30 and 70, to possess,
but not separate, such byproduct and special nucicar materials 4
as may be produced by the operation of the facility.
C.
This license shall be decced to contain and is subject to the conditions specified in the following Coc's.ission renulations in l
10 CFR Chapter I; Part 20, Section 30.34 of Part 30 Section 40.41 of Part 40$ Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70: and is subject to all applicable pro-visions of the Act and to the rules, regulations, and orders of the Coxnission now or hereafter in effect; and is subject to j
the additional conditions specified or incorporated below-l (1) Maxicaun Power _I.cyc1 The licensec is authorized to operate the facility at steady state reactor core power levels not in excess of 240 megawatts (thermal).
(2)
Techni.c.al Spec.i_f_ic.a._t.i._o_n_s The Technical Specifications contained in Appendix A as issued Itay 1, 1964, as revised, are hereby incorporated in this license. The licensco shall operate the facility
{
in accordance trith the Technical Specifiestions as revised
)
by issued chan.qcs thereto through Chan;c No. 46.
l 1
4
- intrecs >
l summa nes >
i mars k F
LIC-)l3 (Rev. 9-3)) AICM 0240 W u. e. sovanquant emisstems errects sota.saa see
4-D.
This anended license becomes effective 30 days after the date of its issuance and shall expire at nidnight, ?'ay 31, 2000 FOR Tl!E NUCLEAR REGifLATORY C0'NISSION kkchkc(.J 27O a dM(q nr Dennis L. Ziemann, Chief i Operating Reactors Eranch #2 Division of Reactor Licensing
Attachment:
Change No. 46 to the Technical Specifications l
Date of Issuance:
l r-~
.?:
h
)
I oprecs >
- evana.es >
oave >
Pens ABC.310 (Rev.SS)) AECM 0240 W v. s. sovanament enenvia.s earscas seva.ese.see
t A-y
-. 3:....
.,s. I
- -- - - :. c -
u
=-n
,i
- t.,. ::
~ "3 vo' 'n
~&
I C'I .
~v.
.g.,.p
'l:..,P.?cf, e..
.,p g..,s.
~ _. J. t
... 3 y
~~~~
TACILI--- TY O P~3 n...
..r.. L z c oc r.
i)
- 0. DP.R 6 pp..r- -.,
r
~~)..
oc.j75 3
pt g
Instructions top 4.- a. :porating Change \\'o.
- 6-I*
RcP1nce cxist I'i 1
addi tional.'.. *.P^i'o 111 with attache 4
'~~3cs).withsid's:.,S2 sed page iii and A
the mur'.in-Changes are 1.je..I,-
P"
\\.
2 2 Incory.> rate,sg5.E isidenc5g3ef701' -lthrough3037, (ya, \\ e nct' Pucc 2n the Iv c. corner' " @_ 'O. 46 3-nletethefo]j'h'.}nsSection5 catire]y'. 7 y, 7.1.1 - 7.1*10* e 7.2, 7.2 ] - -- 7.5, 7,3,y. !.a.6 and 7,7, '--), lr' ) l 1 i. e 4 g f 5 e i J. o 9 E'
CONTENTS (Contd) Page No. 7-1 7.0 Operating Procedures.. 7.1 Deleted. 7.2 Deleted....... 7.2.1 Deleted... 7.2.2 Deleted. 7.2.3 Deleted..... ~................................... 7-3 7.2.4 Administrative and ' Procedural Controls Relating to High Performance Fuel. 7-3 7.3 Normal Operation 7.3.1 General....................... 7-3 7-3 7.3.2 Cold Start-Up After Extended Shutdown............ 7-4 7.3.3 Hot Start-Up,down.. 7-5 Extended Shut 7.3.5 7-6 7.3.6 Short Duration Shutdown................................ 7-6 7.4 Refueling Operation... g 7-8 i 7.5 Maintenance (Deleted Except for 7.5.7) 7-8 7.6 Operational Testing of Nuclear Safeguard Systems 7.7 Deleted........................................ l 8-1 8.0 Research and Development Program (Phase II) 8-1 8.1 Fuel Irradiati-. Program 8-1 8.1.1 Development Fuel Design Features 8-5 8.1.2 Instrumented Assembly D'esign 8-5 8.2 Performance Testing. 8-5 8.2.1 Core Performance and Transient Tests 8.2.2 Sequence of Testing. 8-6 8-8 8.2.3 Analysis of Typical Tests............................... 8-11 8.3 Reactor Operating Limits 8-12 8.* 4 Operating Procedure.................................. 8-12 8.5 Special Review Procedures......... 9-1 9.0 Primary System Surveillance 10-1 10.0 (Section 6.0) Administrative Controls 10-1 6.1 Responsibility 10-1 6.2 Organization 10-1 6.2.1 Off Site 10-1 6.2.2 Plant Staff..................................,,.. 10-1 6.3 Plant Staff Qualifications 10,5 6.4 Training 10-5 6.5 Review and Audit iii
CONTENTS (Contd) Par,e No. 6.5.1 Plant Review Committee (PRC) 10-5 6.5.2 Safety and Audit Review Board (SARB) 10-7 6.6 Reportable Occurrence Action 10-11 6.7 Safety Limit Violation 10-11 6.8 Procedures 10-11 6.9 Reporting Requirements 10-12 6.10 Record Retention 10-20 6.11 Radiation Protection Program 10-22 6.12 Respiratory Protection Program 10-22 u e N,! e iv
~ 6.0 ADMINISTRATIVE CONTROLS (NOTE: 'Ihis is the new fomat of the Technical Specifications to be issued d entirely in the near future. This section replaces the foIIowing sections of the current Technical Specifications: 7.1, 7.1.1-7.1.10, 6.1 RESPONSIBILITY 7.2, 7.2.1-7.2.3, 7.5, 7.5.1-7.5.6, and 7.7 entirely) 6.1.1 The Plant Superintendent shall be responsible for overall plant operation end shall delegate in writing the succession to this responsibility 9_minc his absence. 6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for plant management and technical support shall be as shown on N-Figure 6.2-1. PLANT STAFF 6.2.2 The plant organization shall be as shown on Figure 6.2-2 and: Each on-duty shift shall be composed of at least the minimum shift crew composition shown a. in Table 6.2-1. At least one licensed Operator shall be in the control room when fuel is in the reactor. b. At least two licensed Operators shall be present in the control room during reactor startup (to a power level >5 percent), scheduled reactor shutdown and during recovery from reactor c. trips. is An individual qualified in radiation protection procedures shall be on site when fuel d. g in the reactor. All core alterations after the initial fuel loading shall either be performed by a lictnsed Operator under the supervision of a licensed Senior Operator or a no e. Fuel llandling) who has no other concurrent responsibilities during this operation. 6.3 PLANT STAFF QUALIFICATIONS Each member of the plant staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 6.3 1 i for comparable positions. 10-1 Change No. 16
i CONSUMERS POVER COMPANY OFF-SITE ORGANIZATION ~, PRESIDENT EXECUTIVE VICE PREST 0ENT ib- } } i VICE PRESIDENT BULK POWER OPERATI.ONS i i g Safety and Audi t j EXECUTIVE MANAGER Review BULK POWER OPERATIONS Board ~~ I I i I Manager of Hanager of Director Others Hanager Production - Nuclear Maint & Quality Operating Services Adm Services Assurance g., General Supt QA Adm Services Admin-Plant . i s t ra tor Superintendents (Nuclear) Nuclear Training Administrator Figure 6.2-1 m.
{ CONSUMERS POWER COMPANY BIG ROCK POINT Organization QA-ADMIN OFF-SITE PLAMT SUPERINTENDENT PLANT REVIEV COMMITTEE CLERKS QA ENGlHEER ADMINISTRATIVE SUPERVISOR OPERATIONS SUPERINTENDENT TECHNICAL SUPERINTENDENT MAINTENANCE SUPERINTENDENT Sm SECURITY SUPERVISOR REACTOR ENGINEER IEC SUPERVISOR. ENGINEER TRAINING S.S.: PLANT SHIFT SUPERVISORS ASST MAINTENANCE SUPERVISOR OPERATORS CHEM & RAD PROT SUPV' ENGINEER' lREPAIRME!1l [ CHEM & R h liOT SUPifR i CHEM & RAD PROT TECHNICIANS SOL - Senior Operator Licensed in accordance with 10 CFR 55. FIGURE 6.2-1 10-3 1, diange No. 4fi
TABLE 6.2-1 Minimum Shift Crew Composition The minimum shift crew shall be as follows except when plant conditions specified in paragraph (a) or (b) below have been established (see note (1) below) or when an unexpected absence occurs (see note (2) below): 1 Shift Supervisor - SOL 2 Operators - OL 2 Operators - Nonlicensed \\- (a) Cold Shutdown 1 Shift Supervisor - SOL 1 Operator - OL 1 Operator - Nonlicensed f (b) Refueling Operations (See note (3) below) 1 Shift Supervisor - SOL 1 Operator - OL 2 Operators - Nonlicensed SOL - Senior Operator Licensed in accordance with 10 CFR 55 OL - Operator Licensed in accordance with 10 CFR 55 (1) During control rod motion associated with reactor start-up (to a power Icvel > 5 percent), one 1icensed Operator shall observe the control rod manipulation to ensure established control rod withdrawal procedure: are adhered to. (2) In the event that any member of a minimum shift crew is absent or incapacitated due to illness or injury, a qualified replacement shall report onsite within two hours. ) (3) Does not include additional personnel required when core alterations are being conducted. See 6.2.2.e. 10-4 Change No. M
w ( ( ~. 6.h TRAINING 6.h.1 A retraining and replacement training program for the plant staff shall be maintained under the direction of the Nuclear Training Administrator and shall meet or exceed the requirements and recommendations of Section 5 5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR, Part 55. 6.5 REVIEW AND AUDIT 6.5 1 PLANT REVIEW COMMITTEE (PRC) g 6.5 1.1 FUNCTION The Plant Review Committee (PRC) shall function to advise.the Plant Superintendent on all matters related to nuclear safety. 6.5 1.2 COMPOSITION The PRC shall be composed of the: Chairman: Plant Superintendent Member: Operations Engineer Member: Technical Engineer Member: Maintenance Engineer Member: Plant Instrument and Control Supervisor Member: Reactor Engineer Member: Chemistry and Radia? ion Protection Supervisor 4 Member: Shift Supervisor (One) Member: Engineer With at least One-Year Plant Experience 6.5.1.3 ALTERNATES Alternate members shall be appointed in writing by the PRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate in PRC activities at any one time. 10-5 Change No. 46
W ( ( 6.5.1 (Cont'd) 4 6. 5 1. 14 MEETING FREQUENCY The PRC shall meet at least once per calendar month with special PliC meetings ac required. 6.5.1 5 QUORUM ~ A quorum of the PRC shall consist of the Chairman and four members (including alternates). 6.5 1.6 RESPONSIBILITIES 4 The PRC shall be responsible for: Review of (1) all procedures required by 6.8 and changes thereto, a. (2) any other proposed procedures or changes thereto as determirad by the Plant Superintendent to affect nuclear safety. b. Review of all proposed tests and experiments that affect nuclear safety. Review of all proposed changes to the Technical Specifications. c. d. Review of all proposed changes or modifications to plant systems or equipnent that affect nuclear safety. e. Investigation of all violations of the Technical Specifications. A report shall be prepared and forwarded covering evaluation and recommendations to prevent recurrence to the Manager of Production, Nuclear and to the Chairman of the Safety and Audit Review Board (SARB). f. Review of plant operations to detect potential safety hazards, g. Performance of special reviews and investigations and reports thereon as requested by the Chairman of the SARB. h. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Dnergency Plan to the Chairinan of the SARB. Chatirc Nei. &
i ( ( 6.5.1 (Cont'd) 6.5.1 7 AUTHORITY The PRC shall: s a. Recommend to the Plant Superintendent written approval or disapproval of items considered under 6.5.1.6(a) through (d) above., b. Render deteminations in writing with regard to whether or not each item considered under 6.51.6(a) through (f) above constitutes an unreviewed safety question. c. Provide inanediate written notification to the Manager of Production, Nuclear and the Chairinan of SARB of disagreement between the PRC and the Plant Superintendent. However, the Plant Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above. 6.5 1.8 RECORDS The PRC shall maintain written minutes of each meeting and copies shall be provided to the Manager of Production, Nuclear, the Chairman of SARB, PRC members and alternates. 6.5 2 SAFETY AND AUDIT REVIEW BOARD (SARB) 6.5 2.1 RESPONSIBILITIES SARB is responsible for maintaining a continuing examination of designated plant activities. In all cases, where a matter is fomally considered by SARB, its findings and recomunendations are coausunicated. in writing to the Executive Manager - Bulk Power Operations (BPO) and other appropriate levels of management. A written charter is prepared and approved by the Executive Manager - BPO which designates the membership, authority and rules for conducting the meetings. Board membership, qualifications, meeting frequency, quorum, responsibilities, authority and records are in accordance with the nuclear plant Technical Specifications and ANSI N18.7-1972. 6.5 2.2 FUNCTION The SARB shall function to provide independent review of designated activities affecting safety-related components, systems and structures designated on the plant's Safety-Related Quality List contained in the Consumers Power Conapany Quality Assurance Program. 10-7 Change No. 46 I
_ x_ - _ l ( t 6.5.2 (Cont'd) 6.5.2.3 COMPOSITION AND QUALIFICATIONS Collectively, the personnel appointed for the SARB by the Executive Manager - BPO shall be competent to conduct reviews and technical audits in the following areas: Nuclear power plant operations. j a. b. Nuclear engineerir.g. l Chemistry and radiochemistry. l c. d. Metallurgy. l j e. Instrumentation and control. f. Radiological safety. g. Mechanical and electrical engineering. h. Quality Assurance practices. An individual appointed to the SARB may possess expertise in more than one of the above specialties. He should, in general, have had professional experience at or above the senior engineer level in his specialty. 6.5.2.4 ALTERNATE MEMBERS Alternate members may be appointed by the Executive Manager - BP0 to act in place of members during any legitimate and unavoidable absences including a conflict-of-interest determination. The qualifications of alternate members shall be similar to those members for whom they vill substitute. 6.5 2.5 CONSULTANTS Consultants shall be utilized as determined by the SARB members and/or chairman to provide expert advice to the SARB. SARB members are not restricted as to sources of technical input and may call for separate investigation from any competent source. 10-8 Change No. h
( ( ( 6.5.2 (Cont'd) 6.5 2.6 MEETING FREQUENCY The SARB shall meet at least once per calendar qursrter during the initial ye:ar of facility operation following fuel loading and at least once every six months thereafter. 6.5 2.7 QUORUM A quorum of SARB shall consict of the Chairman or his designated alternate and four (1+) { members or their alternates. No more than a minority of the quorum shall have line responsibility for operation of the facility. It is the responsibility of the Chaiman to ensure that the quorum convened for a SARB meeting contains appropriately qualified members or has at its disposal consultants sufficient to carry out the review functions required by the meeting agenda. 6.5 2.8 REVIEV t ,The SARB shall review: Proposed tests or changes to procedures, equipment, systems which are deemed to involve an a. unrevLeved safety question as defined in 10 CFR 50 59 b. Proposed changes in Technical Specifications or licenses. Significant operating abnormalities or deviations from normal and expected performance c. of P ant equipment that affect nuclear safety. l All events which are required by regulatio..s or Technical Specifications to be reported d, to NRC in writing within 24 hours and other violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements or of internal pro-cedures or instructions having nuclear safety significance. Reports and meet ing minutes of the pRC including safety evaluations for changes to pro-c. cedures, equipment.or systems and tests or experiments completed under the provisions of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question. Operational and major modification Quality Assurance Program audit reports, f. g. Technical audit reports. 10-9 Change No. 16
G.M.8 (Ccna'd) h. Tho status of deficiencies identified by the Quality Assurance Program, including the effectiveness of the corrective actions completed and implewented, at least once every six (6) months.
- i. Audits of the Security Program required by the " Nuclear Power Plant Secur'ity Plan."
652.9 AUDITS Audits of safety-related facility activities during operations are performed by the Quality Assurance Department - BPO in accordance with the policies and procedures of the Consumers Power Company Quality Assurance Program. Quality assurance audit reports are sent to SARB for review. In addition, technical audits are the responsibility of the Technical Services Depart-4 ment and shall be reviewed by SARB. These technical audits encompass: a. The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license conditions at least once per year. b. The performance, training and qualifications of the entire facility staff at least once per year. c. The facility Site Emergency Plan and implementing procedures at least once per two years. d. Any other area of facility operation considered appropriate by SARB or the Vice President - BPO. 6.5 2.10 AUfHORITY SARB shall report to and advise the Executive Manager - BP0 on those areas of responsibility specified in 6.5.2.8 and 6.5.2.9. t 6.5 2.11 RECORDS-Records of SARB activities shall be prepared and distributed as indicated below: a. Minutes of each SARB meeting shall be prepared and forwarded to the Executive Manager - BPO and each SARB member within fourteen (14) days following each meeting. Minutes shall be approved at or before the next regularly scheduled meeting following distribution of the minutes. b. If not included in SARB meeting minutes, reports of reviews encompassed by Section 6.S.2.3 above shall be prepared and forwarded to the Executive Manager e BPO within fourteen (14) days following completion of the review. 10-10 Change No. 46 -g ,,,m
6.5.2 (Cent'd) Audit reports encompassed by 6.5.2.9 above, shall be forwarded to the Executive Manager-c. BPO and management. positions responsible for the areas audited within thirty (30) days after completion of the audit. 6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken in the event of a reportable occurrence: The Commission shall be immediately notified pursuant to 6.7 or a report submitted pursuant a. to the requirements of 6.9. All events which are required by regulation or Technical Specifications to be reported to the NRC in writing within 24 hours shall be reviewed by the PRC. Ihe results of the PRC review j b. i shall be submitted (either by PRC minutes or by separate report) to SARB and the bunager of Production, Nuclear. 6.7 SAFETY LIMIT VIOLATION The following actions shall be taken in the event a Safety Limit is violated: 6.7.1 The reactor shall be shut down immediately and not restarted until Commission authorization a. is received [10 CFR 50.36(c)(1)(1)]. 1 The Safety Limit violation shall be reported immediately to the Commission in accordancc b. with 10 CFR 50.36, the Manager of Production, Nuclear and to SARB Chairman or Vice-Chairman. A report shall be prepared in accordance with 10 CFR 50.36 and 6.9 of this specification. c. The Safety Limit violation and the report shall be reviewed by the PRC. The report shall be submitted within 10 days to the Commission (in accordance with the d. requirements of 10 CFR 50.36), SARB Chairman and the Manager of Production, Nuclear. 6.8 PROCEDURES Written procedures shall be established, impicmented and maintained for all structures, systems, These procedures shall 6.8.1 components and safety actions defined in the Big Rock Point Quality List. meet or exceed the requirements of ANSI-18.7. l l 4 Change No. 46 10-11 ~ ~
6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewei
- by the PRC (except for security procedures which are reviewed by the Company Security Ucpartmua,
and approved by the Plant Superintendent prior to implementation. 6.8.3 Temporary changes to procedures of 6.8.1 above may be mace provided: a. The intent of the original prc cedure is not altered, b, The change is approved by two members (or designated alternates) of the PRC, at least one of whom holds a Senior Reactor Operator's License. The change is documented, reviewed by the PRC at the next regularly scheduled meeting and c. approved by the Plant Superintendent. y, 6.9 REPORTING REQUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted. 6.9.1 Routine Reports a. Startup Report. 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 plant. The report shall address each'of the tests identified in the llazards Summary Report and shall in general include a des-cription of the measured values of the operating conditions or charact6ristics obtained during the test program and a comparison of these values with design predictions and specifications. 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. Startup reports shall be submitted within (1) 90 days following completion of the senrtup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test progran, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed. 10-12 Change No, 46
~- 6.9.1 (Cont'd) b. Annual Operating Report Routine operating reports covering the operation of the unit during the previous calendar year should be submitted prior to March 1 of each year. The initial report shalJ be s.ubmitted prior to March 1 of the year following initial criticality. ^ The annual operating reports made by licensees shall provide a comprehensive summary of the operating experience gained during the year, even though some repetition of previously reported information may be involved. References in the annual operating report to previously submitted reports shall be clear. Each annual operating report shall include: (1) A narrative summary of operating experience during the report period relating to safe operation of the facility, including safety-related maintenance not covered in 6.9.1.b.(2)(c) below. (2) For each outage or forced reduction in power ! of over twenty percent of design po'er level 1 where the reduction extends for greater than four hours: (a) the proximate cause and the system and major component involved (if the outage or " forced reduction in power involved equipment malfunction); (b) a brief discussion of (or reference to reports of) any reportable occurrences pertainin'g to the outage or power reduction; (c) corrective action taken to reduce the probability of recurrence, if apg7ropriate; J/ The term " forced reduction in power" is normally defined in the electric power industry as the occurrence of a component failure or other condition which requires that the load on the unit be reduced for corrective action immediately or up to and including the very next weekend. Note that routine preventive maintenance, surveillance and calibration activities requiring power reductions a're not covered by this section. 1 10-13 Change No, o
~ U.W.LFN oparcting time gst es a result of thn outags or power reduction (for scheduled or (d) J forced outages,- use the generator off-line hours; for forced reductions in pow 2r, use the approximate duration of operation at reduced power); (e) a description of major safety-related corrective maintenance performed during the outage or power ceduction, including the system and component involved and identifica-tion of the critical path activity dictating the length of the outage or power reduction; and (f) a report of any single release of radioactivity or adiation exposure specifically associated with the outage which accounts for more than 10% of the allowable annual values. g, V (3) A tabulation on an annual basis of the number of station, utility and other personnel ~ (includingcontractors)receivingexposuresgreaterthan100mregyrandtheir associated man rem exposure according to work and job functions.-- e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions Small exposures may be estimates based on pocket dosimeter, TLD, or film badge measurements. In the totalling less than 20% of the individual total dose need not be accounted for. aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions. Indications of failed fuel resulting from irradiated fuel examinations, including eddy (4) tests, ultrasonic tests, or visual examinations completed during the report period. currdnt c. Monthly Operating Report Routi..e reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Office of Inspection and Enforcement,U. S. Nuclear Regulatory Commission, Washington, 20555, with a copy to the appropriate Regional Office, to arrive no later than the tenth D. C. of each month following the calendar month covered by the report. 1 2/ The teira " forced outage" is normally defined in the electric power industry as the occurrence of a component failure or other condition which requires that the unit be removed from service for corrective action immediately or up to and including the very ner.t weekend. J This tabulation supplements the requirements of 220.407 of 10 CFR Part 20. 10-14 Change No 46
6.9.2 R;portcbis Occurrenc;s R;porttblo cccurr:nces, including corrective actions and measures to prevent reoccurrence shall be reported to the NRC. final resolution of occurrence. Supplemental reports may be required to fully describe In case of corrected or supplementa'l reports, a licensee event report shall be completed and reference shall be,made to the original report date. Prompt Notification With Written Followup .a. The types of ' events listed below shall be reported as expeditiously as possible, but within 24 hours by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the appropriate Regional Office, or his designate no later than the first torking day following the event, with a written followup report within two vecks. 1 report shcIl include, as a minimum, a completed copy of a licensee event report form.The written followup Information, provided on the licensee event report form shall be supplemented, as needed, by additional narrative i I material to provide complete ex'planation of the circumstances surrounding the event. (1) Failure of the reactor protection system or other systems ' subject to limiting safety system settings to initiate the required protective function.by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function. Note: Instrument drift discovered as a result of testing need not be reported under this
- item but may be reportable under 6.9.2.a(5), (6), or 6.9.2.b(1) below.
(2) Operation of "the unit or affected systems when any parameter or operation subject to a limi condition is less conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications. Note: If specified action is taken when a system is found to be operating between the most conservative and the least conservative aspects of a limiting condition for operation listed in the technical specifications, the limiting condition for operation is not 5 considered to have been violated and need not be reported under this item, but it may be reportable under 6.9.2.b(2) below. (3) Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment. Note: Leakage of valve packing or gaskets within the limits for identified leakage set forth in teclutical specifications need not be reported under this item. 10-15 Changu No. 4b T
6.9.2 (Cont'd) \\ (4) Reactivity anomalics involving disagreement with the predicted value of reactivity balance ~ under steady state conditions during power operation, greater than or equal to 1% Ak/k; a calculated reactivity balance indicating a shutdown margin less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if sub-critical, an unplanned reactivity insertion of more than 0.5% Ak/k or occurrence of any unplanned criticality. (5) Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the Final llazards Summary Report (FilSR). (6) Personnel error or procedural inadequacy swhich prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents y analyzed in the SHSR'. Note: For 6.9.2.a(5) and (6) reduced redundancy that does not result in a loss of system function need not be reported under this section but may be reportable under 6.9.2.b(2) and (3) below. '.7) Conditions arising from natural or man-made' events that, as a direct result of the event ' require plant shutdown, operation of safety systems, or other protective measures required by technical specifications. ~~55 Errors discovered in the transient or accident analyses or in the methods used for such
- Talyses as described in the safety analysis report or in the bases for the technical Lpecifications that have or could have permitted reactor operation in a manner less rAbservative than assumed in the analyses.
(9) Performance of structures, systems, or components that requires remedial action or corrective ~ measures to prevent operation in a manner less conservative than assumed in the accident
- L analyses in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or technical specifications that require remddial action or corrective measures to prevent the existence or development of an unsafe condition.
Note: This item is intended to provide for reporting of potentially generic problems. l 10-16 Change No, 46
@.O.B (Csntd)~ \\ s \\ b. Thirty Day Written Reports The reportable occurrences discussed below shall be the subiect of written reports to the Director of the appropriate Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event. (1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those establishco by the technical specifications but which da not prevent the fulfillment of the functional require ~ents of affected systems. \\ (2) Conditions leading to operation in a degraded mode permitted by a limiting condition for 'g operation or plant shutdown required by a limiting condition for operation. { Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in 6.9.2.b(1) and (2) need not be reported except where test results themselves reveal a de' raded modo as described g above. (3) Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems. (4) Abnormal degradation of systems other than those specified in 6.9.2.a(3) above designed to contain radioactive material resulting from the fission process. Note: Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item. 10-17 Change No. 46
6.9.3 Unique Reporting Requirements a. Radioactive Effluent Release A report shall be submitted to NRC within 60 days after January 1 and Ju?y I of each \\ year specifying the quantity of each of the principal radionuclides released to unrestricted g areas in' liquid and gaseous effluents during the previous 6 months. The format and content s of the report shall be in.accordance with Regulatory Guide 1.21 Revision 1 dated June 1974. \\, (1) Gaseous Effluents (a) Gross Radioactivity Releases i u (i) Total gross radioactivity (in curies), including noble and activation gases released. (ii) Maximum gross radioactivity release rate during any one-hour period. (iii) Total gross radioactivity (in curies) by nuclide released, based on representative isotopic analyses performed. (iv) Percent of technical specification limit. (b) Iodine Releases (i) Total iodine radioactivity (in curies) by nuclide released, based on representative isotopic analyses performed. (ii) Percent of technical specification limits for I-131 released. W.- (c) Particulate Releases (i) Gross radioactivity 01, y) released (in curies) excluding background radioactivity. (ii) Gross alpha radioactivity released (in curies) excluding background radioactivity. (iii) Total gross radioactivity (in curies) of nuclides with half-lives greater than eight days. (iv) Percent of technical specification limit for particulate radioactivity with half-lives greater than eight days. 10-18 Change No. 4b
6.9.3 (Cont'd) (2) Liquid Effluents (a) Gross radioactivity @, y) released (in curies) excluding tritium and average concentration released to the unrestricted area. (b). Total tritium and alpha radioactivity (in curies) released and average concentration released to the unrestricted area. (c) Total dissolved gas radioactivity (in curies) and average concentration released to the unrestricted area. (d) Total volume (in liters) of liquid waste released. ( (e) Total volume (in liters) of dilution water used prior to release from the restricted area. (f) The maximum concentration of gross radioactivity 6, y) released to the unrestricted area (averaged over the period of release). (g) Total radioactivity (in curies) by nuclide released, based on representative isotopic analyses performed. (h) Percent of technical specification limit and 10 CFR Part 20 concentration limits for unrestricted areas. (3) Solid Waste is) The total amount of solid waste packaged (in cubi-feet). (b) The total estimated radioactivity (in caries) involved. (c) Disposition, including dates and destination if shipped off site. i i b. Environmental Monitoring (1) For each medium sampled; e.g., air, sediment, surface water, soil, or fish, include: (a) Number of sampling locations. Change No. 46 10-1,9 e-. . mm em- -e. -.~.m,~ .,e ---r-.s. u
6.9.3 (Cont'd) (b) Total number of samples. ~ (c) Number of locations at which levels are found to be significant1y above local backgrounds. (d) liighest, lowest, and the annual average concentrations or levels of radiation for the sampling point with the highest average and description of the location of that point with respect to the site. (2) If levels of radioactive materials in environmental media indicate the likelihood y' of public intakes in excess of 1% of those that could result from continuous annual exposure to the concentration values listed in Appendix B, Table II, Part 20, estimates of the likely resultant exposure to individuals and to population groups ? and assumptions upon which estimates are based shall be provided. I (3) If statistically significant variations of offsite environmental concentrations with time are observed, correlation of these results with effluent release shall g be provided. l-6.9.4 Special Reports Special Reports shall be submitted to the Director of the appropriate Regional office within j the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable technical specification section: a. In-service inspection reports. 6.10 RECORD RETENTION (Records not previously required to be retained shall be retained as required below commencing January 1,1976. ) 6.10.1 The following records shall be retained for at least five years; Records and logs of facility operation covering time interval at each power level. a. f 10-20 Change No. 46 p ( ...t, - ~ ~ ~h _~..-mmm..-.-
r ah.e w sweesh-
6.10.1 (Cent'd) b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety. c. Records of reportable occurrences. d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications. e. Records of reactor tests and experiments. f. Records of changes made to Operating Procedures. g. Records of radioactive shipments. ~ h. Records of training and qualification fur current members of the plant staff. 1. Records of seal'ed source leak tests and results. j. Records of annual physical inventory of all source material of record. 6.10.2 The following records shall be retained for the duration of the Facility Operating License: Record and drawing changes reflecting facility design modifications made to systems a. and equipment described in the FHSR. b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories. c. Records of monthly facility radiation and contmaination surveys. d. Records of radiation exposure for all individuals entering radiation control areas. e. Records of gaseous and liquid radioactive material released to the environs. f. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles. i I 10-21 Change No. 46 ~.
6.10.2 (Cent'd) g. Records of inservice inspections performed pursuant to the.se Technical Specifications. h. Records of Quality Assurance activities required by the QA Maaual to be retained for the duration of the facility operating license.
- i. Records of reviews perfor1ned for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50 59
- j. Records of meetings of the PRC and the SARB.
6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR, Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 RESPIRATORY PROTECTION PROGRAM ALIDWANCE 6.12.1 Pursuant ' o 10 CFR 20.103(c)(1) and (3), allowance may be made for the use of respiratory pro-t tective equipment in conjunction with activities authorized by the operating license for this facility in determining whether individuals in restricted areas are exposed to concentrations in excess of the limits specified in Appendix "B," Table I, Column 1, of 10 CFR 20 subject to the following conditions and limitations: The limits provided in Section 20.103(a) and (b) shall not be exceeded. a. b. If the radioactive material is of such form that intake through the skin' or other s, additional route is likely, individual exposures to radioactive material shall be controlled so that the radioactive content of any critical organ from all routes of intake averaged over seven (7) consecutive days does not exceed that which would result from inhaling such radioactive material for forty (10) hours at the pertinent concentra-4 tion values provided in Appendix "B," Table I, Column 1, of 10 CFR 20. For radioactive materials designated "Sub" in the " Isotope" column of Appendix "B," Table I, c. Column 1, of 10 CFR 20, the concentration value specified shall be based upon exposure to 10-22 Change No. 46 .=
6.12.1 (Cont'd) the meterial as an external radiation source. Individual exposures to these materials shall be accounted for as part of the limitation on individual dose in 5 20.101. Thesc materials shall be subject to applicable process and other engineering cont rols. PROTECTION PROGAM 6.12.2 In all operations in which adequate limitation of the inhalation of radionctive material by the use of process or other engineerin6 controls is impr.acticable, the licensee may permit an indi-vidual in a restricted area to use respiratory protective equipment to limit the inhalation of airborne radioactive material, provided: a. The limits specified in 6.12.1 above are not exceeded. b. Respiratory protective equipment is selected and used so that the peak concentrations of airborne radioactive material inhaled by an individual wearing the equipment do not exceed the pertinent concentration values specified in Appendix "B," Table I, Column 1, of 10 CFH 20. For the purposes of this subparagraph, the concentration of radioactive material that is in-haled when respirators are worn may be determined by dividing the ambient airborne concen- .tration by the protection factor specified in Table 6.12-1 for the respirator protective i equipment worn. If the intake of radioactivity is later determined by other measurements to have been different than that initially estimated, the later quantity shall be used in evaluating the exposures. c. The licensee advises each respirator user that he may leave the area at any time for relief from respirator use in case of equipment malfunction, physical or psychological discomfort, or any other condition that might cause reduction in the protection afforded the wearer. d. The licensee maintains a respiratory protective program adequate to assure that the require-ments above are met and incorporates practices for respiratory protection consistent with those s_ recommended by the American National Standards Institute (ANSI-288.2-1969). Such a program shall include: (1) Air sampling and other surveys sufficient to identify the hazard, to evaluate indiviilual exposures, and to permit proper seicetion of respiratory protective equipment. (2) Written procedures to assure proper selection, supervision and training of personnel using such protective equipment. (3) Written procedures to assure the adequate fitting of respirators; and the testing of respiratory protective equipment foi operability immediately prior to use. 10-23 Change No. 46
16.12,2 (Cont 'd) (h) Written procedures for maintenance to assure full effectiveness of respiratory protee-tive equipment, including issuance, cleaning, decontamination, inspection, repair and storage. (5) Writt'en operational and administrative procedures for praper use of respiratory protective equipment including provisions for planned linitations on working times as necessitated by operational conditions. (6) Bioassays and/or whole body counts of individuals (and other surveys, as appropriate) to evaluate individual exposures and to assess protection actually provided. h' The licensee shall u.e equipment approved by the US Bureau of Mines under its appropriate e. Approval Schedices as set forth in Table 6.12-1. Equipment not approved under US Bureau of Mines Apr. oval Schedules shall be used only if the licensee has evaluated the equipment and can demonstrate by testing, or on the basis of reliable test information, that the material and performance characteristics of the equipment are at least equal to those afforded by US Bureau of Mines approved equipment of the same type, as specified in Table 6.12-1. Unless otherwise authorized by the Commission, the licensee shall not assign protection f. factors in excess of those specified in Table 6.12-1 in selecting and using respiratory protective equipment. REVOCATION The specifications of 6.12 shall be revoked in their entirety upon adoption of the ' 6.12.3 proposed change to 10 CFR 20, Section 20.103, which would make such provisions unnecessary. s i 10-24 Change No. 46
TABLE 6.12-1 Protection Factors for Respirators Protection Factors 2 s Particulates and Guides to Selection of Equipment Vapors and Gases Bureau of Mines /NIOSil Approval
- Schedules Except Tritium-For Equipment Capable of Providing at Description Modesl oxide 3 Least Equivalent Protection Factors I.
AIR-PURIFYING RESPIRA'IORS Facepiece, Half-Mask"'7 NP 5 21B 30 CFR S lb.h(b)(h) 7 Facepiece, Ful1 NP 100 21B 30 CFR 5 lb.h(b)(5); 1hF 30 CFR 13 II. ATMOSPHERE-SUPPLYING RESPIRATOR 1. Air Line Respirator Facepiece, Half-Mask CF 100 19B 30 CFR 5 12.2(c)(2) Type C(i) Facepiece, Full CF 1,000 19B 30 CFR 5 12.2(c)(2) Type C(i) 7 Facepiece, Full D 100 19B 30 CFR S 12.2(c)(2) Type C(ii) Facepiece, Full PD 1,000 19B 30 CFR 5 12.2(c)(2) Type C(iii) Hood CF 5 6 Suit CF 5 6 2. Self-Contained Breathing Apparatus (SCBAl ,Fe -: piece, Full 7 D 100 13E 30 CFR S 11.h(b)(2)(i) Facepiece, Full PD 1,000 13E 30 CFR S 11.h(b)(2)(ii) Facepiece, Full R 100 13E 30 CFR S 11.h(b)(1) III. COMBINATION RESPIRATOR Any Combination of Air-Protection Factor for 19B CFR S 12.2(e) or Applicable Purifying and Atmosphere-Type and Mode of Opera-Schedules as Listed Above Supplying Respirator tion as Listed Above 1, 2, 3, 4, 5, 6, 7(These notes are on the following pages.)
- 0r Schedule Superseding for Equipment of Type Listed 10-25 Change No. 46
( ( ( TABLE 6.12-1 (Contd) 1 See the following symbols: CF: Continuous Flow D: Demand NP: Negative Pressure (ie, Negative Phase During Inhalation) PD: Pressure Demand (ie, Always Positive Pressure) R: Recirculating (Close.' "ircuit) 2(a) For purposes of this specification the protection factor is a measure of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radioactive material outside the respiratory protective equipment to that inside the equipment (usually inside the face-piece) under conditions of use. It is applied to the ambient airborne concentration to estimate the concentration inhaled by the wearer according to the following fomula: Concentration Inhaled = Protection Factor (b) The protection factors apply: (i) Only for trained individuals wearing properly fitted respirators used and maintained under supervision in a well-planned respiratory protective program. (ii) For air-purifying respirators only when high efficiency (above 99 9% removal efficiency by US Bureau of Mines type dioctyl phthalate (DOP) test) particulate filters and/or sorbents appropriate to the hazard are used in atmospheres not deficient in oxygen. (iii) For atmosphere-supplying respirators only when supplied with adequate respirable air. 10-26 Change No. 46 ~
a ( ( TABLE 6.12-1 (Contd) 3Excluding radioactive contaminants that present an absorption or submersion hazard. For tritium oxide, .approximately half of the intake occurs by absorption through the skin so that an overall protection factor of not more than approximately 2 is appropriate when atmosphere-supplying respirators are used to protect against tritium oxide. Air-purifying respirators are not recommended for use against tritium oxide. See also Footnote 5, belov, concerning supplied-air suits and hoods. "Under chin type only. Not recommended for unie where it might be possible for the ambient airborne concen-tration to reach instantaneous values greater than 50 times the pertinent values in Appendix "B." Table I, Column 1 of 10 CFR, Part 20. 5Appropriate protection factors must be determined taking account of the design of the suit or hood and ito l permeability to the contaminant under conditiens of use. No protection factor greater than 1,000 shall be used except as authorized by the Commission. 6No approval schedules currently available for this equipment. Equipment must be evaluated by testing or on basis of available test inforination. ( 70nly for shaven faces. I l Protection factors for respirators, as may be approved by the US Bureau of Mines and/or NIOSit according 1 NOTE 1: to approval schedules for respirators to protect against airborne radionuclides, may be used g to the extent that they do not exceed the prctection factors listed in this table. The pro-f' tection factors in this table may not be appropriate to circumstances where chemical or other {, respiratory hazarda exist in addition to radioactfye hazards. The selection and use of } respirators for such circumstances should take into account approvals of the US Bureau of Mines and/or NIOSil in accordance with its applicable schedules. j, i' NOTE 2: Radioactive contaminants for which the concentration values in Appendix B, Table I of this part are based on internal dose due to inhalation may, in addition, present external exposure hazards j. i i at higher concentrations. Under such circumstances, limitations on occupancy may have to be II governed by external dose limits. f 1 10-27 Change No. 46 t .}}