ML20024G123
| ML20024G123 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/22/1976 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20024G111 | List: |
| References | |
| NUDOCS 9102070555 | |
| Download: ML20024G123 (33) | |
Text
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i NORWEPR STATT;S_POWPR_ COMPANY i
1
_ DOCKET No. 50_-263 M. ON.TICELLO.NUC.L. EAR GENERATIN. G STATION AMENRfNT *Q PROVISI,0NAL_0PERATING LICENSE l
j
, Amendment No. 15-
' License No. DPR-22 1.
W t.uclerr i:crulatory Cor:nission (the Cormission) has found thnt.
'ino application for amendment by the Northern States Power Company (the licensee) dated *,ecember 16, 1974, couplies tvith tho' standards.
and requirements of tne Atomic Energy Act of'1954, as amended (the Act). and the Coumission's rules and regulations set forth in 10 Crn Cin;pter I -
'f ue facility vill operate in conformity with the application, the n.
- irovnions 01 tac Act, and the rules and regulations of the Comnission.
L.
mere is reasonable assurance (1) that the cetivities authorized by tais anendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; g3, s
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n.
'The issuance of this a:aendment will not be inimical to the~ common defense and security or to the health and safety of the public: and E.
An environmental statement or negative decle. ration.need not be prepared in connection vith the issuance of this amendment.
Accordingly the license is amended by a change to the Technical SpcW41 cations as indicated in the attachment to this license at endnent and Tararraph 5.B of racility License No. DPR-22 is ht:reby Jec to rerJ - follow.
9102070555 760122 PDR ADOCK 05000263 P
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'B.
Technical Spec _ifications The '4echnical Specifications contained in Appendix A, as revised, are hereby incorporated in the license.
Tne licensee shall operate the facility in accordance with the Technical Specifications, as revised.
3.
This license amendment is effective 30 days after its date of issuance.
FOR TIE NUCLEAR REGULATORY C0!911SS10N
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a.
a Dennis L. Zierann, Chief Operating Reactors Ilranch 82 Division of Operating Peactors A t t n W ~1 -
Ol'in?ns to the Tec'm1 c,1 Srnci fice t i nnt ITt0 O f 3 f S 8.1; t'C C j'f
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j ATTACiB!ENT TO LICENSE AMENDMENT NO. 15 PROVISIONAL OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the existing pages of the Technical Specifications listed below with the attached revised pages bearing the same numbers, except as otherwise noted.
Changed areas on these pages are shown by marginal lines.
Pages lii, y,
vi and vii Table of Contents Pages viii and ix - AdCtion (as result of expansion of T.L'e of Contents)
]
1 1A - Deleted l
81 134 136 150 164 170A 189 P.
195 196 197 200 201 210 211 212 213
- 214 215 216 216A - Deleted 217 21S 219 - Deleted 220 - Deleted 2
1 Ir l
f O.
IIPCI Systew.
103 4
li.
Automatic Pressure Itelief System 101 1
F.
RCIC System-106 G.
Minimum Core and Containment. Cooling System Availability 107 11.
Deleted 1
l.
Recirculation System 108A I
3.5-Bases 109 i
4.5 11ases 114 3.6 and 4.6 Primary System 11oundary 115 A.
'Ihermal Limi tations 115 i
l B.
Pressurization Temperature 116 l
I t
C.
Coolant Chemistry lib D.
Coolant I.eakage 118 E.
Safety / Relief Valves 119 F.
Structural Integrity 1.'O I
i G.
Jet Pumps 120 i
i i
11.
Ilydraulic Snubbers 121 l
j.
3.e,and 4.6 nases 130 i
i t
4 Amendment No. 15 iii
,.f.
4 i
(,
3.
Standby Diesel Generators 182 4.
Station Battery Sys tems 183 3.9 Bases 185
)
i 4.9 Bases 186 i
i 3.10 and 4.10 Refueling 187 A.
Refueling Interlocks 187 B.
Core Manitoring 188 4
C.
Fuel Storage Pool Water level 188 D.
Movement of Fuel 188 I
1 E.
Extended Core and Control Fud Drive Itnintenance 188A~
l i
j 3.10 and 4.10 Eases 189 i
3.1I and 4.I1 Ileactrr Fuel AssenhIles 189B l
i j
A.
Average Plantir Linersr lleat tienc rat ion Itate lh9B f
i B.
Local LIIGil IB9C C.
Flinimum Critical Power 1(at io Ib90 4
i I
3.I1 liases 18911 4.11 Bases 169G 3.1.' and 4.12 Scaled Source Contamination 169N f
A.
Contamination Ib9N l
L.
Itecords th9P 3.12 and 4.12 11ases 189Q t
1 j
L Amendnient No. 15 v
a
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1' 5.0 DESIGN FEATURES 190 i
6.0 AD.N.INISTRATIVE CONTROLS 192 6.1 Organization 192 6.2 Review and Audir.
195 i
6.3 Deleted 6.4 Action to be taken if a Safety Limit. is Exceeded 201 6.5 Plapt Operating Procedures-202 6.6 Plant Operating Mecords 209 6.7 Reporting Requirements 211 l
i Amendment No. 15 vi
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1.IST OF:FIGUPFS 4
i Figure No.
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.Page No.
3.II.1-C Maximum Average Linear lleat Generat ion Itate versus Planar Average I!xposure Monticel10 8D262 Fuel 189J
-3.ll.1-D Maximwn Average Linear lleat Generation Itate versus Planar Average' Exposure Monticello 8D250 Fuel 189K i
1 3.ll.1-E Maximum Average Linear !! eat Generation Itate vetsus Planar Average Exposure t
Monticelle 8D219 Fuel 189L t
3.I1.2 K
Factor versus Percent of 1:ated Core Flow g
189M.,
I' 6.1.' 1 NSP Corporate Organizational itelationship to 'On-Site Operating Organization 193 L
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- unendiaua t No. 15 viii
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1.lST (W TAhl.LS 3.1.1 1:eactor Protection System (Scram) instrument Requirements 30 4.1.1 Scram lustrument Functional Tests - flinimum Functional Test Frequencies for Safety Instrumentation and Cont rol Ci rcuits 34 4.1.2 Scram Instrument Calibration - Flinimum Calibration Frequencies for Reactor Protection Instrument Channels 36 3.2.1 Instrumentation th at Initiates Primary Containment Iselation 1: unctions 50 3.2.2 Instrumentation that Initiates timergency Core Cooling Systems 53 3.2.3 Instrumentation that Initiates 1:od 111ock S7 3.2.4 Instrumentation that initiates Reactor liuilding Ventilation isolation and Standby Gas Treatment System Initiation 60' i
I b
3..'. S Trip Functions and Deviations 69 l
4.2.1 blinimum Test and Calibration Frequency For Core Cooling, Rod lilock and Isolation lustrumentation 61 3.0.1 Safety helated ilydraulic Jnubbers 12115 1.0.1 In-Service inspection Requirements f or Nuticello 124
- 3. /.1 Primary Containment isolation 153 4.h.]
Sample Collection and Analysis f.lont icello finclear Plant - linvi ronmental Flonitoring Program 174 6.5.1 Protection Factors for Respirators 206 i,
f t
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Amendment rio. 15 ix
IrrritoDIR: TION i
These Technical Specifications are prepared in accordance with the requirements of 10 CI'R 50.36 and apply to the Monticello Nuclear Generating Plant, Unit No. 1.
The bases for these Specifications are included for information and understandability purposes.
1.0 DEFINITIONS The succeeding frequently used terns are explicitly defined so that a uniform interpretation of the Specifications may be achieved.
A.
Deleted' t
B.
Alteration 'of the Reactor Core
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l The act of. moving any component in the region above the core support plate, below the upper i
grid and within the shroud.
(Normal operating functions such as control rod movement using 1
the normal drive mechanism, tip scans. Slot and II H detector movements, etc., are not to be considered core alterations.)
C.
Ilot Standby 5
i llot Standby means operation with the reactor critical in the startup mode at a power level just sufficient to maintain reactor pressure and tuaperature.
r i
i Amendment No. 15 1
. _ ~
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30 LIMITING CONDITIOIIS FOR OPEHATION '
'.0 ntutVEILIAUCE 11EQUIR104EITI"1 I
i E.
Reactivity Anomalies E.
Ipactivity Anomalies e
At, a specific steady state base condition During the startup test program and at
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of the reactor act.ual control ro<l inventory each startup following refueling outages, t
will be periodically compared to a norm l-the acttn1 rtxl inventory shall be com-ized computed prediction or' the inventery.
pan d to a nontalized computed prediction If the difference exceeds one per cent, d k, o f the i nven t ory.
These comparisons will j
reactor power operation shall not be per-te used as base data. for niacti vity mon-mitted until the cause has been evaluated it oring during subsequent power op3 ration l
and appropriate corrective action has been throughout the fuel cycle. At speci fic completed.
p.wer opn ating corditions, the actual
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rul confi unation will be compared to
(
d the configuration expected based upon appropriately corzveted past data. This l
l F.
If Specifications 3.3. A thr.>t ch D above comparison will be made at least every are not met, an orderly shutdown shall equivalent full power month.
be initiated and have reactor in the cobt l
shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
f r
i t
l 81 E
3.3/4.3 Amendment No. 15 l
Bases Continued 3.6 and 4.6:
D.
Coolant leakage lhe fomer 15 spm limit for leaks fr.na unidentified sources was established assur:iing such leakage was coming f rom the primary system. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the probability that the crack will propagate.
From the crack size a leakage rate can be determined.
for a crack size which gives a leakage of 5 gpm, the probability of rapid propagation is less char.10-5 Thus, an unidentified leak of 5 gpm when assumed to be from the prizaary system had less than one chance in 100,000 of propa-gating, which provides adequate margin. A leakage of 5 gpm is detectable and measurable. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allowed for determination of leekage is also based on the low probability' of the crack propagattug.
The capacity of the dryvell sump pumps is 100 gpm and the capacity of the dryvell equipment drain tank pumps is also 100 gpm.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
E.
Saf ety/ relief Valves Testing of all safety / relief valves each refueling outage ensures that any valve deterioration is detected.
A tolerance value of 17. for safety / relief valve setpoints is specified in Section III of the ASME Ik>iler and Pressure Vessel Code.
Analyses have been performed with all valves assumed set 17. higher (1080 psig + 17.) than the nor nal setpoint; the 1375 psig code limit is not exceeded in any case.
The safety / relief valves are used to limit reactor vessel overpressure and fuel tl.armal duty.
The required safety / relief valve steam flow capacity is determined by analy:Ing the tr.1nsient the mainsteam flow stoppage resulting from a postulated MSIV Closure from a power of 1670 accompanying rM. The analysis assumes a analtiple-failure wherein direct t
scram (valve position',
neglected. Scram is assumed to be from indirect meens (high flux)..
In this event, the safety / relief valve c00acity is assumed to be 717. of the full power steam generation rate.
3.6/4.6 IIASES 134 Asaendment No. 15
+
ibscu Continued 3.6 and h.6:
i besi n confirmation and construction udequoey uill be demonstrated durisq: the plant startup und 6
power ascension t.est procrum. As part of this progrsui, cold tmd hot, vibration tests on certain reuctor vessel internals vill be perfo2neJ. 7he tests, descrited in a letter to Dr. P. A. Itorris, dated March 5,1970, are designed to obtain confizuatory dat., on the deuten reatures or11onticello as canpored to Dresden Unit 2 desi n.
Thus, the basis for the Monticello vibration test procra.a is E
- prediented on obtaining satisfactory data which conrima etvidon desica reatures from earlier IMH plants such as Dresden Unit 2.
In the event that data from these earlier plant.s are not available before
)
{ routine power cperation of Monticello, the matter will be reviewed by the NRC.
She progrnm outlined in Table h.6.1 is limited to irupections or the primary coolant system. It is i
I unticipated that, the data collected durirq: the first five years or operation will provide n suitable buuis to evaluate the need for inspecting other portions or the racility (such as the main steam lines 3
downstream of the main steamline isolution valves). These dntu along with the overall operating i
experiences will be reviewed to datemine the inspection program to be implement.ed for the lifetime of the facilit,y. The results of thia study t.ocether with the proposed liret.ine inspection program vill be submitted to the NRC in accordance with Specification 6.7.c.2.
'1he special inspection of the main feed and utca:a linea in to provide added protection against pipe whip. 4he Group I welds are selected on the Louis or un analysia that. shows these velds are the
,~
h1 heut stress velds and that due to their physical location, o break would result in t he least, 6
f interference and maximtun energy upon impact with the dryucil. These velds are the only ones which offer any significant risk and vill be included in future impections au detemined by the study described i
above.
Group II welds are selecte<1 because eithout regard for the operut.iry; streau levels and int er-fering equi uent, they have sufficient, theoretical enercy t.o puetrate un1 would propel the pipe l
toward the contnitunent. 'lhey nre therefore included in tim rir at inspection. Upon consideration of impact uncle, interfering equi enent and distance pipe travelu, m aubutant,ial risk is involved and l
no extra inspection is needed.
In addition, exiensive visual inspection for let, ulil he rwk perlottically ou erttical systen.s.
'1he inwpection procrum specified encompasaca the major creau or th.
vessel on1 piping systems within the dryvell. The inspection porlod is based on the observed rate or crowth of defects fran fatigue I
l studies sponsored by the NRC. These studies show that it requires thoua:nals or stress cycles at 3.6/4.6 136 Y
Amendment No. 15
E l
l 3.0 LlHITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE HEQUIREMENTS C.
- 1.. Secondary containment integrity, shall be 1.
Secondary containment surveillance shall i
maintained during all modes of plant be performed as indicated below:
operation except when all of the following conditions are met, l
The reactor is subcritical and Specifi-a.
Secondary containment capability to l
a.
cation 3.3.A is met.
maintain at least a 1/4 inch of water I
2 vacuum under cal:n vind (2 (u < 5 mph) conditions with a filter train flow a
rate of64,000 scfm. shall be dem-onstrated at each refueling outage prior to refueling.
b.
The reactor water temperature is below 0
212 and the reactor coolant system is vented.
i t
c.
No activity is being performed which f
can reduce the shutdown margin below that specified in Specification 3.3.A.
t 150 1.7/4.7 l
4 Amentbi cnt No. 15
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4 3.0 LIMITING CONDITIONS OF OPERATION 4.0 SURVFll I.AMCE REQUIREMENTS i
t a.
Investigate te identify the 5.
A determination shall be made of the causes for.such release rates.
i total I-131 released weekly. An analysis j
b.
Define and initiate a program shall be performed on a sample at least monthly ' for I-133 and 1-135.
i to reduce.such release rates to.the as low as practical levels.
6.
A deteretnction shall' be made of the I
total radioactive particulates with half-Provide a report describing c.
lives gre atet-than 8 days released weekly.
[
these. actions within 30 days.
The particulate filters shall be removed l
and analyzed for gross beta particulate radioactivity with half-lives greater than 8 days. Monthly, a composite of l
j-10.
At least one of the two stack munitors, i
including the charcoal cartridge and those filters used during the month particulate filter, shall be operable shall be prepared and analyzed for the l
principal gamma emitting radionuclides.
at all times that the stack is releasing effluents to the environs.
7.
. analysis for Sr-89 and Sr-90 shall be 11.
If both stack monitors are made or found in-made quarterly. Cross alpha radioactivity shall be determined quarterly.
operable, the. reacter shall be placed in the hot standby condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
12.
Except as specified in 3.8.A.13, the off-l I
gas stack and reactor building vent moniturs shall have automatic isolation set points j
consistent with Specification 3.8.A.1 and alarm set points consistent with Specifi-ca t ion 3.8. A.2.
13.
If operation is necessary with the Of f gas lloldup System recombiners bypassed, the off gas stack monitors sha11' serve only an alaru function.
i i
3.8/4.6 170A I
Amendment No. Ib J
3.0 LIMITING CONDITittJS FOR OPERATIONS 4.0 SultVI:ll.l ANCE REQtlIRI.MHITS
~
4 3.11 REACTOR FUEL ASSDiBLIES 4.LL Ri%CTOR FUEL Assail! LIES Applicability Applicability The Limiting Conditions for Operation 1he Sutveillance Requirements apply to associated with the fuel rods apply to the parameters which monitor the fuel those parameters which monitor the fuel ros! operating conditions.
rod operating conditions.
Objective Oldec t E ve The objective of the Limiting Condit ions the object we of the Surveillance Requiremeats' for Operation is to assure the perfor-is to spe.:ify the type and f requency of surveil-mance of the fuel rods.
lance to be applied to the fuel rods.
Specificatio s Spee l f len t ions A.
Average Planar Linear lleat Gener:n-A.
Ave rage Planar I.inear IIcat Genera-tion Rate (API.IICR) ion Ita te ( API.llGR)
During steady state power opesation, I. The APLIICR for each type of fuel as a the API.llGK for eact. type of f uel a3 f unct ion of average planar exposure shall function of average pla~nar exposure be detennined daily during reactor operation a
shall nat exceed the Ifuiting value at 2 257. rated thermal power.
shown in Figures 3.11-1.
If at any time i t.
is determined that the Iimit-
- 2. Whenever the plant technical staff determines ing value for APLilGR is being exceed-that more frequent survettlance of APL!!GR cd, action shall be taken inaued ia t e ly is n ec e s sa ry, it shall specify an augmented to restore opetation to within the surveillance program cousuensurate with prescribed limits.
r eactor conilitions.
3.11/4.11 gp Amendment No. I r,
6.2 neview and Audit Organizational units for the review nnd audit of f uellity operationc stmll be constituted and have the responsibilities and aut.horities outlined below:
A.
Safety Audit Committee (SAC)
The SAC must: verify that openttion of the pinnt. is conniutent wit.h company policy and rules, approved operating procedures anl ci ennting lleense prm inionc; review import. ant proposed plant changes, tests und procedures; veril'y th:.t unusunt events are prumptly investigated and corrected in a wnu:er which reduccu the psobabliity of n. urrence of such events; and detect trends which naay not be apparent to a day-to-dny observer.
Audits of selected aspects of pinut operation rhall be 1.ci ronr.ed with a fmetuency connensurate with their safety significance emd in such a it muer as In annute that an audit of e.11 nuclear safety related activities is conplet ed within n period of two years.
Periodic review of the audit prognuns should be perfons d I y the CAC nt. least Iwice n year to at cure that such nudits are being accompliol ed in accor d me-with reilutrements of Techniani Ugecifications.
'lhe au-di ts shall be ucrformed in acec.bl ume with uppii,priat e u s i t ten inctruct.lons or procedures and shall include verification of c..cyllance with internal rules, procedures (for ex.auple: no nna 1,
'of f-nonnul, emer6ency, operatini, milutenance, t,u rveill a r. e, test and radiation cont.roi pru-cedures and t.he emetgency and cecui ity pinus), recuinti -
involving nuclear safety and operu-ting license provisions; traintnc,.potli f t erition and pei t e i a.ance of ope rutinir, staff; and cor-rective actions following reportable occurrences or unusual events. A representat ive portion of prucedures and reconis of the activities perronaed during the nuuit period t. hall be audited and,in tuldition, obcervations or pe rformance or operating and saintenance activities shall be l
i included. Written reports of such nudits ohn11 be mvicued at t scheduled meeting of the SAC l
and by apprupriate enembers of runnnfrment, including thone having responsibility in t.ne uren audited. Follow-up action, inclu< ling renudit of deficient arens, shall be taken when indicated.
1.
&mbership
'1he SAC shall consist of ut leuc t, five ( S) percor.6.
n.
l D
Amendment No. 15
b.
The SAC Chaltuan shall be a frJp ennoncreent reprecentative nppointed by the Crcup Vice i
President-pouer Supply.
Other SAC c.c 21ers ohn11 Le oppointed by the Gzoup Viec Preci-dent-Pover Supply or such othar percon no be er.ny denic,nate.
'lle Chairman chall appoint a Vice Chnisman from the SAC membersh.p to net in his abcence.
No more thou tuo meanbers of the SAC shall be from grotips holding line responcibility c.
for eperation Of 1.he plant.
d.
The SAC members should collectively have the capability required to review problems in the following areas: nuclear powei pl.uit operations, nuclear engineering, chem-istry and radiochemistry, metallurgy, ins t rurnent a t ion and enut r ol, radiological 4
safety, mechanical and electrical engineeninn, and other appropriate fields asso-cinted with the unique characteristles of t he nucle.ir power plant involved. A minimum of fcur voting members shall have a minir.tum of a II.whelors Degree in Engineering or a scientific discipline and possess a minhaum of three years of professional level experience in nuclear services, nuclear plant operat ion, or nuclear engineering. When the nature of a particular problem dictates, special consultants will be utilized, as necessary, to provide expert advice to ahe SAC.
2.
Meeting Frequency The SAC shall meet on call by the Chairman but not less ircquently than twice a year.
3.
Quorum A quorum shall include a majority of the permanent ruember s, including the Chairman or Vice Chai rtuan.
No more than a minority ut the quus ma shall be from groups holding line responsibility for the operation of the plant.
4.
Responsibilities - The tollowing subjecto should be reported to and reviewed by the SAC:
Proposed tests and experiments, and their results, when such tests or experiments a.
may constitute an unreviewed safety question as defin.d in Sect ion 50.59( a)(c), pa rt 50, g
Title 10, Code of Federal Regulations.
b.
Draluations of proposed changes to procedures, equipment and systems cocpleted under pro-visions of Section 50.59(a)(1), part 50, Title 10, Code of Federal itegulatwns to ver2 y l
that such proposed changes do not constitute unreviewed safety questions.
6.2 1%
Amendment No. 15
Proposed changes in proceduies, cyniguaent or systems uliich way involve a unreviewed c.
safety question as defined in Sect ion 50.59(a)(2), Part 50, Title 10, Code of Federal Regulations or changes referred to it by the operating organization.
d.
Proposed changes in Technical Specifications or operating license.
e.
Violationsofapplicablestatuteu, codes, reSulations, orders, Technical Specifications, operating. license requiremients, or of internal procedures or instructions having safety significance, f.
Significant operating abnormalities or deviations f rom normal and expected performance of plant equipment.
All events which are required by s'egulations or Technical Specifications to be reported l
g.
to NRC in writing within 24 heurs, li.
Any indication of.an unanticipated deficicacy in sona: aspect of design or operation of salaty related structures, systems or components.
1.
Operations Conanittee proceedings and ininutes to determine if matterrs considered by that Committee involve unreviewed or unresolved safety questions.
l j
- j. Training, qualification and performance of operating staff.
k.
Disagreement $between tite seconnendations of the Opetutions Connittee and the Plant llanage r.
1.
Security and einergency plans and their liuplementing procedures.
Environraental }ionitoring Program and its re sie l t s.
m.
n.
Quality Assurance prograus and evaluate its adequacy.
Review of events covered under 4.e-4.h above shall include reporting to appropriate n. embers of management on the results of inveuttgations and recomendations to prevent or reduce the probability of recurrence.
5.
Authority The SAC shall bd ad,'isory to the Group Vice President-Power Supply.
Auteudmen t No, IS 197
f.
All events which are reipiired by regulations or Technical Specifications to be repot ted to NitC in writiiig within 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />.
g.
Drills on enw rgency proceduren (including plauit, evacuationi) und adequacy of cosamuni-Cation with off-site support gauups.
h.
All procedures required by these 'Ibchnien1 Spect rications, including leiplementing procedures of the Dnergency Plan and the Security Plitu shall be reviewed with a frequency conenensurate with their safet.y signific.ince but, at un interval of not,
.inore than two years.
1.
Perform special reviews and investigationo, as requested by the Safety Audit Obm-rnittee.
5.
Authority
'ihe OC shall be nelvisory to the Plant Ibnuger.
In the event of disngreeixnt betweert the arconsnendations of the OC and the PlustL thnnger, Ihe course dete:Tuined by the j
Plant thnager to be the suore conuervntive will be followed. A written sununary of the
. disagreement will be sent to the Generni Sog erinten+:ot of fluelens' Power Plant. Opern-tion and the Chainrnn of the SAC for review.
6.
Deconis liinutes shall be recorded for ali nicetings of the OC and shall identify all documentury materlat reviewed. The minutes shall be dist ributed to each saember of the OC, the Chairman and each member of the Safety Audit Couuni t tee, t he Ceneral Superintendent of 11uclear Power Plant Operation and othero desipated by OC Chaliman or Vice Chairman.
7, Procedures A written clutrter for the OC slutil Le prelatient that contalus:
Itesponsibilit.y and authority of the gavup a.
b.
Oontent and method of submissior of prvuentations to the Operutions Conanittee i
i Ame,?idment fjo. 15 200
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Principal maintenance activities, including inspection, repairs and substitution or replacement of principal items of equipment pertaining to nuclear safety.
6.
Records of changes-to plant procedures and records of special tests and experiments.
'7.
Records of wind speed and direction.
8 Records of individual plant staff members showing qualifications, training and retraining.
I i
9.
Reportable Occurrences.
I B.
Records 8tetained for Plant Life j'
Records and logs relative to the following items shall be retained for the life of the plant:
j 1.
Liquid and gaseous radioactive releases to the envi rons
{
2.
Radiation exposures for all plant, visitor and contractor personnel 3.
Off-site environmental monitoring surveys 4.
Fuel accountability including new and spent fuel inventories and transfers, and fuel assembly histories 5.
Radioactive shipments
.t 6.
Plant radiation and contamination surveys e
7 Changes made to the plant as it is described in the Final Safety Analysis Report, l
reflected in updated, corrected and as-built drawings 1
8 Cycling beyond normal limits or those components that have been designed to operate safely for a limited number of cycles beyond such limits 9.
Reactor coolant syste:n in -ser vice inspections L
i 10.
Flinutes of meetzy of the Safety Audi t Commit tee t
i Amendment No. 15 210
l 6.7 REPORTING RE0llII DIENTS In addition to ti.e applicable reporting requirti: ent s of Title 10, Code of Federal hegulations, the following identified reports shall be submitted so the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted.
A.
Routine Reports 1.
Startup Report j
A sucuaary report of plant startup and power escalation testing shall Le submitted following l
(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) modificaticas that have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
- na report shall address each of a
the tests identified in the FS 2 and shall in general include a descr3ption of the measured values of the operating conditions or characteristics obtained during the test program and a _omparison of these values uith 4tesign pred it.t ions and specificat ions.
Any corrective actions that vere reciuired to ott ain satisfactory operation sh..ll also be described.
Any additional specific detaf f s required in license conditions based on other coenituents shall be included in this report.
Startup reports shall be s"bmit te-I wit hin (1) W days followinn completion of the startup test program, (2) 90 days follouing resumption or cor.aencement of c. umerchi power operation, or (3) 9 months following initial criticality, saichever is earliest.
If the Startup Report does not cover all three events (i.e.,
initial criticality, completion of startup tert program, and renumption er comaiencemer t of cornerc ial pouer operation), supplementary reports shall be submit ted at least every three moaths antil all three events have been completed.
Annual Operating Repart /
1 2.
Routine operating report s < svm iw the operation of the unit during the previous calendar year should be submit ted prior t. P irc h 1 o r each year.
The init ial repot t shall bs. submitted prior to March 1 of the year f ollowin;; initial c rit ical ity.
[I A single submit tal may be nade for a multiple uait station.
The submittal should combine those nections t h.+ t are coctmon to all units at the station.
Juacndment tu. I5 211
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in the technical cpecificat.:an; sho r t - t e rr. t ract ivit y i nc rea:,es that currespond to a rea tor period of less than S secena ur, 11.;uu-et it ical. an unplanin d reactiv hy insertion el more than 0.5% AL/k or occutrence of any unplanned criticality.
e.
Failure or saalfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional rujuirements of system (s) used to cope with accidents analyzed in the SAR.
f.
Personnel error or procedural inadequacy,which prevents or could pre sent., by itself, the fulfillment of the f unctional requireu.ents 01 systems required to cope with accidents l
analyzed in the SAR.
Note:
For it ems 6. 7. B. I.e. and 6. 7. B. I. f. reduced redundancy that does not result in a loss of system function need not be reported under this section but may be reportable under items 6.7.B.2.b. and 6.7.B.2.c. below.
g.
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 speci fications.
or in the methoda used for sdeh h.
Error: discovered in the t r, ins i en t or accident analy-ca analyses as described in the safety analysiu report or in the bases for the technical specifications that have or could have permit ted reactor operation in a manaer less conservative than assuued in the analyses.
1.
Periormance of structures, o y s t e.r.s, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assemed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not apecitically considered in the :.af ety analysis report or technical specifications,that require remediai action or corrective measures to prevent the exiatence or development of an unsafe condition.
l Note: This item is intended to provide for reporting ut potentially generic problenu.
Amendment No. 15 215 l
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