ML20095E054
| ML20095E054 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/08/1995 |
| From: | Stone J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20095E057 | List: |
| References | |
| NUDOCS 9512140162 | |
| Download: ML20095E054 (15) | |
Text
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION -
.p WASHINGTON, D.C. 30006 4001 i
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WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION c
DOCKET NO. 50-482 i
i ANENDNENT TO FACILITY OPERATING LICENSE f '
Amendment No. 92' License No. NPF-42 t
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The Nuclear Regulatory Commission (the Commission) has' found that:
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.A.
The application for amendment to the Wolf Creek Generating Station (the facility) Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated
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June 14, 1995, as supplemented by letters dated July 13, 1995, and August 22, 1995, complies with the standards'and requirements of the L
Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of l
the Commission; 4
C.
There is reasonable assurance:
(1) that the activities authorized F
by this amendment can be conducted without endangering the health and safoty of the public, and (ii) that such activities will be i
conducted in compliance with the Commission's regulations; I
D.
The issuance of this license amendment will not be inimical to the F
common defense and security cr to the health and safety of the i-
'public; and E.
The issuance of this amenkont is in accordance with 10 CFR Part 51 4
of the Commission's regulations and all applicable requirements have
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bee.n satisfied.
i 9512140162 951208 PDR ADOCK 05000482 P
PDR-e
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Accordingly, the' license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-42 is hereby amended to read as follows:
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Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 92, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are i
hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented prior to restart from the eighth refueling outage, which l
i is scheduled to begin in March 1996.
1 FOR THE NUCLEAR REGULATORY COMMISSION i
d I
ames C. Stone, Senior Project Manager Project Directorate IV-2 I
Division of Reactor Projects III/IV j
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical i
Specifications Date of Issuance:
December 8, 1995
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ATTACHMENT TO LICENSE AMENDMENT NO. 92 FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain marginal-lines indicating the areas of change. The corresponding l'
overleaf pages are also provided to maintain document completeness.
REMOVE INSERT 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-10 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 6-21 6-21 6-21a 6-21a l
6-21b 6-21b
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i THIS PAGE HAS BEEN DELETED 4
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WOLF CREEK - UNIT 1 3/42 Amendment No. J,61
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POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z) l t
LIMITING CONDITION FOR OPERATION i
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3.2.2 F,(Z) shall be limited by the following relationships:
F,(Z) s [F/") [K(Z)) for P > 0.5, and T
l F,(Z) s [F/"] [K(Z)] for P s 0.5.
i T
Where:
F/" - the F. (I) Limit at RATED THERMAL POWER (RTP),
i as specified in the CORE OPERATING LIMITS REPORT (COLR),
l P
THERMAL POWER
, and
=
i RATED THERMAL POWER K(Z) - the normalized F,(Z) limit as a function' of l
core height, as specified in the COLR.
APPLICABILITY: NODE 1.
ACTION:
With F,(Z) exceeding its limit:
a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit within 15 minutes and similarly reduce the Pow,er Range Neutron Flux-High Trip Setpoints within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; POWER OPERATION may 4
proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F,(Z) exceeds the limit; and b.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F,(Z) l is demonstrated through incore mapping to be within its limit.
WOLF CREEK - UNIT 1 3/4 2-4 Amendment No. fi,,s?
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1 POWER DISTRIBUTI0d LIMITS l
SURVEILLANCE RrPUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
i 4.2.2.2 F,(Z) shall be evaluated to determine if F,(Z) is within its limit l
by:
a.
Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER; l
b.
Increasing the measured F (Z) component of the power distribution map by 3% to account for, manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.-
Verify that the requirements of Specification 3.2.2 are satisfied.
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i c.
Satisfying the following relationship:
[F '"][K(Z)]
F,"(Z) s --,---------- for P > 0. 5
[P][W(Z))
[F ""][K(Z)]
1 l
F,"(Z) s --,---------- for P s 0. 5 j
[0.5][W(Z)]
where F " Z) is the measured F (Z) increased by the allowances for manufaclu(ring tolerances and m,easurement uncertainty and W(Z) is th j
j cycle dependent function that accounts for power distribution transients encountered during normal operation. This function is
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provided in the COLR.
d.
Measuring F,"(Z) according to the following schedule:
3.
Upon achieving equilibrium conditions after exceeding, by 10%
L or more of RATED THERMAL POWER, the THERMAL POWER at which F,(Z) was last determined *, or j
c 2.
At least once per 31 Effective Full Power Days, whichever cccurs first.
_e.
With measurements indicating
' F,"(Z) maximes over z K(Z) has increased since the previous determination of F,"(Z), either of the following actions shall be taken:
1 1.
F,"(Z) shall be increased over that specified in 4.2.2.2.c by an appropriate factor specified in the COLR, or
- During power escalation at the beginning of each cycle, THERMAL POWER may be increased until a power level for extended operation has been achieved after which a power distribution map may be obtained.
WOLF CREEK - UNIT I 3/4 2-5 Amendment No. E,92 i
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POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS (Continued) 2.
F/(Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that F/(Z) '
is not increasing.
maximum over z K(Z) f.
With the relationships specified in 4.2.2.2.c above not being satisifed:
1.
Calculate the percent F[(Z) exceeds its limit by the following expression:
Ff F,'(z)X w(Z)" )
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F$" X K(Z) 1-
> X 100 for P10.5
.h wk, P
3 F[ (Z)X V(Z) \\
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F[" X K(2) l -l
> X 100 for P(0.5
,(~
_ 0.5
_)
2.
Either one of the following actions shall be trken:
a.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, control the AFD to within new AFD limits which are determined by tightening both the negative and positive AFD limits of Specification 3.2.1 by 1% AFD for each percent F " Z exceeds its limit and declare the AFD monitor alarm ln(op)erable until the AFD alarm setpoints are changed to the modified limits, or b.
Comply with the requirements of Specification 3.2.2 for F,(Z) exceeding its limit by the percent calculated above.
g.
The limits in Specification 4.2.2.2.c, 4.2.2.2.e and 4.2.2.2.f are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:
1.
Lower core region from 0 to 15%, inclusive, 2.
Upper core region from 85 to 100%, inclusive, WOLF CREEK - UNIT 1 3/4 2-6 Araendment No. E,92
l POWER DISTRIBUTION LIMITS l
3/4.2.3 NUCLEARENTHALPYRISEHOTCHANNELFACTOR-F5 l
4 LINITING CONDITION FOR OPERATION 3.2.3 F5 shall be limited by the following relationship:
F5 s Fj' [1.0 + PF, (1.0 - P)]
- where, Fj' -
TheF511mitatRATEDTHERMALPOWER'(RTP)specifiedin the Core Operating Limits Report (COLR).
PF, -
ThepowerfactormultiplierforF5specifiedintheCOLR, i
THERMAL POWER
, and P
RATED THERMAL POWER Measured values of F[btain a power distribution map.
F5 obtained by using the movable incore detectors to o The measured values of F5 shall be used since an uncertainty
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of 4% for incore measurement of F5 has been included in j
the above limit.
APPLICABILITY: MODE 1 ACTION:
With F5 exceeding its limit:
j a.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either i
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1.
Restore F5 to within the above limit, or j
2.-
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER i
and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
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b.
Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of initially being outside the above limit, verify through incore flux mapping that F5 has been restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
I c.
Identify and correct the cause of the out-of-iinit condition prior to increasing THERMAL POWER above the reduced limit required by Actions a. or b.,
above; subsequent POWER OPERATION may proceed i
provided that FE is demonstrated through in-core flux mapping to be within its limir at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THE".1AL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greatar MTED THERMAL POWER.
WOLF CREEK - UNIT 1 3/4 2-7 Amendment No. 4h n,92
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POWER DISTRIBUTION LIMITS l
SURVEILLANCE REQUIREMENTS 4.2.3.1 F5 shall be determined to be within its limit by using the movable incore detectors to obtain a power distribution map:
n.
Prior to operation above 75% of RATED THERMAL POWER after each feel loading, and b.
At least once per 31 Effective Full Power Days, and c.
The provisions of Specification 4.0.4 are not applicable.
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I WOLF CREEK - UNIT 1 3/4 2-8 Amendment No. 61,92 (Next page is 3/4 2-11) l
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j 3/4.2 POWER DISTRIBUTION LIMITS i
RASES i
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The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:
(a) maintaining the minimum DN8R in the core greater than or equal to the DNBR design limit specified in the CORE OPERATING LIMITS REPORT (COLR) during normal operation and in short-term transients, and (b) limiting the j
j fission gas release, fuel pellet temperature, and cladding mechanical proper-ties.to within assumed design criteria.
In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance l
criteria limit of 2200*F is not exceeded.
The definitions of certain hot channel and peaking factors as used in j
these specifications are as follows:
r,(Z)
Heat Flux Hot Channel Factor, is defined as the local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, J
F5 Nuclear Enthalpy Rise Hot Channel Factor, is defined as the j
ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.
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3/4.2.1 AXIAL FLUX DIFFERENCE j
The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F Z) and F5 limits are not exceeded during either normal operation or in T(he event of xenon redistribution following power changes. The AFD limits have been adjusted for measurement uncertainty.
i Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-l mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the AFD limits and the THERMAL POWER is r
greater than 50% of RATED THERMAL POWER, lj 214.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE t
HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria j
limit.
d WOLF CREEK - UNIT 1 B 3/4 2-1 Amendment 2. h41,92
1 1
POWER DISTRIBUTION LINITS j
BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) f 4
l Each of these is measurable but will normally only be determined j -
periodically as specified in Specifications 4.2.2 and 4.2.3.
This periodic surveillance is sufficient to insure that the limits are maintained provided:
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a.
Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the i
group demand position, b.
Control rod groups are sequenced with overlapping groups as J
described in Specification 3.1.3.6, f.
c.
The control rod insertion limits of Specification 3.1.3.6 are maintained, and d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
i Fg" will be maintained within its limits provided Conditions a. through l
d.abov,egremaintained. The limits on the nuclear enthalpy rise hot channel factor, F, are specified in the COLR.
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L remain,(Z) and F,",ir limits.are measured periodically to provide assurance that the l
F i
within the A peaking margin calculation is performed, when l
1 necessary, to provide the basis for reducing THERMAL POWER or for reducing the j
width of the AFD limits. The hot channel factor f(Z) is measured periodically and increased by a cycle and height d,ependent factor, W(Z), to provide assurance that the limit of F,(Z) is met. W(Z) accounts for '.he l
effects of normal operation transients and is determined from expected power control maneuvers over the full range of burnup conditions in the core. The i
W(Z) functions are specified in the Core Operating Limits Report.
3/4.2.4 OVADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power L
distribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
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The limit'of 1.02, at which corrective ACTION is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt, i
5 WOLF CREEK - UNITS 1 B 3/3 2-2 Amendment No. 4 6L 61,92 7 7 i
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ADMINISTRATIVE CONTROLS i
j CORE OPERATING LIMITS REPORT (COLR) 6.9.1.9. Core operating limits shall be established and documented in the CORE -
OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle, for the following:
1.
Specification 3.1.1.3: Moderator Temperature Coefficient (MTC) EOL limits 2.=
Specification 3.1.3.5: Shutdown Rod Insertion Limit s
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3.
Specification 3.1.3.6: Control Rod Insertion ~ Limits 4.
Specification 3.2.1:
Axial Flux Difference (AFD) l 5.
Specification 3.2.2:
Heat Flux Hot Channel Factor - F,(Z) l l
6.
Specification 3.2.3:
Nuclear Enthalpy Rise Hot Channel Factor -
l F5 l
1 7.
Specification 3.9.1.b: Refuel ng Boron Concentration The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those i
described in the following documents.
a.
NRC Safety Evaluation Report dated October 29, 1992, for the " Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating
+
Station" (ET-90-0140, ET 92-0103)
(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor - F5).
l b.
NkC Safety Evaluation Report dated ~ January 17, 1989, for the
" Acceptance for Referencing of Licensing Topical Report WCAP-11397, Revised Thermal Design Procedure."
(Methodology for Specification 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor - FM.
c.
NRC Safety Evaluation Report dated September 30, 1993, for the l
" Transient Analysis Methodology for the Wolf Creek Generating 1
Station" (ET-91-0026, ET 92-0142, WM 93-0010, WM 93-0028)
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient [MTC))
4 L
d.
NRC Safety Evaluation Report dated November 26, 1993, " Acceptance l
for Referencing of Revised Version of Licensing Topical Report WCAP-10216-P-A, Relaxation of Constant Axial Offset Control - F, Surveillance Technical Specification" (TAC No. M88206).
WOLF CREEK UNIT 1 6-21 Amendment No. 4h 61,92
ADMINISTRATIVE CONTROLS l
CORE OPERATING LIMITS REPORT (COLR) (Continued)
'(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor
- F (Z): Specification 3.1.1.3 - Moderator Temperature Coefficient (MTO): Specification 3.1.3.5 - Shutdown Rod Insertion Limit:
' Specification 3.1.3.6 - Control Rod Insertion Limits:
Specification Enthalpy Rise Hot Channel Factor - F,{fication 3.2.3 - Nuclear-3.2.1 - Axial' Flux Difference: Spec Specification 3.9.1.b -
Refueling Boron Concentration).
e.
NRC Safety Evaluation Report dated March 10, 1993, for the " Reload i
Safety Evaluation Methodology for the Wolf Creek Generating Station" (ET 92-0032, ET 93-0017).
(Methodology for Specification 3.1.3.6 - Control Rod Insertion Limits; Specification 3.2.1 - Axial Flux Difference) f.
NRC Safety Evaluation Report dated March 30, 1993, for the " Revision I
to Technical Specification for Cycle 7" (NA 92-0073, NA 93-0013, NA l
93-0054)
(Methodology for S cification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor - F
[Use of WRB-2 Correlation with VIPRE-01 Code])
l g.
NRC. Safety Evaluation Report dated November 13, 1986, for "The 1981 l
i Version of the Westinghouse ECCS Evaluation Model Using the BASH Code" (WCAP-10266-P-A, Rev. 2)
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor '- F,(Z))
h.
NRC Safety Evaluation Report dated May 17, 1988, " Acceptance for Referencing of Westinghouse Topical Report WCAP-ll596 -
Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores."
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Coefficient (Z): Specification 3.1.1.3 - Moderator Temperature Factor - F (MTC): Specification 3.1.3 5 - Shutdown Rod Insertion Limit:' Specification 3.1.3.6 - Control Rod Insertion Limits:
Specification 3.2.1 - Axial Flux Difference: Specification 3.2.3 -
Nuclear Enthalpy Rise Hot Channel Factor - Fh Specification 3.9.1.b - Refueling Boron Concentration).
i.
NRC Safety Evaluation Report dated June 23, 1986, " Acceptance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP-ANC: A Westinghouse Advanced Nodal Computer Code."
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Coefficienk(Z): Specification 3.1.1.3 - Moderator Temperature Factor - F (NTC): Specification 3.1.3.5 - Shutdown Rod Insertion Limit: Specification 3.1.3.6 - Control Rod Insertion Limits:
Specification 3.2.1 - Axial Flux Difference: Specification 3.2.3 -
WOLF CREEK - UNIT 1 6-21a Amendment No. E,92
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ADMINISTRATIVE CONTROLS l
CORE OPERATING LIMITS REPORT (COLR) (Continued)
Nuclear Enthalpy Rise Hot Channel Fcctor - Fh Specification 3.9.1.b - Refueling Boron Concentration).
The core operating limits shall be detersiried so that all applicable limits (e.g., fuel thermal-hydraulic. limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and 4
accident analysis limits) of the safety analysis are met.
4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, i
to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the L
NRC Regional Office within the time period specified for each report.
i 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
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6.10.1 The following records shall be retained for at least 5 years:
a.
Records and logs of unit operation covering time interval at each power level; 1
b.
Records and logs of principal maintenance activities,. inspections, repair and replacement of principal items of equipment related to i
nuclear safety; c.
All REPORTABLE EVENTS; d.
Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; e.
Records of changes made to the procedures required by Specification 6.8.1; f.
Records of radioactive shipments; i
g.
Records of sealed source and fission detector leak tests and results; and t
h.
Records of annual physical inventory of all sealed source material of record.
1.
1 WOLF _ CREEK - UNIT 1 6-21b Amendment No. M,92 f
s ADNINISTRATIVE CONTROLS RECORD RETENTION (Continued) 6.10.2 The following records shall be retained for the duration of the Unit Operating License:
Records and drawing changes reflecting unit design modifications a.
made to systems and equipment described in the Final safety Analysis Report; b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories; Records of radiation exposure for all individuals entering radiation c.
control areas; d.
Records of gaseous and liquid radioactive material released to the environs; Records of transient or operational cycles for those Unit components e.
identified in Table 5.7-1; f.
Records of reactor tests and experiments; g.
Records _ of training and qualification for current members of the j
Unit Staff;
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h.
Records of in-service inspections performed pursuant to these Technical Specifications; i
l 1.
Records of Quality Assurance activities required by the QA Manuel; j
j.
Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59; i
k.
Records of meetings of the PSRC and the NSRC; 1.
Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.8 including the date at which the service life commences and associated installation and maintenance records;
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Records of secondary water sampling and water quality; and m.
i n.
Records of analysis required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of 2
the analysis at a later date.
This should include procedures effective at specified times and QA records showing that these procedures were followed.
Records of reviews performed for chan ws made to the 0FFSITE DOSE o.
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CALCULATION MANUAL and the PROCESS COMROL PROGRAM.
l WLF CREEK UNIT 1 6-22 Amendment No. 42 4
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