ML20205T654
| ML20205T654 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire, 05000000 |
| Issue date: | 06/09/1986 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20205T651 | List: |
| References | |
| TAC-61680, TAC-61681, NUDOCS 8606160129 | |
| Download: ML20205T654 (14) | |
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ATTACHMENT I PROPOSED CHANGES TO A) McGUIRE TECHNICAL SPECIFICATIONS B) CATAWBA TECHNICAL SPECIFICATIONS 1
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B606160129 860609 PDR ADOCK 05000369 P
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
No PRESSURE BOUNDARY LEAKAGE, a.
b.
1 gpm UNIDENTIFIED LEAKAGE, I gpm total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator, c.
10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, d.
40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of e.
2235 1 20 psig, and 2235 1 20 psig 1 gpm leakage at a Reactor Coolant System pressure of f.
from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Ab a.
any Reactor Coolant System leakage greater than any one of t h
b.
9 *h om d dian PRESSURE BOUNDARY LEAKAGE an Reactor Coolant System Pressure Asvi- '-- 'Q, reduce the leakage above limits, ours or be in ac ;-_:t unT STANDBY g
rate to within limitours and in COLD SHUTDOWN within the within the With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of c.
the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
P SEE risk f M 6 Jt McGUIRE - UNITS 1 and 2 3/4 4-19
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F b.
With primary-to-secondary leakage, IDENTIFIED LEAKAGE, or CONTROLLED LEAKAGE greater than the appropriate limit, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
d.1 With UNIDENTIFIED LEAKAGE greater than 1 gpm, but less than or equal to 3 gpm, identify leakage or reduce to less than or equal to 1 gpm within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
In addition, verify that the total of UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, and Reactor Coolant Pump Seal Leakage is not greater than 26 gpm, or declare the Standby Makeup Pump INOPERABLE.
d.2 With UNIDENTIFIED LEAKAGE greater than 3 gym, identify or reduce leakage to less than or equal to 3 gpm within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and take ACTION dl, above, or be in at least HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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PO 4 ppd-gh REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued)
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to 10 CFR Sections 50.72 and 50.73 prior to resumption of
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plant operation.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory exami-nations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is O
expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of.less than 1 gpm.
This threshold value is sufficienti low to ensure early detection of additional leakage.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited l
amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in the accident analyses.
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The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break.
The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents.
The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
Amendment No.52((Unit 1 McGUIRE - UNITS 1 and 2 B 3/4 4-4 Amendment No.33 Unit 2)
A leak which develops under operating temperatures and pressures may become virtaully undetectable under Cold Shutdown conditions. Therefore, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are allowed to find the leak before compensatory measures are required, i
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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 gpm UNIDENTIFIED LEAKAGE, 1 gpm total reactor-to-secondary leakage through all steam generators c.
and 500 gallons per day through any one steam generator, d.
10 gpm IDENTIFIED LEAKAGE from the Reactor. Coolant System, 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of e.
2235 1 20 psig, and f.
I gpm leakage at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
I APPLICABILITY:
MODES 1, 2, 3, and 4 ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY a.
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With any.t -" ar Coolant System leakage greater
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^.h-, gb above limits, excluaing .:""DF "CL:Gniu LEAKAGE and leakage from gf gg Reactor Coolant Sys sure Isolar.i reduce the leakage STANDBY g/ g rate to wit ts within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at wi next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the fo
- J hours.*
pg With any Reactor Coolant System Pressure Isolation Valve leakage l
c.
greater than the above limit, isolate the high pressure portion of t +
I-the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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operation in Modes 3 and 4 p < 5 gpm. g b 2
'un
- ^^'? H"es. October 23, 1985,=%ntified leakagegM vviii but with the Reactor Cooian6 Sp.;;
^ 7 ' ca= by the above time, M If the unidentified leakage rate is not ren-the...M. '; :..,. ; ;; :.. ;Gw anuluuWN within the following e oud f5b CATAWBA - UNITS 1 & 2 3/4 4-20
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b.
With primary-to-secondary leakage, IDENTIFIED LEAKAGE, or 4
CONTROLLED LEAKAGE greater than the appropriate limit, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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d.1 With UNIDENTIFIED LEAKAGE greater than 1 gym, but less than or
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equal to 3 gym, identify leakage or reduce to less than or equal to 1 gpa within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the 4
following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
In addition, verify that the total of UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE, and Reactor Coolant Pump Seal Leakage is not greater than 26 gpe, or declare the Standby Makeup Pump INOPERABLE j
and apply Specification 3.7.13.a. " Standby Shutdown System".
l d.2 With UNIDENTIFIED LEAKAGE greater than 3 gpa, identify or reduce leakage to less than or equal to 3 gpa within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and take ACTION dl, above, or be in at least HOT STANDBY within the l
following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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BASES These Dete: tion Systems are consistent with the recommendations of Regulatory
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Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems,"
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May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE i
l PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
Industry experience has shown that while a limited amount of leakage is expected from the Reactor Coolant System, the unidentified portion of this leakage can be reduced to a threshold value cf less than 1 gpe. This thres-hold value is sufficiently low to ensure early detection of additional leakage.
The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the Reactor Coolant System ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break.
The I gpm limit is consistent with the assumptions used in the analysis of these accidents.
The 500 gpd leakage limit per steam d
generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
l The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited l
amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal Reactor Coolant System pres-sure of 2235 psig.
This limitation ensures that in the event o" a LOCA, the safety injection flow will not be less than assumed in the safety analyses.
The 1 gpm leakage from any Reactor Coolant System pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure.
It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required.
Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure.
The Surveillance Requirements for Reactor Coolant System pressure isolation valves provide added assurance of valve integrity thereby reducing the prob-ability of gross valve failure and consequent intersystem LOCA.
Leakage from the pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
CATAWBA - UNITS 1 & 2 8 3/4 4-4
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1 A leak which develops under operating temperatures and pressures may become virtually undetectable under Cold Shutdown conditions. Therefore, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cre allowed to find the leak before compensatory measures are required.
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ATTACHMENT II TECHNICAL JUSTIFICATION This change would facilitate identification of small leaks while the RCS is l.
hot and pressurized. A small (between 1 and 3 gym) leak at Mode 1 or 2 temperatures and pressures may become virtually undetectable at Cold Shutdown conditions. Current requirements to begin shutdown within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the detection of a leak greater than 1 gym were predicated upon concerns that a small leak would suddenly and uncontrollably result in a large LOCA. Recent analyses (references 1 and 2) have shown that such is not the case. There-fore, it is considered appropriate to increase the allowed time of operation, with a leakrate between 1 and 3 gym, to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The complexity of modern plants such as McGuire and Catawba may preclude a speedy identification or isolation of a small leak. The myriad of instrument lines and other connec-tions to the Reactor Coolant System (RCS) create numerous possibilities for leaks, each of which must be investigated until the source of the leak is found. The search effort is impeded by requirements for protective equip-ment, and the design of the ice-condenser containment at each station. The containment is small and compartmentalized, necessitating a painstaking and thorough search, which requires more than the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> currently allowed.
An upper limit of 3 gym was chosen to permit continued operation (for up to 1
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) on the basis of historical need, confidence in the ability of makeup systems to compensate for the leak, and because flaws associated with a 3 gpa leak are most demonstrably not precursors to catastrophic failures of RCS piping, while still providing a higher threshold of leakage which may be tolerated before entry to lower modes must be initiated.
Past events at both stations, involving leaks in the 1-2 gym range, have resulted in either unnecessary shutdowns or the unnecessary expenditure of 3
both Duke and NRC staff resources in the pursuit and granting of discretion-ary enforcement to allow continued operation.
This change would obviate the need for " crisis mode" response to situations which have been demonstrated to I
have no significant adverse safety implications.
According to MNS FSAR Section 5.2.7.5, the ratio of allowable leakage to normal makeup is given as:
.005,
2 gpm Lu l
Rum
=
=
=
M 398 gpa Where:
Run = Ratio of Unidentified Leakage Rate to Normal Makeup Rate Lu = Unidentified Leakage Rate M
= Normal Makeup Rate Increasing Lu to 3 gpm,
.008 3 ape Rum
=
=
398 gym
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. This increase in the ratio of unidentified leakage rate to makeup. rate can be compared to the ratio of assumed identified leakage to makeup capacity
(.025). The additional required makeup capacity is only 10% of the capacity assumed available for identified leakage. Typical station practica does not involve operation with identified leakage approaching 10 gym; therefore, the additional allowed leakage does not place a probable additional demand on the system.
As noted above, current requirements are-based upon IOCA concerns.
In accordance with current leak-before-break theory, pipe breaks will be pre-ceded by leaks of increasing magnitude.
Analyses performed for Catawba and McGuire Nuclear Stations show that flaws of a size which produce leaks of the magnitudes associated with this Tech Spec are not expected to propagate.
In the Catawba analysis (the more conservative of the two analyses with respect to flaw size vs. leak rate) it has been calculated that a flaw of critical size (that which may be expected to propagate) is 32.5 inches. A flaw size of 3.2 inches was calculated to produce a leak of I spa, which would initiate the LCO. Similarly, a flaw size of less than 5 inches is calculated to produce a leskrate of 3 gym. An unidentified leak which is first detected in the range of 1-3 gpa is not expected to grov quickly (within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) due to the relatively small size ( 5 inches) of the associated flaw with respect to the critical size flaw (32.5 inches).
Furthur, it is extremely unlikely that a flaw would grow from the 3 gym leak-size flaw ( 5 inches) to critical size in the 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> allowed to achieve Cold Shutdown.
Catawba and McGuire each have requirements regarding the Standby Makeup Pump (McGuire's is administratively implemented, and Catawba's is Specification 3.7.13a).
Both require that if the total of identified, unidentified, and reactor coolant pump seal leakages is greater than 26 gym, the Standby Makeup Pump is declared inoperable. That initiates a 7-day Limiting Condition for Operation, to make the pump operable (in this case, reduce leakage) or comply with the associated action statement.
The 26 gpm includes 3.5 gpm seal leakoff per RCP,10 gpa Identified Leakage,1 gpm Unidentified Leakage, and.1 gpa extra margin. It is not anticipated that the additional allowed unidentified leakage will cause the total actual system leakage to exceed 26 gpa, because, as noted above, it is not station practice to operate with identified leakage approaching 10 gpa.
Since the time allowed to operate with unidentified leakage is less than that allowed for an inoperable SS, this change does not create a significant additional safety concern.
Given that the revised leakage rates do not constitute a challenge to the normal system makeup capability, and are not considered to be precursors to catastrophic failures of RCS piping, no significant detriment to safety is being introduced by this change.
In addition, a footnote in Catawba's Tech Specs on Page 3/4 4-20, the applicability of which expired on October 23, 1985, is being deleted. This change is administrative in nature and has no safety significance or Significant Hazards Consideration.
ATTACHMENT III l
In accordance with 10CFR50.91, following is an analysis of the proposed J
amendment with respect to the criteria of 10CFR50.92; i.e.,
significant hasards considerations.
1)
The proposed amendment will not significantly increase the probability or consequences of an accident previously evaluated in the FSAR.
The only credible accident which could increase in probability would ts a Loss-of-Coolant Accident (LOCA). The proposed change has no effect l
until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the event (i.e., operation may continue without l
commencing shutdown), therefore the probability of any event is unaf-fact ed.
The magnitude (" consequences") of the LOCA are not expected to increase significantly because:
a)
The volume of coolant which could be released (at the rate of 3 gpa) is well within the capability of the containment sump pumps to The containment sump pump removal rate is 200 gym (refer-remove.
ence MNS FSAR Section 5.2.7.5, CNS FSAR Section 11.2.2.2).
.l b)
Analyses have shown that leaks caused by flaws in piping systems will not increase rapidly or result in catastrophic pipe rupture.
(References 1 and 2, Attachment IV) c)
The leak is not expected to cause total leakage to exceed the capacity of the Standby Makeup Pump, because the allowed uniden-tified leakage will be bounded by existing allowances for total leakage as defined in the respective Standby Shutdown System Tech Specs ; or the LCO associated with those Tech Specs will be initi-ated.
d)
The consequences of LOCAs have been exhaustively analyzed in Chapter 15 of each stations' FSAR. The consequences of any leak to which this change is applicable are bounded by these analyses.
2)
The proposed amendment will not create the possibility of any new accident not previously evaluated in the FSAR.
This change does not affect the design of the station, so it does not create the possibility of any new accident. The respective Final Safety Analysis Reports for McGuire and Catawba sufficiently examine the consequences of all credible accident scenarios to which the stations' l
designs are susceptible.
3)
The proposed amendment will not significantly reduce any margin of safety.
I Margins of safety which could be affected by this change are:
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a)
Potential increase in time available for leaks caused by piping flaws to cause catastrophic failure of piping.
b)
Potential to exceed the capacity of Standby Makeup Pump, thus j
rendering the Standby Shutdown Facility (SSF) inoperable.
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. Item a, catastrophic failure of RCS piping has been exhaustively analyzed by Westinghouse and reviewed by the NRC in various submittals (e.g., References 1 and 2).
Consistent with current leak-before-break theory, the amount of time (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) available will not allow a small flaw to propagate. There-fore, this margin of safety is not reduced.
Item b, exceeding the capacity of the Standby Makeup Pump, is possible only if all three contributors to the SMP flow requirements, i.e., Identified Leakage, Unidentified Leakage, and RCP seal leakage, are at their maximum values concurrently. In that event, pursuant to each stations' existing SSF requirements, the SSF is declared inoperable and a 7-day action is entered.
Use of an existing action for an identical LCO does not constitute a reduc-tion in a margin of safety.
Therefore, Duke Power Company concludes that the proposed changes do not involve a Significant Hazards Consideration.
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ATTACHMENT IV REFERENCES 1.
Letter, H.B. Tucker to H.R. Denton, August 30, 1985, " Pipe Break Cri-teria Relief for Reactor Coolant Loop" (McGuire) 2.
Letter, H.B. Tucker to H.R. Denton, May 11,1984, (untitled) (Catawba) i i
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