ML20235E706
| ML20235E706 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/31/1988 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20235E705 | List: |
| References | |
| NUDOCS 8902220110 | |
| Download: ML20235E706 (20) | |
Text
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CRYSTAL' RIVER UNIT 3 ANNUAL REPORT OF FACILITY CHANGES, TEST, AND EXPERIMENTS IN ACCORDANCE WITH 10CFR50.59(b).
JANUARY 1 - DECEMBER 31, 1988 In-the attached report, each number refers to the Safety Evaluation questions listed below:
1.
Is-the probability of occurrence or the consequences of an accident or malfunction evaluated in the FSAR increased?'
YES NO 2.
Is the possibility for an accident or malfunction of a different type than.any evaluated-in the FSAR created?
YES
, NO 3.
Is the margin of ' safety, as defined in the bases for any Technical Specification reduced?
YES
' NO l,
1 t
8902220110 09021o DR ADOCK 0500o3o2 PDC
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I 10 CFR 50.59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION 6.3.1 1.
No, alignment of the "3AB" MCC to either safety train is presently l
accomplished via a
manual transfer switch.
To alleviate the automatic loading on the 3A Emergency Diesel Generator resulting from a coincident LOOP and
- DBA, the 3AB MC is procedurally committed to be connected to the "B" train.
In the event of a single failure associated with the "B"
train (i.e.
loss of emergency diesel generator 3B) a loss of power to the 3AB MCC will occur until operator action is taken to realign the MCC to the "A"
train while maintaining the emergency diesel generator 3A within its 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating.
The following is a listing of the Safety Related equipment supplied from the 3AB MCC and a
description of the impact resulting from their unavailability due to loss of power immediately following an accident.
Tac No.
Description Jmoact AHF-1C Reactor Building Cooling No Impact - Redundancy Fan provided by AHF-1A CFV-5 CF Tank 3A Outlet ISO No Impact - Valve is Valve mechanically maintained in the open position during Plant Operation CFV-6 CF Tank 3B Outlet ISO Same as CFV-5 Valve MUP-2B MU&P Pump 3B Main No Impact - Redundancy 2
10 CFR 50.59 EVALUATIONS PERFORMED "J FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION Tao No.
Description Imoact Lube oil Pump provided by MUP-1A and i
related support MUP-4B MU&P Pump 3B Main Gear Same as MUP-2B Oil PuIrp MUV-18 Reactor Coolant Pump Seal No Impact - Redundancy of ISO Valve isolations provided by check valve MUV-27 MU ISO Valve to Reactor Same as MUV-18 Inlet Lines Loop A (Normal Makeup Line)
DWV-160 Reactor Building Same as MUV-18 Demineralized Water Supply Valve DHV-41 DH Removal Outlet No Impact - Redundancy Containment ISO Valve provided by DHV-3, 4
(Inside Containment) and DHV-39, 40 (Outside Containment)
DHV-91 DH to Press Spray Outside Same as MUV-18 ISO Valve DHV-7 DH Discharge Crosstie Valve No Impact - Valve can be manually operated if required prior to Emergency Diesel Generator 3A capacity being made available DHV-8 DH Discharge Crosstie Valve Same as DHV-7 2.
No, as this procedural realignment does not establish any new methods of plant operation or require any physical modifications, the possibility of an accident or malfunction of a different type than any evaluated 3
10 CFR 50.59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION previously in the 1M hits not been created.
3.
No, although the margin of safety has not been specifically addressed in the technical specification bases this procedural change is considered to have no effect on the existing plant safety margins.
12.1.2 1.
No, job title changes and responsibilities have been approved by management.
The
- Director, Nuclear Plant.
Operations (DPO) responsibilities have not changed and the probability or consequences of an accident or malfunction have not increased.
2.
No, changing job titles under the DPO does not create new accidents.
3.
No, the technical specifications do not define the margin of safety when applied to management position responsibilities.
7.1.3.3.4 & Table 7-2 1.
No, the change to Table 7-2 correct the trip terminology to agree with Tech. Specs. Section 7.1.3.3.4 changes correct als error in the number of tests performed on output relays.
These changes do not affect the probability or consequences of an accident.
l 2.
No, these changes do not create new accidents or malfunctions.
Table 7-2 is corrected to insure that the FSAR and Tech. Specs.
i agree.
Section 7.1.3.3.4 is an l
4 1
.s,.
10 CFR 50.59 EVALUATIONS PERFORMED ON j
FSAR TEXT CHANGES IN ACCORDANCE WITH j
PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION error correction.
]
3.
No,
the changes bring the FSAR into agreement with Tech. Specs.
which are approved by the NRC.
The correction of an error in Section 7.1.3.3.4 does not reduce the margin of safety.
Sections 5.5.2, 9.4, 9.5 and l'.
No,
raising the Ultimate Heat 0
Tables 4-7, 6-10, 9-8, 9-9, Sink (UHS) temperature to 95 F 9-12 does not increase the probability or consequences of an accident.
The revised operating parameters do not physically alter any
- system, and will not require a revision to any plant operating procedure, thus the provability of an accident has not been increased.
There can be no increased consequences
'ollowing an acc.ident without an increase in any parameter related to boundary performance.
" Boundary" refers to the
- RCS, fuel / cladding and containment.
Two (2) parametric increases have been identified and evaluated.
Containment
- pressure, the only parameter which must be considered in evaluating containment boundary performance, is not influenced by RB cooler performance.
The Building Spray system controls post-LOCA RB pressure, and any reduction in RB cooler heat removal rate is not considered significant.
Temperature is the only parameter impacting fuel / cladding boundary performance which c..
be 5
i 10 CFR 50.59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION influenced by an increase in UHS temperature, and then only during normal cooldown and holding modes of operation.
The temperature of the fuel / cladding during normal operation is much higher than those temperatures found during cooldown and holding
- modes, provided by the decay heat removal rate exceeds the rate of decay heat production.
Gilbert calculation DCC-0428-5510J117-
- 002, REv.
1 confirms that the decay heat removal rate exceeds-the rate of decay heat production at system operating temperatures resulting from UHS temperatures 0
of 95 F.
2.
No, by raising the maximum permissible UHS temperatur.
to 950 F, the temperature the s_
cooling water supplied to safety related components will be slightly higher than design for a short period of time following an accident.
Cooling water temperatures in excess of design could challenge component operability and degrade system performance.
- However, every safety related component was requalified for service at-the elevated cooling water temperature (Ref.
G/C report "SW/DC/RW Vendor Qualification Data" and B&W Doc.
No.
51-1172534-00).
Since all impacted systems will fulfill their safety
- function, raising the UHS temperature limit will not create a new accident scenario.
3.
No, the analyses performed to confirm operability at elevated 6
l i
{
l
10 CFR 50.59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH 1
PROCEDURE NOD-11 l
1 i
FSAR SECTION IDENTIFICATION / SAFETY EVALUATION I
UHS temperatures demonstrates the inherent overdesign of the SW, DC and RW systems.-
Raising the upper limit of UHS temperature reduces the amount of overdesign, 1
but maintains a
sufficient margin-of safety since the operability of all equipment was confirmed at system operating temperatures resulting from a UHS 0
95 F.
In all temperature of probability, the new UHS 0
temperature limit of 95 F will never be challenged--seawater temperature at CR-3 has exceeded 0
0 90 F (90.5 F max) only three (3) hours in ten (10) years.
Section 5.2.2.6 1.
No, the insulation on the SW lines inside the Reactor Building provides anti-sweating protection for these lines only.
It does not affect function or the consequences of an accident.
2.
No, the insulation has been evaluated and found to be able to withstand a
LOCA and the requirements of SP-5953, the original requirements for SW system insulation inside the Reactor Building.
3.
No, insulation on the SW lines inside the Reactor Building is not addressed in any Technical Specification.
Insulation of these lines provides anti-sweating protection.
Section 5.6.4.1, 2,
& 3 1.
No, the probability of an accident or malfunction previously evaluated in the FSAR occurring will not be increased since this FSAR change does not 7
=
10 CFR 50.59' EVALUATIONS PERFQRMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 l
l' j
FSAR SECTION IDENTIFICATION / SAFETY EVALUATION l
describe a modification'~of any equipment - or a. change in i system -
operating conditions..
The ' consequences. of an accident or malfunction previously evaluated in the FSAR will'not be
. increase since the -radiological consequences of a Design Basis LOCA. and a f Maximum Hypothetical Accident have been analyzed at a containment
'l e a k -
rate corresponding to -the design pressure and temperature of the reactor building.
Increasing the test pressure to a value - still below design pressure ' : ensures that the leak. rate E used 'in the analysis still envelopes the leak rate at worst case post-accident containment pressure.
2.
No, this FSAR change' -will not create an accident or malfunction different than that evaluated in the FSAR since no equipment has.
been modified and no system operating conditions have been changed.
3.
No, the margin of ' safety,.as defined in the bases of the Technical Specifications will not be reduced since the new test pressure is still below design pressure
- and, therefore, the containment leak rate used in the accident analyses (based on design pressure) is still enveloping.
Sections 14.0.1 & 14.0.2 1.
No, Each fuel reload is evaluated to insure that the probability and consequences of an accident or malfunction is not increased.
8
1 10 CFR 50.59 EVALUATIONS-PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION l
l j
L 2.
No, Each Chapter 14 accident 'is i
evaluated against the fuel cycle to insure that accidents or malfunction different from those previously evaluated are not created.
3.
No, the T.S.
margin of safety is not affected by adding Cycles 6 and'7 to the Chapter 14 tabular data.
Section 5.2.5.2.3.1.j 1.
No,' this change is not a design modification.
The personnel airlocks.
were originally acceptance tested at 1.25 times design pressure in the interspace between the doors.
This FSAR i
change is a correction to make the FSAR agree with the design specification.
2.
No, no' new system functions or operating conditions have been created.
This FSAR change is a j
correction to make the FSAR agree
~
with the design specification.
3.
No, the interspace between the airlock doors is periodically tested at the maximum calculated post-accident containment pressure, Pa, in accordance with plant technical specifications.
The airlocks will continue to be tested at Pa.
This FSAR change is a correction to make the FSAR agree with the design specification.
1 Section 7.4.5 1.
No, this modification does not alter the systems or activities directly involved in operating the plant.
Therefore, the series 9
i
j 0
10 CFR 50.59 EVALUATIONS PERFORMED ON FSAR TEXT-CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION
'of events leading to a previously evaluated accident are not changed.
'2.
No, existing safety. systems allow the plant to be remotely shut down in the event the Control Room is totally uninhabitable.
This accident scenario is more conservative than any accident possible due to the modification.
This modification does not alter any plant systems - or components which are directly involved in operating the plant.
2.
No, the types of accidents possible due to the modification have been addressed in the FSAR.
This modification does not alter the function-of :any plant component.
3.
No, this modification does not affect any of the Plant Fire Suppression Systems as defined in Tech Spec, Sect.
3/4.7.11, nor does it exceed the maximum fire loading for the Control Room Area listed in the FHSA.
4.2.4.4 1.
No, accident mitigation assumes the availability of pressurizer heaters to maintain pressure.
This description tells how the alignment of emergency power is made to the heaters.
2.
No, this description provides a discussion of how the pressurizer heaters are supplied with power in normal and emergency conditions.
3.
No, this description does not 10
', [,-
10 CPR 50.59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH
. PROCEDURE NOD-11
-FSAR SECTION IDENTIFICATION / SAFETY EVALUATION affect the Technical Specifications or the margin of safety.
9.11 1.
No, this change to..the FSAR adds a
description of the PASS installed to satisfy NUREG-0737, Item II.B.3.
2.
No, this text addition describes a system installed to provide a means to sample and analyze the RCS after an accident.
3.
No, this text change does not have anything to do with the Technical Specification margin of safety.
7.1.3.2.4 1.
No, Procedure CP-123 does not alter any analysis performed in the FSAR.
This change is' to insure compliance with FSAR Chapter 5.
Administrative control of ES channel and actuation cabinet doors does not alter the performance of the ES System.
2.
No, the function of the ES System is not altered by administrative control of ES channel and actuation cabinet doors.
3.
No, Procedure -CP-123 does not alter equipment operation.
12.4 1.
No, none of the changes reduce the effectiveness of the Radiological Emergency
Response
Program and do not address safety-related equipment.
2.
No, many of the changes are administrative in nature and/or 11
10 CFR 50,59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANC2 WITH PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION are the result of revisions to implementing procedures which do not address
. safety-related equipment.
3.
No, the changes serve to enhance the Radiological Emergency
Response
Program and do not affect the basis for any Tech Specs.
1.4, 1.5.8, 6.1.3 1.
No, these changes only clarify the CR-3 licensing basis and do not affect previously evaluated FSAR information.
2.
No, these changes are editorial and do not change accident scenarios.
3.
No, these changes clarify CR-3 licensing basis and do not change the Technical Specifications.
7.2.4 1.
No, ASV-5 and ASV-204 are motor operated valves having identical functions of supplying steam to the turbine driven Emergency Feedwater Pump (EFP-2).
Since EFP-2 is the ES "B" channel pump, ASV-5 and 204 were electrically connected in parallel to a common 250/125 VDC ES "B"
channel power and control source.
This modification electrically separates ASV-204 from ASV-5 and repowers ASV-204 from 250/125 VDC ES "A"
channel power.
- Also, separate control room controls and separate "A"
channel EFIC interlocks are being provided for ASV-204.
Automatic control logic of ASV-204 has not changed.
Therefore, the probability of an occurrence or the consequences of 12
10 CFR 50.59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 FSAR SECTION IDENTIFICATION / SAFETY EVALUATION an accident or malfunction of equipment important to safety as previously evaluated in the FSAR is not increased since the logic of automatically opening ASV-204 whenever the EFIC System calls for emergency feedwater has not been altered.
The reliability of EFP-2 has actually been increased because with this modification either "A"
or "B"
train power will control and operate one of the steam inlet valves to EFP-2 as opposed to both valves being "B" trained powered.
2.
No, the electrical separation of ASV-204 from ASR-5 does not impact the design function of either valve to supply steam to the EFP-2 turbine.
Power and control for ASV-5 is not affected by this modification and ASV-5 retains its automatic control
- logic, remote manual
- control, local manual control and remote shutdown isolation and control.
ASV-204 is being powered from the redundant power channel, and will be provided with its own remote and local manual control and with separate EFIC interlocks for automatic operation.
The type of manual control and automatic operation of ASV-204 is the same as for ASV-5.
Therefore, based on the above, the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR is not created.
3.
No, this modification enables the turbine driven Emergency Feedwater Pump (which is the "B"
13
7
.10 CFR 50.59 EVALUATIONS PERFORMED Ol{
FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE NOD-11 FSAR SECTION.
IDENTIFICATION / SAFETY EVALUATION channel pump) to be operational even if a. failure should occur on the "B"
channel power system.for which shutdown operation would be via the "A"'
channel. systems.
With.this capability, the turbine -
driven EFW pump. is able to operate
.and share the EFW requirements'with the "A" channel motor driven EFW pump.
This.will-reduco the electrical-load on the "A"
channel diesel. generator for the condition 'of an ES actuation coincident 'with a
loss-of-offsite-power and failure of the "B"
channel-power system.
Consequently, with
.this modification the. margin of.
safety, as defined in the basis for any Technical Specification, is not reduced.
It is actually enhanced because of the increased availability of the turbine driven Emergency Feedwater Pump.
3.2 1.
No, Section-3.2 of the FSAR deals with reactor design.
The text revisions update the' FSAR to reflex current Cycle 8 conditions for which we have already received approval from the NRC.
2.
No, the Section of the FSAR reviewed here deals with reactor design for which FPC has received approval for CR-3's current design from the NRC.
Approval was received on December 14, 1987 with Amendment #103 to the CR-3 operating license.
3.
No, the changes here only effect the FSAR which is being updated to reflex current licensed conditions.
14
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10 CFR 50.59-EVALUATIONS PERFORMED ON PIANT MODIFICATIONS l
IN ACCORDANCE WITH MAR PROCEDURES FACILITY MODIFICATION DESCRIPTION / SAFETY EVALUATION Increased Pressurizer Level 1.
No, FPC investigated the effects (87-03-09-01) of increasing the normal operation pressurizer level form 200" to 220" and found no increase in the probability or consequence of an accident or malfunction.
2.
No, FPC's evaluation of changing pressurizer level did not reveal the possibility for an accident.
or malfunction of a
different type.
3.
No, there was no reduction in the margin of safety based upon the analysis.
Remove Incore Monitoring 1.
No, incore themocouples provide Panel (IMT-1) an indication of core temperature (88-01-25-01) during normal plant operations and transients.
They do not increase malfunction of equipment important to safety as evaluated in FSAR Sections 1.3.2.5 or 7.3.3.2.1.
2.
No, since the incore monitoring and data recording is performed
)
by the recall system and recorders on the main control board no possibility for an accident or malfunction different than any previously evaluated in the FSAR is created.
3.
No, the instrument (IMT-1) being removed is not required by Tech.
Spec.
3/4.3.3.2, 3/4.3.3.6 or Table 3.3-10 therefore, the margin of safety is not required.
Automatic Closure 1.
No, this modification does not of Fire Dampers involve any equipment considered (86-02-09-01) to be important to safety.
15
[.'
10 CFR 50.59 EVALUATIONS PERFORMED ON PLANT MODIFICATIONS IN ACCORDANCE WITH MAR PROCEDURES FACILITY MODIFICATION DESCRIPTION / SAFETY EVALUATION 2.
No,
this modification does not involve or affect any previously evaluated accident or malfunction addressed in the FSAR.
3.
No, the detection of TB switchgear room smoke and closure of fire dampers does not affect any Tech. Spec. defined margin of safety.
Addition of Radwaste 1.
No, the majority of CR-3's liquid Demineralized radwaste disposal system is (87-10-21-01) classified as non-safety related per figure 11-1 of the FSAR.
I Exceptions to this are the containment penetrations.
The demineralized modification interfaces with existing CR-3 equipment only in two location; the discharge of WDP-6A, 6B and the suction of WDP-14A, 14B; both of which are within the non-safety related region of the system.
The equipment added to the existing system will perform no function important to safety as it is not a
part of the reactor coolant pressure boundary; it is not required for the safe shutdown of the reactor or to maintain in a
safe configuration; and it is not designed to prevent or mitigate the consequences of an accident that could result in offsite exposures.
2.
CR-3's radioactive waste and holdup processing equipment is housed within the Auxiliary Building; a reinforced concrete Seismic Class 1
structure i
designed to withstand tornados, 16
1 t
5 s
10 CFR 50.59 EVALUATIONS PERFORMED ON PLANT MODIFICATIONS IN ACCORDANCE WITH MAR PROCEDURES FACILITY MODIFICATION DJ_SCRIPTION/ SAFETY EVALUATION l
tornado driven missiles and the Maximum Hypothetical Earthquake.
The NUS demineralized and its interface points are housed
.within the " Yellow Room", a room I
located inside this Seismic 1
i structure.
(REF. FSAR sec. 11.2, Reg. Guide 1.143 para. 1.1.3)
To follow the concept of ALARA, those pieces of processing equipment prone to having high dose rates,(demin beds) have been located behind a shield wall with.
I no access other than from above.
j In this manner possible exposure-l is virtually eliminated.
In the event of component failure or spillage due to operational
- error, each area of the Yellow Room ie equipped with a dike and a floor drain.
The floor drains are routed to the auxiliary building sump 'such that any spillage is captivated and can be reprocessed through the radwaste system.
The Yellow Room is also vented to the auxiliary building HVAC system at a rate of 5100 CFM and processed through RM-A8 with final monitoring at RM-A2.
]
Should there be a gaseous release with activity levels exceeding the set point of RM-A2, (5.4E3 uci/sec of Kr-85) an interlock will shutdown the supply fans such that the release is minimized.
Normal process effluent from the demineralized equipment is sent to WDT-10A, 10B where it is then processed for discharge through RM-L2 utilizing the standard release procedures.
(FSAR sec.
11.2.1.2)
Spills of contaminated resin during resin transfer operations are minimized 17
10 CFR 50.59 EVALUATIONS PERFORMED ON PLANT MODIFICATIONS IN ACCORDANCE WITH MAR PROCEDURES FACILITY MODIFIC1 TION DESCRIPTION / SAFETY EVALUATION by level indication and through the use of a television camera j
incorporated into the fill head j
of the solidification system.
To prevent the inadvertent release of radioactivity during dewatering operations, air is pulled through the vessel and j
filtered through a coalescer and J
a high efficiency filter prior to exhausting to the Yellow Room atmosphere.
Vacuum is used in the dewatering process such that should a line break occur, any leakage will be into the system, not out.
The set points of RM-A2 and RM-L2 have been established to ensure that the estimated radioactive releases to the environment will not cxceed those evaluated in section 11.2.4 and Appendix 11A of the FSAR.
The afore-mentioned design features and set points have been established to mitigate the consequences of component failure.
To reduce the possibility of component failure, all pressure retaining components, such as
- vessels, piping and hoses, are designed and fabricated in accordance with the code and standards listed in Table 1 of Reg. Guide 1.143.
The failure mechanisms for this equipment are no different than those for the rest of the radwaste system.
Sufficient plant features already exist to prevent a radiological release should a component fail.
These features have been evaluated and can meet the requirements of 10 CFR 100 as documented in both Appendix 11A l
l 1
18
l 10 CFR 50.59 EVALUATIONS PERFORMED ON PLANT MODIFICATIONS J_N ACCORDANCE WITH MAR PROCEDURES FACILITY MODIFICATION DESCRIPTION / SAFETY EVALUATION and Appendix 11B of the FSAR.
3.
No, liquid radwaste discharge to unrestricted areas must meet the requirements of Tech.
Spec.
3/4.7.13.2,
'3 / 4.11.1.1, and 3/4.11.1.2 which are based on the requirements found in 10 CFR 50 Appendix I.
Utilization of the NUS domineralizer for processing miscellaneous waste will provide compliance with these technical specifications.
EFW Thermal Upgrade 1.
No, At the higher system design (88-10-14-01) temperatures given in this MAR the EFW system will maintain its original design function as defined in FSAR section 10.5.
- Thus, the probability of increasing an accident occurrence or increasing the consequences of an accident as defined in FSAR section 10.5.1.a, b,
c, d,
& e does not exist.
Reference also Section 5.4.4 for piping design basis.
2.
No, this modification increases the design temperature of the EFW system.
This increase is within the original design intent of the l
EFW system and does not add any new system, operation, or process that would create the possibility for an accident or malfunction different than any previously evaluated in the FSAR, section 10.
3.
No, the function of the EFW system as defined in Tech Spec 3.7.1.2 has not changed as a
result of confirming the integrity of the system at the 19
a 10 CFR 50.59 EVALUATIONS PERFORMED ON PIANT MODIFICATIONS IN ACCORDANCE WITH MAR PROCEDURES FACILITY MODIFICATION DESCRIPTION / SAFETY EVALUATION higher temperatures specified in this MAR.
Thus the margin of safety has not been reduced.
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