ML20235T807
| ML20235T807 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/31/1988 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-89-26 VPNPD-89-112, NUDOCS 8903080484 | |
| Download: ML20235T807 (126) | |
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l WISCONSIN _ ELECTRIC POWER COMPANY ANNUL >L RESULTS'AND DATA REPORT 1988 FOR THE
-POINT: BEACH NUCLEAR PLANT
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.l UNITS NOS.'l AND"2 Docket Nos. 50-266~and 50u301 l
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-Facility Operating Licenses DPR-24 and DPR-27 j
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e PREFACE This Annual Results & Data Report for 1988 is submitted in accordance with' Point Beach Nuclear Plant, Unit Nos. 1 & 2, Technical Specification 15.6.9.1.B and filed under Docket Nos. 50-266 & 50-301 for Facility Operating License Nos. DPR-24 & DPR-27, respectively.
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s Fir TABLE OF CONTENTS l
,n Page 1 0 ' INTRODUCTION 1;
2.0 HIGHLIGHTS 2.1'
. Unit l' 1
2 '. 2 L Unit 2 :
1 3.0 ~ FACILITY CHANGES, TESTS & EXPERIMENTS 3.1 Amendments to Facility Operating Licenses
'2' 3.2 Facility or' Procedure Changes' Requiring NRC Apprctali
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' 3.3 Tests or Experiments Requiring NRC Approval L43 3.4
. Designi. Changes -
45-3.5, Temporary Modifications 80 3.6 Core Reloads-106-
'4.0. NUMBER OF PERSONNEL & PERSON-REM BY WORK GROUP AND JOB-FUNCTION
.108 5.Q' STEAM GENERATOR TUBE INSERVICE INSPECTION' 5.1
- Unit 1 109 5.2 Unit 2 110 6.0 REACTOR COOLANT SYSTEM RELIEF VALVE CHALLENGES 6.1-Overpressure Protection During Normal Pressure &
Temperature. Operation 122-6.2 Overpressure Protection During Low Pressure &_
Temperature Operation 122 7.0 REACTOR COOLANT ACTIVITY ANALYSIS 122
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1.0 INTRODUCTION
The Point' Beach Nuclear Plant, Units 1 and 2, utilize identical pressurized water reactors rated at 1518 MWt each. Each turbine-generator is capable of producing 497 MWe net (524 MWe gross) of electrical power. The plant is located ten miles north of Two Rivers, Wisconsin', on the west shore of Lake Michigan.
2.0 HIGHLIGHTS 2.1 Unit 1 Highlights for the period January 1, 1988, through December 31,
".38, included a 41-day refueling / maintenance outage.
, construction of.a five story service building was completed during the period. The building provides a new storeroom, office area for the training, instrument and control, chemistry and technical services groups; as well as a new makeup water treatment plant.
Unit 1 operated at an average capacity factor of 88.9 percent and a net electrical / thermal efficiency of 32.7 percent. The unit and reactor availability were 88.7 percent and 89.4 percent, respectively. Unit 1 generated its 56 billionth kilowatt hour on January 16, 1988; its 57 billionth kilowatt hour on April 4, 1988; its 58 billionth kilowatt hour on August 6, 1988; and its 59 billionth kilowatt hour on October 26, 1988.
2.2 Unit 2 Highlights for the period January 1, 1988, through December 31, 1988, included a 45-day refueling / maintenance outage, a plant trip caused by the inadvertent parallel operation of instrument bus power supplies and a scheduled 39-hour maintenance outage to repair a stem-disk separation on a main feed pump discharge valve.
Unit 2 operated at an anerage capacity factor of 87.3 percent and a net electrical / thermal efficiency of 32.7 percent. The l
unit and reactor availability were 86.9 percent and 87.7 percent, respectively. Unit 2 generated its 56 billionth kilowatt hour on January 18, 1988; its 57 billionth kilowatt hour on April 9, 1988; its 58 billionth kilowatt hour on July 1, 1988; and itt 59 billionth kilowatt hour on September 19, 1988.
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3.0' FACILITY CHANGES, TESTS, AND EXPERIMENTS 3.1:
Amendments to Facility Operating Licenses t-
.During the~ year 1988, there were six license amendments issued i
by.the U.S. Nuclear Regulatory Commission to Facility Operating License DPR-24 for Point Beach Nuclear Plant Unit 1 and six license amendments ' issued to Facility Operating License DPR-27 j
for Point Beach Nuclear Plant Unit 2. 'These. license amendments.
y are listed by date of issuance and are summarized as follows-3.1 ' 1-01-05-88, Amendment 110 to DPR-24,
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Amendment:113 to DPR-27
- These snendments modify Technical Specification 15.6.10, " Plant Operating Records,";and modify other Technical Specifications-to correct minor administrative errors.
3.1.2 02-03-88,-Amendment 111 to DPR-24, Amendment.114 to DPR-27 These amendments delete Technical Specification 15.5.3.A.8, which specifies a limiting quantity of fissionable material in the form of fabricated neutron flux detectors.
3.1.3 03-02-88, Amendment 112 to DPR-24, Amendment'115 to DPR-27 These amendments (1) change the number of channels indicated in Technical Specification Table 15.3.5-5, Item 10, " Containment Hydrogen Monitors," from four to two; (2) modify Technical Specification Table 15.3.5-2,
" Instrument Operation Conditions for Reactor Trip," to accurately indicate the number of channels required to trap; and (3) change the term "zero power physics testing" to " low power physics testing" in a footnote toi j
Technical Specification Table 15.3.5-2.
r 3.1.4 04-14-85, Amendment 113 to DPR-24,
'Amendmen'. 116 to DPR-27 These' amendments revise the asterisked notes to Technical Specification Table 15.4.1-1, Item 25, "Stean Generator Pressure," which clarify the requirements for testing of the steam generator pressure channels during refueling shutdowns.
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04-18-88,~ - Amendment 114'to DPR-24,.
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Amendment 117 to DPR-27
- These amendments make numerous administrative-changes
' toithe Technical Specifications,' clarifying or-correcting several items in the Technical--
Specifications.
3.1 6 06-09-88, Amendment 115 to DPR-24,
. Amendment 118 to DPR-27 These amendments modify paragraph-3.Fiof-the licenses:
to; require compliance-with-the amended Physical; Security ~ Plan. This plan was' amended to conform to the
- retirements of 10 CFR 73.55.
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a 13. 2 Procedure Changes
'here.were~no procedure changes.made during 1988 beyond those-T
' authorized with license' amendments as noted.above, which required
' Nuclear Regulatory Commission approval. The following procedure changes made at Point Beach Nuclear Plant during 1988, required a 10 CFR 50.59 review:
3.2.1 CLORT #3 Unit 1: Safety Injection With Loss of AC Checklist.
' Summary of Safety Evaluation: A safety evaluation was required because the.above. procedure revision constituted -
a change to the procedures as. described in the FSAR.
SI pump breakers are'placed and' maintained in the a.
test position for ORT #3.
Technical Specification 15.4.5.1.A.1.a specifies this configuration. FSAR Section 6.2.4 for system testing states, "A test signal is applied to initiate automatic action pnd verification made that SI pumps attain required discharge heads. The test demonstrates the operation of valves, pump circuit breakers and automatic circuitry."
For operation. sequence testing, the fourth' paragraph.
of this section states in the last. sentence that, "The latter cases are performed without delivery.of water to the RCS, but include. starting of all pumping equipment involved in each test."
ORT #3, as changed, does not. singularly meet the FSAR test requirements; however, the overall combination of testing (ORT #3, ITs and ORT flow testing)' demonstrates high head capabilities (see alse Item c below).
b.
Reactor contadtrcit fan cooling 'mits an tested with the fan motor breakers in test. This prevents the electrical leading trom' approximating that in FSAR Section 8.2 (see Item c below).
Technical Specification Section 5.4.6, regarding c.
diesel generator electrical load testing, requires,
...will start and assume required load in FSAR Section E.2..."
Section 6.2 not only describes sequencing loads but also contains kW load limits.
By not actually operating equipment (e.g., high head SI pumps and fan cooler units), the load of 8.2 is not reached during ORT #3, although the sequence and timing is checked, some component loads are not sequenced.
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- Upon' recovery from the annual diesel generator
outage, the diesel generator is loaded beyond the kW requirements of the FSAR,- verified by receipt of the overload ~ alarm. However, this check is done via manual loading, not ESF sequencing. Additionally, individual ESF component electrical. supplies
'(breakers, for example) are not tested. The'FSAR is not' explicit if this is part of Section 8 or not.
Breakers are multi-amp tested' individually, providing some assurance of load carrying capability.
- d. -At various times throughout ORT #3, the procedure takes decay heat removal for the unit out of service.
This occurs due to testing of RHR pumps, RHR pump
- primary and redundant power sources, component cooling pumps and service water pumps. Technical Specification 15.3.1.A.3.b(4) allows one of two RHR-loops to'be temporarily out of service to meet surveillance requirements. In.all testing, the, time.
7 components are not in service is minimized and~
personnel are aware that decay heat removal equipment is'affected. Finally,-at this point in refueling, decay heat remaining in the fuel is-
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' minimal.
e.
The FSAR sections which were reviewed do not^ address electrical shedding of charging pumps on the combination of undervoltage plus SI.
This feature is tested during ORT #3.
Other equipment having.
such logic (th, motor-driven auxiliary feedwater pumps) is specifically identified Thus, the basis of analysis.with respect to electrical capabilities may not consider retention of charging pumps as electrical loads under emergency conditions.
Moreover, the lead'from the charging pumps is small relative to total emergency diesel generator capability, and procecural guidance e::ists directing -
operators to check diesel generator loading.
H The change does not pose an unreviewed se.fety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Techniral Specifications is not reduced.
3.2.2 CSP-C.2:
Response to Degraded Core Cuoling.
Summary of Safety Fvaluation: An evaluation is required 1
because this revision changes procedures described in the FSAR. Steps 6b and 8b were revised to be high level
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steps to conform to standard EOP format. The changes do not affect the manner in which plant equipment is operated.
'The changes do not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The changes do not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.3 CSP-H.1:
Response to Loss of Secondary Heat Sink.
Summary of Safety Injection: An evaluation is required because this revision changes procedures described in the FSAR. Step 9 was moved from one page to the next in order to keep the entire step on one page. Steps 3,'7 and 10 were changed to be high level steps to conform to the standard practice of identifying E0P transitions as high level steps. The changes do not affect the manner in which plant equipment operates or is operated.
The changes do not pose an unreviewed safety question.
The probability of occurrence or the ccasequences of an accident or realfunction of equipment important to safety i
is not increased. The changes do not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
l' 3.2.4 CSP-I.3:
Response to voids in the Reactor Vessel.
Summary of Safety Evaluatioy An evaluation is required because this revision changes procedures described in i
the FSAR. Symptoms and entry conditions contained in E0P-1 were added for reference purposes. The changes do not affect the manner in which plant equipment is operated.
The changes do not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The changes do not create the j
possibility for an accident or malfunction which has not I
been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.5 ECA-0.0:
Loss of all AC Power.
Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR.
In Appendix C, an extra line space was removed to correct a typographical error.
In Step 5, a l
line space was added to conform to standard format.
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'b Circles were~added around the first four step numbers to
~F correct clerical errors. Steps 3 and 10 were changed.to' l
realign words and add line spaces to maintain..the l
two-column format. Appendix A was, reworded to more j
clearly. define the steps needing to be performed and to-conform to standard format. The changes do not affect the: manner in which' plant' equipment is operated.
l The changes do not pose an'unrevived safety question.
The probability of occurrence or the consequences of an-accident'or malfunction of equipment important to safety is not increased...The changes'do not create the-possibility for an accident or malfunction which has.not been previously evaluated. The margin of safety as.
defined in.the. Technical Specifications is not reduced.
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1 3.2.6-ECA-0.1:
Loss.of All AC Power Recovery Without Safety.
Injection Required.
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Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR'.
Step 3 was revised to correct a grammatical error. Step 13 was revised to maintain 2-column format.
Neither of the changes affects the manner in which plant equipment is operated.
The changes ~do not pose an unreviewed safety question'.
The probability of occurrence or the consequences of an accident or malfunction of equipment important-to safety l
is not increased. The changes do not create the possibility for an accident-or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
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-3.2.7 ECA-1.1:
Loss of Containment Sump Recirculation.
Summary of Safety Evaluation: An evaluation is required because this revision changes' procedures described in the FSAR. Figure.1 and Steps 10a and 10b were changed to allow the use of charging pumps at low flow rates where the SI pump can 'only go 100 gpm. A charging pump is morr. easily controlled at lower flow rates.
This change gives the possibility of either/or pumps using the same type of water, giving the operator more j
flexibility. The change is also contained in Revision 1A of the ERG-LP where the NSSS vendor has 1
already approved the use of this method.
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The change does not pose an unreviewed safety question.
The~ probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the l
possibility for an accident or malfunction which has not-been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
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3.2.8 ECA-3.1:
Steam Generator Tube Rupture with Loss of Reactor Coolant Subcooled Recovery Desired.
Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR. Step 10 was revised to correct a typographical error.
Step 29 was revised to provide direction to complete-Subset e, which was inadvertently omi:ted.
In Step 29a, a line space was removed to correct a clerical error. The.
changes do not affect the manner in which plant equ pment is operated.
The changes do not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The changes do not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.9 E0P-0
Reactor Trip or Safety Injection.
Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR. The change to Step 18 constitutes correction of two typographical errors.
Step 27 adds the word
" valves" in the RNO column for clarification. Step 28 depositions a word for consistency.
In Step 37, a line space was removed in accordance with standard practice in the EOPs. These changes do not affect the manner in which plant equipment is operated.
The changes do not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The changes do not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.10 EOP-0.1:
Reactor Trip Response.
Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR. Step 8 was revised to remove a line space which was a typographical error.
Step 3 was revised to add an equivalency to the boric acid storage tank.
Neither of these changes affects the manner in which plant equipment is operated.
The changes do not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The changes do not create the 8
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h possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.11 EOP-0.2:
Natural Circulation Cooldown.
Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR. The symptom and entry conditions step number was revised to reflect changes in other procedures. The change does not affect the manner in which plant equipment is operated.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.12 EOP-1: Loss of Reactor or Secondary Coolant.
Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR. The reference number for the symptoms and entry conditions was revised as a result of changes to other procedures.
In Step 14, a typographical error was corrected.
In Step 25, the sentence was reworded to reflect standard practice for transitional steps. None of the changes affects the manner in which plant equipment is operated.
The changes do not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The changes do not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.13 EOP-1.1:
Safety Injection Termination.
Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR. The reference numbers for the symptom and entry conditions were revised to reflect changes made to other procedures. Steps 5, 6 and 11 were corrected to add / remove line spaces to conform to standard E0P formatting practice. These changes do not affect the manner in which plant equipment is operated.
The changes do not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety 9
is not increased. The changes do not. create the i*"
7ossibility for an accident or malfunction which has not been previously. evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.14 E0P-1.3:
Transfer to Containment Sump Recirculation.
Summary of Safety Evaluation:
An' evaluation is required because.this revision changes procedures described in the FSAR. Step ~3d was added to ensure' adequate-flow to i.
the accident fan cooler with the ability to still provide more than adequate flow to the. component cooling heat exchanger for this type of accident. The change does not affect the manner in which plant equipment is operated.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the
. possibility for an accident or malfunction-which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.15 EOP-3:
Steam Generator Tube Rupt ure.
. Summary of Safety Evaluation: An evaluation is required because this revision changes procedures described in the FSAR. Step 16 (RNO) was revised to change the word
" required" to " target temperature" to more clearly identify the required means of operation. The change does not affect the manner in which plant equipment is operated.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased..The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.16 FTI-PBNP-01: Pipe Freeze.
Summary of Safety Evaluation: Use of the freeze sealing process constitutes a potential change to the facility or its operation as described in the FSAR. This evaluation documents a generic review of this process.
The review does not constitute a safety evaluation for the individual uses of this process.
Although each freeze plugging instance will require an individual engineering review, the following is intended to justify the generic usability of the process of 10
, t.m tfreeze plugging. The review presents the process of freeze plugging and references tests'that have been done to analyze freeze plugging effects on piping integrity.
The freeze plug process 'uses an -insulated jacket placed around a pipe into which liquid nitrogen'is introduced-to'obtain temperatures of -320 F.
Jul extensive series-of tests on-piping material. properties and freeze plug integrity was done by an analysis laboratory as documented in a report dated November 15, 1982.~
That report supports this evaluation. The tests were done on 12-3/4" OD API. Grade 5LX42 and SLX52:and 24" OD API Grade 5LX65 pipe. The SLX42 has very.-similar properties as A53 and A106 Grade B.
This is supported by the Battelle Laboratories report.
It was found that yield strength and tensile strength-increase as temperature decreases; as much as two times the. room temperature values. 'Obviously,-however,,the fracture toughness decreases with decreasing, temperature and the point will be reached and surpassed at which crack propagation is not depressed by ductility.
since the ductility has decreased with decreased L
temperature.
Therefore, to ensure that the -initiation of propagating t
cracks does not occur,'it must be verified before the freeze plugging takes place that there are no cracks detectable via an NDE inspection, although the Battelle l
report provides a means for analyzing whether a crack / indication (if found) is acceptable for the operation (if necessary).
The same NDE inspection shall be done after the freeze plug is removed to verify any existing indications have
-not grown or new ones started.
It should also be noted that the mechanical properties of piping which has been subjected to low temperatures return to normal when temperature is back to normal. A laboratory report of July 14, 1976, and a letter of transmittal dated May 13, 1977, document'this issue.
The tests that have been done on the 12-3/4" OD samples by Battelle included a freeze plug test where the water in the pipe was not pressurized. The circumferential stress found, indicated at each gauge location through the cooldown and heatup process, did not exceed 10,000 psi. Obviously, with increasing pressures, stresses increase.
The laboratory testing on the various freeze plugs included having a 2240 psi differential across a freeze plug, with no spillage occurring. The ice plug resistance to slippage is due to an ice-to-pipe bond and also the hourglass shape the pipe attains in the freeze 11
plug area at the low temperature. This'high' pressure'is typicali of those pressures seen during hydros.of piping systems. Ice plugging operations at PBNP shall be limited to static and low pressure situations.
Precautions must be taken to ensure.that any trapped areas of water between the ice plug and any isolated
- equipment is monitored for pressure increases during the freezing operation. This'shall be part of the work plan-and is also addressed in'the vendor. Studies have shown that pipe' ruptures due to freezing occur in areas where the liquid phase still-exists and becomes excessively pressurized by the expanding ice adjacent to it.
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Conclusions:==
The information available indicates that the process is usable if proper precautions, as.
mentioned above, are exercised. Therefore, it is felt that use of a freeze plug and the potential for. failure do. not pose significant problems.
A' listing of precautions, as a minimum, are contained in FTI-PBNP-01.
There are items that have to be considered in-an engineering review of the-ice plugging case. The piping yield and ultimate tensile strength is higher at.
the cold temperatures; thus, if the pipe is' verified to have no flaws, it is capable of higher loads. No permanent deformation or change in material properties occurs as a result of freeze sealing. Therefore, the piping will still meet its design. specification requirements after it is returned to normal operating temperature. 'If the freeze sealing process is to be used on stainless steel piping, the precautions listed must still apply.
The process does not pose an.unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The process does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.2.17 Furmanite Procedure N-88073: Unit 2 B Steam Generator Outboard Secondary Manway Leak Repair.
Summary of Safety Evaluation: The repair constitutes a potential change to the facility or its operation as described in the FSAR. The FSAR states that the steam generator was analyzed in accordance with the rules of the ASME Boiler & Pressure Vessel Code for Nuclear Vessels,Section III, 1968 Edition. Also, a fatigue i
analysis was performed at potentially critical regions.
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.The3 stress report-for:the steam generator was reviewed and1 calculations performed by the leak seal' vendor and reviewed by WE to verify that theLinjection sealant
-process.would not result in stresses exceeding-Section III allowables.
This area 'the secondary manway cover) was not
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identified in the~ original stress report as an area of concern for cyclic stresses; thus a fatigue analysis was not done, and fatigue was not considered here.
Injection sealant'can be. broken into two aspects.
First, the actual process and the change in the cover sealing configuration; and se'cond, the holes,that must be drilled in the cover to inject the sealant. The.
-process itself will be controlled by an approved procedure and the cover will be restored to the original configuration during the next refueling outage. The holes, which already exist-in this cover, will remain.
.The existence of these holes in closure covers has been addressed before and determined acceptable for the
- pressurizer cover (MWR #84231) and the Unit l'B' steam.
generator secondary manway cover (MWR #52313).
Based upon the analysis performed and previous justification and acceptance, the stress report-conclusions remain valid and were still within Code allowable stresses as stated in the FSAR.
The repair does not~ pose an unreviewed safety question.
The. probability of occurrence or the conse'quences of an accident or malfunction of equipment important to. safety is not increased. The repair.does not create the' possibility for an accident or malfunction which has not-been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.18 Furmanite Procedure N-88076: -Unit 2 B Steam Generator.
Outboard Secondary Manway Cover.
Summary of Safety Evaluation: Reinjection of the steam generator manway cover was performed in the same manner.
as the original injection. Thus, aspects of the analysis performed and attached to MWR #880830 apply with respect' to the original safety evaluation, the leak seal vendor calculation forming part of the leak seal vendor procedure N-88073, and PBNP calculation P-88-006, which addressed manway cover stresses as needed.
The FBNP calculations concluded that the stresses incurred as a result of the Furmanite sealing of the manway cover are below Code allowable. Part of the control over the first injection process was to limit injection pressure to 3000 psi, which is again the limit per the reinjection procedure.
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1Since.previously; injected sealant will-not be as fluid-
-as it:was'when first injected, injection pressure A
. greater than' 3000 psi' mayfbe needed,to initiate flow.
into cavities. A lower injection pressure' of-3000 psi.
should then be~ sufficient ~for. completing injection of sealant and. stopping.the'~1eak.
l' The-change 'does not poselan unreviewed ' safety question.
The probability of occurrence or the consequences of.an accident orfmalfunction of equipment important:to safety is not increased. The change does not-create the possibility for~an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced; 3.2.19-Furmanite Procedure N-88216: On-Line Sealing of.
2MS-2018A Stuffing Box Flange' Joint.
Summary of Safety Evaluation: The evolutions constituted a potential change to the facility or its operation as described in the FSAR. Leak repairing of the stuffing box' flange area on a nonreturn valve l involved drilling L
holes ~in the stuffing box' flange'and injecting sealant 4
via injection adapters. Calculation P-88-023. addressed
. bolt stresses as a result.of changing the gasket design to a full-face gasket and thus'. increasing the stud-
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loading. Stresses at system design pressure were acceptable. Stresses seen during sealant injection were above Code allowable (25,000 psi), but.are not considered above those values which could be seen during stud torquing.
Sealant injection did not impair the disc / arm stroking l-capability because the Furmanite' sealant did not touch j
the valve shaft.
L Furmanite sealant of the'nonreturn. valve stuffing box flange was considered an acceptable fix. Failure of a steam line outside of containment was considered in the FSAR alcng with high energy line breaks.
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The evolution does not pose an unreviewed safety question. The probabilities of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
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3.2.20 HP 9.4:
Approved Temporary Lead Shielding Installation.
Summary of Safety Evaluation: The revision added Step 10.4 to install temporary lead shielding on the A and B steam generator platforms during an outage condition.
There were administrative controls in place to prevent lead shielding from being left in containment; thus the shielding was removed prior to unit startup. However, in the unlikely event that shielding is left in containment, the added weight would not affect the steam generator or reactor coolant pump supports.
It would affect the platform.
If the lead shielding fell from the platform, it would not move and therefore, does not present a concern for blocking or diminishing flow to the residual heat removal systeA during the recirculation mode. The failure of the platform resulting from leaving the lead shielding in place would likewise not present a problem for RHR operation, and since therp is no safety-related equipment located in the vicinity under the platform, the failure, if it occurred, would not affect safety-related equipment.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.21 ICP 2.19 Appendix A.
I&C Surveillance Test, Safeguards System Logic.
Summary of Safety Evaluation: An evaluation is required because performance of this procedure constitutes a potential change to the facility or its operation as described in the FSAR.
Technical Specification Table 15.3.5-4 Items 2.a and 2.b state that the minimum number of operable channels for steam flow and Tavg is one per loop. Also, Section 14.2.5 of the FSAR addresses steam line isolation as a result of a steam line break.
The performance of ICP 2.19 Appendix A could conceivably reduce the number of operable steam flow and Tavg channels to zero in either or both loops, while maintaining the high and high-high steam flow and low Tavg bistable test switches in their normal (untripped) j position. This would violate the minimum number of
~
operable steam flow and Tavg channels required by 15
f Technical Specifications. Also, the automatic steam line isolation feature will be defeated in both trains of safeguards. Steam line isolation is addressed in the steam line rupture analysis in Section 14.2.5 of the FSAR.
ICP 2.19 Appendix A will only be performed to satisfy post-refueling, pre-startup testing requirements prior to criticality. However, it is possible for.the plant to be in or approaching hot shutdown status at the time ICP 2.19 Appendix'A is performed. This would place the plant in a Limiting Condition for Operation and require either returning the steam flow and Tavg channels to normal within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or proceeding to a cold shutdown condition.
To ensure Technical Specification adherence and steam line isolation requirements are satisfied, ICP 2.19 Appendix A incorporates placing the steam line isolation valves in the closed position during the performance of the test. A caution has been added to ICP 2.19 Appendix A alerting the DSS and the I&C technician to the possible LCO.
The date and time prior to starting and upon completion are recorded in the procedure to document Technical Specification compliance.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.22 ICP 11.402: Emergency Fuel Oil Tank Level Switch Calibration Procedure.
Summary of Safety Evaluation: This procedure constttues a change to an NRC commitment.
In the WE response to 10 CFR 50 Appendix R, Step 4.5.2, a commitment was made to modify the fuel oil system to gravity feed fuel oil to the emergency diesel day tanks from the fuel oil storage tanks. This commitment was satisfied by the installation of a gravity feed line to the emergency diesel day tanks via the emergency fuel oil tank overflow (reference MR 83-150). As a result of this commitment, emergency fuel oil tank integrity must be maintained.
The calibration and adjustment of the emergency fuel oil tank level switch requires that the level switch be removed from the tank. The opening created by the removal of the level switch will compromise the commitment to gravity feed fuel oil to the emergency diesel day tanks.
16
i T
The procedure,;ICP 11.402, to remove and calibrate' thel level switch requires that a gasket and blank. flange'be
~
temporarily installed in place of the-level switch to maintain ~ tank integrity while the switch is not in
,a
-place. 'The~ calibration procedure also contains a precaution alerting the DSS'and the procedure user about maintaining emergency fuel oil tank integrity.
However, there was approximately a.40-minute time interval during switch removal and reinstallation that the integrity of the emergency fuel tank was compromised.
The fuel oil gravity feed system was out of service during these times.
The calibration procedure requires that sufficient' fuel oil will be available to.the emergency diesels from the diesel day tanks and the diesel sump tanks while the.
gravity feed system:is out of service. Both diesels.
have approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of fuel available from their-respective day tank and sump tank.
The calibration procedure also-incorporates a precaution which requires that the fuel' oil gravity feed system be returned to operable status.if either emergency diesel is' required to operate. The maximum time required to return the gravity feed system to operable status was approximately 40 minutes; thus emergency diesel operation was not affected by the performance of this procedure. As an added precaution, the procedure required that both fuel transfer pumps be operable.
The procedure does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The procedure does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.2.23 OP-10E: Discharge of B CVCS HUT.
Summary of Safety Evaluation: 'Use of this procedure constitutes a change to the facility as described in the FSAR. The purpose of the procedure was to align and discharge the B CVCS holdup tank via radiation monitor RE-218.
The margin of safety was not reduced by the temporary modification placed by OP-10E or the conduct of the effluent release administratively controlled by OP-10E.
17
s.
l i
f FSAR safety analysis,14.2.2 demonstrates that release' limits:would not be violated upon the. rupture of'a full
~ holdup tank. The complete volume can be contained L.
within the auxiliary building. Thus, anyl leakage, including a rupture of the temporary hose, falls within the bounds of this' analysis.
Analysis 14.2.2 also describes the' procedure'to'be use'd to discharge liquid waste.
Included.are isolation'of the tank to be released; a sample of contents.taken and' analyzed; assessment of the quantity of activity to be-released versus permitted limits; and the use of'an automatic stop valve which shuts upon a high radiation signal and which fails shut. OP-10E addresses all of-these-concerns. For example, a discharge permit-is required and valve lineups are made to isolate the B holdup tank.
The use of installed radiation monitor RE-218, along with its automatic features, ensures lthe monitoring requirements of Section.14.2.2 are met.
Further, RETS 15.7.5.G.c allows.the pre-planned batch releases of unprocessed effluents so long as the criteria.specified in Section 15.7.5.G.c.1 through 4 are met.
In this case, the criteria were met or did not
-apply. Procedure OP-10E contains the administrative controls to accomplish the effluent release.
The change does not pose'an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change'does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.24 ORT #3 Unit 1:
Safety Injection Actuation With Loss of Engineered Safeguards AC.
Summary of Safety Evaluation: A safety evaluation was required because the above procedure revision constituted a change to the procedure'as described in the FSAR.
a.
SI pump breakers are placed and maintained in the test position for ORT #3.
Technical Specification 15.4.5.1.A.1.a specifies this configuration.
FSAR Section 6.2.4 for system testing states, "A test signal is applied to initiate automatic action and verification made that SI pumps attain required dis,.harge heads. The tes*. demonstrates the j
operation of valves, pump circuit breakers and j
automatic circuitry."
18
_-.___________________________J
- o For operation. sequence testing,_the fourth paragraph ofEthis section states in the last sentence that,;
"The.latter cases are performed without delivery of water to the RCS, but include starting of.all
. pumping equipment involved in each test."
ORT #3, as changed, does not singularly. meet the_
FSAR test requirements;'however,-the overall-combination'of testing (ORT #3, ITs and ORT flow.
testing). demonstrates high head capabilities _(see also Item c below).
b.
Reactor. containment fan cooling units are tested with the-fan breakers in test. This prevents the electrical loading from approximating that in FSAR Section 8.2 (see Item c below).
c.
Tedhnical Specification Section 5.4.6, regarding diesel generator electrical load testing, requires,
...will start and assume required load in FSAR Section-8.2..."
Section 8.2 not only describes
- sequencing loads but also contains kW load limits.
By not actually operating equipmentc (e.g., high head SI pumps and fan cooler units), the_ load of 8.2 is not reached during ORT #3, although the sequence and timing is checked, some component loads'are not sequenced.
Upon recovery from the annual diesel generator outage, the diesel generator is loaded beyond the 163 requirements of the FSAR, verified by receipt of the overload alarm. However,'this check is done via manual loading, not ESF sequencing. Additionally, individual ESF component electrical supplies (breakers, for example) are not tested. The FSAR is not explicit if this is'part of Section 8 or not.
Breakers are multi-amp' tested individually, providing some assurance of load carrying capability, d.
At various times throughout ORT #3, the procedure takes decay heat removal for the unit out of service.
This occurs due to testing of RHR pumps, RHR pump primary and redundant power sources, component cooling pumps and service water pumps. Technical Specification 15.3.1.A.3.b(4) allows one of two RHR loops to be temporarily out of service to meet surveillance requirements.
In all testing, the time components are not in service is minimized and personnel are aware that decay heat removal equipment is affected. Finally, at this point in refueling, decay heat remaining in the fuel is minimal.
19
r, q.
.i e.
'The FSAR sections which were. reviewed do"not address-g electrical shedding of charging pumps on the,
combination of undervoltage plus SI.
This feature is tested during ORT,#3. ;Other equipment having such logic.(the motor-driven auxiliary _feedwater pumps) is specifically identified. Thus, the basis.
of. analysis with respect to electrical capabilities.
may not consider retention of charging pumps as electrical loads under emergency. conditions.
Moreover, the load from~the_ charging pumps is small; l
relative to total emergency diesel generator l
i capability, and procedural guidance _ exists directing.
operators to check diesel generator loading.
'The change does not pose an unreviewed safety question.
The probability of occurrence'or the consequences of an-accident or. malfunction of equipment ~important to safety is not increased. The change does not create the j
[L possibility for an accident or malfunction which has not been previously evaluated. The margin of: safety as defined in the Technical Specifications is not reduced.
j 3.2.25 ORT #3 Unit 2: _ Safety Injection Actuation with Loss of Engineered Safeguards AC.
i Summary of' Safety Evaluation: The addition of Sections 5.7--
and 5.8 represented a potential change to the facility or its operation as described in.the FSAR. These i
sections test the A train and B train 480 V and 4160 V-
-safeguards logic, respectively. The purpose of these l
sections is to repeat the loss of AC and SI portions of this test and isolate the operation'of the AC undervoltage relays. The steps being performed are similar to.
Section 5.4 with a slower time sequence such that relay and component operation can be verified.
This change was being evaluated because the test is described in Sections 15.4.5.I.A.1'and 15.4'6.A.2 of the Technical Specifications.
Initial conditions of ORT #3 I
state that all shared systems and components are operable and neither system is in a Limiting Condition
]
for Operation. During this portion of the test,.some safeguards equipment is removed from service. The removal of this equipment is allowed by the Technical Specifications. No limiting condition or other degraded mode of operation will result.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change-does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
20
l-3.2.26 RMP.#83:
ISI-870A Valve Operator Maintenance.
-Summary of Safety Evaluation: This evolution constituted a change.to the facility as described in the FSAR.
Valve 1SI-870A is the suction isolation from the RWST for the A spray pump. This valve is open during normal operation. When Unit 1 is in a cold shutdown, refueling outage or IP14A is out of service, ISI-870A can be closed. Via RMP #83, the valve was clamped in the closed position so the valve operator could be removed for maintenance.
The clamp was designed to not damage the stem and to provide adequate force to keep the valve closed at system design pressures. The valve was only clamped shut during cold shutdown or refueling when operability.
of the spray system was not required. The fact that the valve was clamped closed did not pose a problem for the RWST. The seismic qualification of the system was not be adversely affected since the removal of the operator (200 lbs) and installation of the clamp (5 lbs) did reduce the system weight and the stresses on the system.
I Additional seismic concerns were the possible effects of loading on the clamp and the unsupported valve stem with the motor operator removed. Since the valve is shut and mounted vertically, the clamp can only be loaded due to vertical down acceleration. Since the weight of the stem and disc was not subtracted out of the clamp force calculation, the vertical down "g" load would have to be greater than 1 g.
Our SSE loadings are not that high, assuming the piping was designed as required with a natural frequency higher than 30 Hz.
Removal of the motor operator weight would increase the natural frequency.
The stem is unsupported in the horizontal direction above the gland with the motor operator off. By engineering judgment, the approximately 12" cantilevered, 1-1/8" diameter stem would not be a problem at our "g" loads.
Because the unit was in the cold shutdown or refueling shutdown condition, the spray system and RWST are not required for nuclear safety of the plant. The clamp was installed to keep the valve shut so it could be used as an isolation for the RWST for working on the out-of-service spray system. Maintaining the RWST as a possible source of boric acid for injection to meet Technical Specification 15.3.2.A was desirable, however. Since the pressure boundary was maintained and seismic capability was maintained, the RWST remained operable as a boric acid source.
21
i S
This. valve is shown as'an MOV in the P&ID figure:in-the FSAR..Thus, removal of the-motor operator.was considered
-as a change:to-the. facility as described in the FSAR.
j It'was recognized tha't'the clamp did not exert a force j
~
on the valve stem to push the wedge.into the seat..
This
]
fact was not considered to be a problem with respect'to j
maintaining a leak-tight shutoff. The reason for.this i
was'because the valve could have been shut with the motor operator' prior to installing the clamp'and removing 1 the motor operator. Thus, the valve disc would have l
been wedged into the' seat with.the normal closure force.
The valve stem must-have some travel prior:to~ starting.
Lto move the disc. Also, from MOV testing, it has been noted that a' considerable force is required to pull thei disc out of the seat. Based on this, it was expected that the disc would remain wedged in the' seat and would have the same leak-tightness with the clamp installed.as it had with the motor operator installed; The change does not pose.an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.
is not increased.- The change'does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in'the Technical Specifications is not' reduced.
3.2.27 RMP #84:
151-870B Valve Operator. Maintenance.
Summary of Safety Evaluation: This evolution constituted a change to the facility as described'in the FSAR.
Valve ISI-870B is the suction isolation from the RWST-for the B spray pump. This valve is open during normal operation. When Unit 1 is in a cold shutdown, refueling outage or IP14B is.outLof service,'1SI-870B can be closed. Via RMP #84, the valve was clamped in the closed position so the valve operator can be removed for maintenance.
The clamp was designed to not damage the stem and to provide adequate force to keep the valve closed at system design pressures. The valve was only clamped shut during cold shutdown or refueling when operability of the spray system was not required. The fact that the valve was clamped closed did not pose a problem for the RWST. The seismic qualification of the system was not adversely affected since the removal of the operator (200 los) and installation of the clamp (5 lbs) did reduce the system weight and the stresses on the. system.
22
1 a
U l
Additional' seismic concerns: Were the possible effects'of' loading on the clamp and'the unsupported valve stem with<
~
the motor operator removed. Since.the valve is-shut and mounted vertically, the: clamp can only be loaded due^to-vertical down acceleration.
Since the weight of th'e
'l stem and disc was not subtracted out of the clamp force calculation, the vertical down "g" load would have to be
' greater than 1 g.
Our SSE loadings.are not that-high,
~
assuming the piping was designed as required with a natural frequency higher than-30 Hz.
Removal of the-
. motor operator weight would increase the natura1' frequency; The stem is unsupported in.the horizontal direction i
above the gland with the motor operator.off.~ By engineering judgment, the approximately 12" cantilevered, 1-1/8" diameter stem would not be a problem at our "g"
loads.
Because the unit was in the cold shutdown or refueling shutdown condition, the spray system and RWST are not required for nuclear safety of.the plant. The clamp was installed to keep the valve shut'so it could be used as an isolation for the RWST for working on the out-of-service spray system. Maintaining the RWST as a possible source of boric acid for injection to meet Technical-Specification 15.3.2.A'was desirable, however. Since the pressure boundary was maintained and seismic capability was maintained, the RWST remained operable as a boric acid source.
L This valve is shown as an.MOV in the P&ID figure in the FSAR. Thus, removal of the motor operator could be considered as a change to the facility as described in the FSAR.
It was recognized that the clamp did not exert a force on the valve stem to push the wedge into the seat. This fact was not considered to be a problem with respect to-maintaining a leak-tight shutoff. The reason for this was because the valve could have been shut with the motor operator prior to installing the clamp and removing the motor operator. Thus, the valve disc would have been wedged into the seat with the. normal closure force.
The valve stem must have some travel prior to starting to move the disc. Also, from MOV testing, it has been noted that a considerable' force is required to pull the disc out of the seat. Based on this, it was expected that the disc would remain wedged in the seat and would have the same leak-tightness with the clamp installed as it had with the motor operator installed.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the 23
~
l:
p w
possibility'forfan accident or malfunction which has not N
been previously evaluated. The margin of, safety: as -
defined in the Technical Specifications:is;not reduced.
l n.
'3.2.28k"RMP#85:
1SW-2880 Valve Maintenance, nr Summary of Safety Evaluation:.The' evolution constituted' a change to.the facility as described in the FSAR.
Valve 1SW-2880 is normally open during Uriit 1 operation.
This valve supplies service water to the Unit 1 turbine-hall. This valve would be required.to close when Unit 1
-safety injection is actuated and less-than 4/6 service water pumps are running within 30= seconds'. Unit 1 in
-cold shutdown or in a refueling outage are. prerequisites-to operator removal and the valve stem being clamped ~
open.
Since the safety injection system does notihave to be operable during a cold shutdown or refueling outage, this isolation feature does not have to be operable, The design of the stem clamp was adequate to-hold the valve open without damaging the stem. Manual isolation valves are available should the Unit 1 turbine building service water need to be. isolated for nonsafety-related.
reasons.
Clamping of this valve in the open position during cold shutdown or.a refueling outage did not present an unreviewed safety concern.. The valve, stem clamp weighs less than the valve operator, therefore, seismic stresses was reduced. The seismic capability of'the stem clamp was not addressed since failure.of.it has no nuclear safety consequences. By engineering judgment, seismic loading on the 1-3/6" diameter stem is not a problem.
The Technical Specifications' require that two service water loop headers be operable. This was' maintained with 1MOV-2880 clamped open since the. seismic capability of-the piping s maintained. The FSAR states that,
" Motor-operated isolation valves are provided which are controlled remotely from the control room and which automatically isolate nonessential services in the event of a safeguards actuation signal." The previous discussion addressed this factor, however,-the FSAR goes on to say, "The only portion of the system in a non-Class I structure is the piping in the turbine building to nonessential equipment. This portion of the system can be rapidly isolated by remotely-operated valves with redundant manual backup valves in the supply headers located in the Class I section of the control building."
24
t With the motor operator removed from ISW-2880 and the valve clamped open, the remote operation capability waa lost. The probability of a failure of the piping downstream of this valve was not increased, only the time to isolate it would take longer, since a manual isolation valve would have to have been used. The capability to remotely isolate the service water header still was available to the operators to ensure adequate flow to the Unit 2 safeguards systems if required.
Because the redundant manual isolction valves were operable and the capability to remotely separate the service water headers remained, it was felt that the consequences of a failure of the Unit I turbine hall service water supply piping was not increased.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not 3
been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.29 RMP #86:
ICV-112C Valve Operator Maintenance.
Summary of Safety Evaluation: Valve ICV-112C permits isolation of the VCT outlet to the charging pump suction header. The valve stem clamp was required to hold the valve shut during maintenance on the motor operator.
Maintaining this valve closed during refueling did not adversely affect the ability of the CVCS to perform its functions, i.e.,
chemical control, purification, etc.
The valve is not described nor are accident analysis assumptions made regarding its ability to be opened.
This valve is shown as an MOV in the P&ID figure in the FSAR. Thus, removal of the motor operator could be considered as a change to the facility as described in the FSAR.
The valve, as part of the CVCS, is Seismic Class I per FSAR Appendix A.
Since the clamp weight was less than the motor operator, any seismic stresses in the clamped configuration was less than.if the motor operator was installed. The possible effects'of seismic loading on the clamp and the unsupported valve stem with the motor operator off were considered.
Since the valve was shut and mounted vertically, the clamp could only be loaded due to vertical down acceleration. Since the weight of the stem and disc was not subtracted out of the clamp force calculation, the vertical down "g" load would have to be greater than I g.
Our SSE loadings are not that high, assuming the piping was designed as required with a natural frequency higher than 30 Hz.
25
i t
s Removal of the motor operator weight would increase the natural frequency. The stem was unsupported in the horizontal direction above the gland with the motor operator off. By engineering judgment, the approximately 12" cantilevered, 1" diameter stem was considered to be a problem at our 'ag" loads.
It is recognized that the clamp would not exert a force on the valve stem to push the wedge into the seat. This fact was not cor.sidered to be a problem with respect to maintaining a leak-tight shutoff. The reason for this is because the valve would have been shut with the motor operator prior to installing the clamp and removing'the motor operator. Thus, the valve disc would have been-wedged into the seat with the normal closure force. The valve stem must have some travel prior to starting to move the disc.
Also, from MOV testing, it has been noted that a considerable force is required to pull the disc out of the seat.. Based on this, it was expected that the disc would remain wedged in the seat and wpuld.
have the same leak-tightness with the clamp installed as it had with the motor operator installed.
[
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced, 3.2.30 RMP #87:
ICV-112B Valve Operator Maintenance.
Summary of Safety Evaluation: Valve ICV-112B isolates makeup to the CVCS from the RWST. The stem clamp is required for maintenance on the motor operator., The valve is not described nor are accident analysis assumptions made regarding its ability to be opened.
This valve is shown as an MOV in the P&ID figure in the FSAR. Thus, removal of the motor operator could be considered as a change to the facility as described in the FSAR.
Two paths of boration were available. With this valve closed, even if it leaked by, the inleakage would have been water from the RWST, which was maintained at 2000 ppm boron.
Since this is above the boron concentration requited for refueling, no inadvertent dilution can occur via this path. The valve, as part of the CVCS, is seismic per FSAR Appendix A.
Since the clamp weighed less than the motor operator, seismic stresses would be less than design during a seismic event.
26
n:
1 I;
L 1
1 II The possible' effects of seismic loading on the clamp and.
l the. unsupported valve stem with the motor operator off
~
C were considered. Since-the valve is shut and mounted l
vertically, the clamp can only be loaded-due to vertical L
_down acceleration.
Since the weight of the' stem'and disc was not subtracted out of the clamp force calculation, the vertical down "g" load would have to.be greater than'1.g.
Our SSE loadings are not.that high, assuming the piping was designed as required with a natural frequency higher than.30 Hz.
Removal of the.
motor operator weight would increase-the natural'
-frequency. The stem was unsupported in the horizontal direction above the gland with the motor operator.off.
By ' engineering -judgment, the approximately 12" cantilevered, 1" diameter stem was not considered to be; a problem at our "g" loads.
It is recognized that the clamp would not exert a force on the valve stem to push the. wedge into the seat. This l
fact was not considered to be a problem with respect to maintaining a' leak-tight shutoff. The reason for this-is because the valve would have been shut with the motor operator prior to installing the clamp and removing the motor operator; Thus, the valve disc would have been wedged into the seat with the normal closure force. The valve' stem must have some travel prior to starting to move the disc. Also, from MOV testing, it.has been noted that a considerable force is required to pull the disc out of the seat. Based on this, it was expected that the disc would remain wedged in the seat and will have the same' leak-tightness with the clamp installed as it had with the motor operator installed.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased.' The c!;ange does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.31 RMP #88:
ICV-350 Valve Operator Maintenance.
Summary of Safety Evaluation: Valve ICV-350 is kept in the closed position during normal operation. The valve could be required to open in order to provide a path for emergency boration to bring the reactor to a cold shutdown condition. The use of this valve would only be needed in the event of an abnormal at power accident situation.
?
27
~l i
The required condition for stem clamping is cold shutdown or a Unit I refueling outage. With the unit in a cold shutdown condition, emergency boration is not required. The valve was clamped shut so 12% boric acid was not leaked into non-heat traced piping which would allow the boric acid to crystallize. The stem clamp was adequately designed to maintain the valve. closed without damaging the stem. The weight of the stem clamp was less than the weight of the valve operator. Therefore, the seismic stresses were reduced.
The valve was clamped closed instead of isolating it so a path for boric acid injection from the boric acid storage tank remained available. Also, the boric.
t acid recirculating pump and its flow path remained in service. The pressure-retaining capability of the valve remained intact; thus, the piping on either side of the valve remained operable. This path was required to meet Technical Specification 15.3.2.A.
Leakage through the valve was not considered to be any l
more likely than with the motor operator on just prior to removing it.
The valve was shut using the motor operator, which wedged the disc into the seat. The clamp was installed prior to removing the operator; thus the established seal was not disturbed. Even though the clamp did not exert a closing force on the valve, the seal was expected to be maintained based upon the fact j
that the stem must undergo some movement before it would start to move the disc. MOV testing has shown that a j
considerable force is required to pull the disc out of the seat.
The clamp did not have to be seismic since failure of the clamp and opening of the valve had no nuclear safety consequences. By inspection, however, the clamp was able to handle our seismic loads since it would only be subjected to loads from a vertical down acceleration.
Assuming the piping was designed as required with a natural frequency above 30 Hz, the vertical "g" loads would not be enough to offset the weight of the disc and stem. The natural frequency would increase with the remo'.'al of the motor operator weight. Seismic loading on the approximately 10" cantilevered, 3/4" diameter stem was not considered a problem by engineering judgment.
With the unit in cold shutdown, removal of the motor did not affect any description in the FSAR. The P&ID figure in the FSAR, however, shows the valve as an MOV.
Thus substitution of the motor operator with a clamp was considered as a change to the facility as described in the FSAR.
28 j
The change does not pose an unreviewed safety question.
-The probability _of occurrence or the consequences of'an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.32 SLP 4: Auxiliary Building Main Crane Safe Load Path.
Summary of Safety Evaluation: The temporary procedure change constituted an alteration to a prior technical commitment to the NRC. The change modified the designated safe load path for the resin storage tank (T112) and cask. This change was required because the refueling bridge had been removed for maintenance and was blocking the normal path.
Safe load paths have been identified as a response to NUREG-0612. The drawing, SLP 4, contains both a designated path and shaded area which is designated as allowable.for moving heavy loads through. The change made was still within the allowable path area, therefore, no additional justification was required.
The new path allowed a lift directly over a 10" service water pipe leading to the spent fuel pool heat exchanger. This was acceptable because the auxiliary building crane has been designed to meet the " single failure-proof" criteria.
In addition, a rupture of this pipe would have only'affected spent fuel pool cooling and could be isolated through SW-650.
Since this system does not have a limitation on being out of service, isolation w/2s acceptable.
The Technical Specifications and FSAR Section 9.5 detail specific requirements for lifts over the spent fuel pool but did not apply to this lift.
The change does not pose an unreviewed safety question.
i The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.33 SMP #850: AF-4035 Repair.
Summary of Safety Evaluation: Performance of this special maintenance procedure constituted a potential change to the facility or its operation as described in the FSAR. During the removal and reinstallation of the relief valve, the mini-recire lines from all of the 29
t auxiliary feedwater pumps was' isolated. The mini-recirc line isLonly.for the protection of the pumps during a no-flow condition to prevent' overheating; it has no' safety-relatedffunction during an accident. On an automatic start of the auxiliary feedwater pumps,
~
flow to the steam' generators will exist and the recirc-line is not'needed.
Flow from the turbine-driven-auxiliary feedwater pumps will occur as soon as' pressure
. builds to greater than steam generator pressure;land for
.the. electric-driven pumps, as soon'as pressure builds'to greater than the pressure controlfvalve setpoint..'
Failure.of the pressure control valves to open could.
have resulted in damage to the pump, but'the consequences of this are no more than.the failure of the pressure control valve to open since it would have prevented flow from getting to-the steam ~ generator.
The time period that the mini-recire line was isolated was short and the special maintenance procedure contains a precaution to not operate the pump with less-than 39 gpm flow. Thus, the potential for damaging the auxiliary feedwater pumps was minimal.
Protectionof/thelowpressurepipingwhentherelief valve was removed was by administrative 1y maintaining the isolation valves in the line to the condensate storage tank open. Should an isolation valve be shut, rupture of the low pressure non-QA piping could occur, but loss of auxiliary feedwater would still be limited as before by the flow orifice and pressure control valve in the recirc line.
Based upon the above, it was concluded that the auxiliary feedwater systemLwould still perform its design function during performance of SMP #850. The auxiliary feedwater system drawing in the FSAR shows this relief valve, therefore this evaluation was required.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences'of an accident or malfunction of equipment important to safety is not increased. The change does-not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.34 SMP #879:
Installation of Unit 1 (Existing) Instrument Air Header Flow Meter, Isolation Valve and Pressure Taps.
Summary of Safety Evaluation: Performance of this evolution constituted a potential change to the facility i
I or its operation as described in the FSAR. This 30
e---
p b
l p
p procedure controlled the work to install an isolation g
~
valve and flow orifice in the 3" NB-1 instrument air' (IA) line. A portion of the main instrument air header was depressurized to perform this work..An air supply was maintained to plant components either from the newly installed IA header or from temporary connections from the service air system. The only components which were not supplied were the 1P29 recire valve and the Unit 1 turbine building strainer.
These components were not required during cold shutdown.
The new header, which supplied mainly Unit 2 components during normal operations, was installed per' original plant IA piping requirements and leak tested per PBNP 3.2.5.
The connection from the service air system was rubber hose or P-tubing.
The service air system was supplying IA loads downstreem of IA-16, IA-17, IA-19, IA-30 and IA-65.
The service air system had sufficient capability to supply these loads; only one of the two compressors is needed to supply, normal demands. The service air compressors are oil free, so oil contamination of the lines was not a problem. A low service air pressure alarm in the control room was available to alert the operator to a dropping header pressure. Because the connections from service air were temporary, the possibility of losing a jumper, and thus, air pressure to those components was addressed. Loss of air to components considered significant to plant operation downstreem of the valves temporarily supplied is addressed below.
a.
IA-16: This jumper will supply the pumphouse service water strainer, screen AP and forebay level indication. Loss of air would prevent auto start of the screens on high AP.
The forebay level indication would fail. Screens can be run in manual if necessary.
Forebay level is not a concern since we are not in an ice melt condition. Service water Zurn strainer blowdown valves are normally closed and can be blown down manually if necessary.
b.
IA-17: The G01 and G02 service water valves are supplied by this jumper. The valves fail open on loss of operating air which is a safe condition.
The glycol thermostats would prevent excessive circulation of cooled glycol to the oil cooler / heater. The control room would get a lube oil temp hi/lo alarm and the operator could isolate the raw water as necessary until the jumper and/or air supply were restored.
c.
IA-19: This valve is the main instrument air isolation valve for the Unit I turbine hall. Many of the loads are not of concern during an outage while the secondary systems are secured. The 31
kg 4
g domestic water fill valve and hydropneumatic tank-fill (air makeup) will'be supplied by this jumper.
g
'These can fail for.an extended period without-n.,
adverse effects until the jumper.is restored.
The purge'.. supply'and exhaust control panel, purge exhaust-fan and purge valves are to be. supplied by L
this jumper. During the refueling outage or cold shutdown if containment requires isolation, such as during-fuel movement, a' loss.of operating air would..
not immediately be a problem. The purge valves fail' shut and their associated accumulators could supply the inflatable seal until the jumper was-reconnected.
The emergency fuel oil' tank' fill valve.will be-supplied by this jumper. -This valve is fail closed on loss of' operating air. The control room.will get 3
a fuel oil tank level hi/lo alarm to indicate a problem and the valve can be gagged open'at that time,.if necessary.
The jumper will alsoLsupply the' fuel oil, m
' transformer and~ seal oil deluge' systems. A prerequisite for the procedure is to have secondary systems secured; thus the fuel oil deluge system is the only one of concern.
If this air supply is lost, a supervisory alarm.is received in the control' room and the problem can be investigated.
d.
IA-30: This jumper supplies 1&2A0V-825C, bypass around safety injection pump suction 825A&B. If air is lost, the valve fails open. This valve is maintained"open on an operating unit.
It may be isolated on Unit 1 for pump isolation and thus is not a problem.
The boric acid evaporator will be lost if the air supply is lost.
The jumper also supplies RCV,-018 and its flow transmitter.
If air is. lost,'a discharge, if in progress, would secure. 'The jumper will supply charging pump speed controllers and HCV-142.
Charging pumps,-if running, will go to minimum speed and HCV-142 will fail open. Depending on plant conditions, labyrinth seal AP may reduce to be lost until the air supply is restored.
Reactor makeup water pump discharge valves are supplied by this jumper and will fail open (thus having water available) on a loss of air.
RC-508, reactor makeup water to containment, is supplied by this jumper and on a loss of air would 32
fail shut. The valve. supplies the reactor coolant pump standpipe and pressurizer relief tank and a loss of air would cause no immediate problems. The jumper will also supply chemistry lab equipment and laundry and chemical tank levels. A loss of air to these facilities is not considered to be an hmnediate problem.
Another item supplied is air-operated sample valves which fail shut to a safe condition upon loss of air.
e.
IA-65: This jumper supplies the gas strippers and blowdown evaporator. A loss of air would ultimately result in losing the gas strippers, requiring recovery after restoring the jumper. The blowdown evaporator would become inoperable and would be recovered after the air supply is restored.
Based upon the above considerations, the loss of service air to instrument air jumpers results in conditions that require actions to return a system to normal but do not create unforeseen undesirable conditions. The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.35 SMP #891 Unit 1:
Residual Heat Removal Flow Orifice IFE-626 Change / Inspect 1HCV-624 and IFCV-626.
Summary of Safety Evaluation: The special maintenance procedure constituted a potential change to the facility or its operation as described in the FSAR. This was an evaluation of the methods, techniques and equipment to be used to drain portions of the Unit 1 A residual heat removal loop to facilitate maintenance, specifically the reversing of an orifice and maintaining of two flow control valves. The evolution was proceduralized using the above-identified special maintenance procedure.
Redundant decay heat removal was maintained in accordance with Technical Specification 15.3.1.A.3.b(3),
e.g., one loop out of service with the reactor vessel head removed and the cavity flooded.
An evaluation in accordance with the requirements of 10 CFR 50.59 was required because the SMP: (1) involved the installation and removal of a drain line and its use in draining portions of the RHR systen;; and (2) introduced air to and removed it from the RHR system.
33
Using an SMP ensured the installed drain line would be removed following the evolution, thereby returning the l
RHR system to its design configuration.
l The use of the SMP directed the activities to be performed and defined the activities and responsibilities of the personnel involved, including parameters observed, conditions monitored, and parameter limitations to be met.
The SMP also provides specific guidance on equipment manipulation.
Protection from air entrainment into the operating RHR loop is provided by isolation valves, elevational head differences in RHR system piping and procedural limitations placed upon equipment operation and the recovery (air removal) phase. All air exhausting was done into the reactor vessel; subsequently vented to the containment, thereby preventing entrainment into the operating RHR loop.
In the unlikely event that problems did occur, AOP-9'C,
" Degraded RHR System Capability," provided guidance to recover decay heat removal capability.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of ar.
accident or malfuntion of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.36 SMP #897: HAD Replacement, Unit 1 Oil Reservoir and Hydrogen Seal Oil Deluge Systems.
Summary of Safety Evaluation: SMP #897 covered the installation of new HADs and local control units on the Unit 1 turbire lube oil reservoir and Unit I hydrogen seal oil deluge systems.
These deluge systems are not specifically indicated or described in the FSAR. There are no Technical Specification requirements.
The fire protection systems description section of the safety evaluation report dated August 2, 1979, indicates that the turbine lube oil reservoirs and the hydrogen seal oil systems are provided with automatic sprinkler systems with pneumatic rate-of-rise detectors.
The replacement of pneumatic rate-of-rise detectors with electric rate-of-rise detectors does not change the operation of the deluge system.
l 34
t While the new HADs and control unit were.being installed,
' the system was not automatic as indicated in the safety evaluation report.
Manual operation of the deluge system was available.
There were also local fire hoses available. The fire
. protection and safety coordinator was' contacted in.
addition to the duty shift superintendent prior'to.any
. work being done.so appropriate compensatory measures could be taken.
. There:are no nearby vital areas that take' credit for; these deluge. systems.
The' change does not pose'an unreviewed safety question.
The probability of occurrence or the consequences of an
. accident or malfunction of equipment important to safety is not increased. The change do3s not create'the possibility for an accident or malfunction which has not-been previously evaluated. The margin of safety as.
defined in the Technical Specifications is not reduced.
3.2.37 SMP #905:
HAD Replacement, Unit 2 011 Reservoir.and-Hydrogen Seal Oil Deluge Systems.
Summary of Safety Evaluation: The SMP covered the installation of.new HADs and local control units on'the Unit 2 turbine lube oil reservoir and Unit 2 hydrogen seal oil. deluge system.
The evolution. constituted a potential change'to the facility or its operation as described in the FSAR.
These deluge systems are not specifically indicated or described in the FSAR. There are no Technical Specification requirements.
The fire protection systems description of.the SER dated August 2, 1979, indicated that the turbine lube oil reservoirs and.the hydrogen seal oil systems'are provided with automatic sprinkler systems with pneumatic rate-of-rise detectors.
The replacement of pneumatic rate-of-rise detectors with-electric rate-of-rise detectors does not change the operation of the deluge system.
While the new HADs and control units were being installed, the deluge system was not automatic as indicated in the SER.
35
Manual operation of the deluge system was available.
There was also local fire hoses available. The fire protection and safety coordinator was contacted prior to any work done (in addition to the duty shift superintendent) so appropriate compensatory measures could be taken.
There are no nearby vital areas that take credit for these deluge systems.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.2.38 SMP #907 Unit 2:
Installation / Removal of Steam Generator Nozzle Dams.
Summary of Safety Evaluation: An evaluation was required because the proposed change could affect the function or method of a system, structure or component which is described in the FSAR. The functicn of the PORV is to retain RCS pressure during pressurized operation of the RCS and to relieve RCS pressure if RCS pressure exceeds the high pressure setpoint or the low temperature overpressure protection (LTOP) setpoint.
Installation of PORV blocks to block the valves open defeats this function.
In the_ cold shutdown, depressurized, drained down condition, the function of the PORV is not required. The blocks were procedurally installed and removed prior to fill and vent of the reactor coolant system when the PORV function was required.
The evolution does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The evolution does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.2.39 SNP #919: Fuel Oil Storage Tank Cleaning and Inspection.
Summary of Safety Evaluation: The special maintenance procedure constituted a potential change to the facility or its operation as described in the FSAR. The SMP controlled the draining and isolation of the fuel oil storage tanks to allow for cleaning and inspection.
l
Except when transferring the fuel oil between tanks, only one tank was isolated at a time.
Steps 3.1 and 3.15 allowed both tanks to be isolated for short periods of time.
The availability of the emergency fuel oil storage tank was not affected by this procedure. Should fuel be needed during this procedure, the fuel can easily be obtained from the tank trucks and pumped into the operable tank.
The Technical Specification requirements of Section 15.3.7.A.1.c were adhered to throughout the procedure. Administrative controls were utilized to prevent spillage of fuel oil during the transfer process.
The remaining fuel oil was removed from the storage tank and stored in the north parking lot in tank trucks.
The procedure does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The procedure does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.2.40 SMP #920: Replacement of the Hydrogen Storage Banks.
Summary of Safety Evaluation: This special maintenance procedure contained a temporary modification which served to supply hydrogen gas to the plant while the permanent storage equipment was removed and replaced with equipment belonging to the new supplier. This activity constituted a change to the facility as described in the FSAR.
It was estimated that the changeout of the permanent equipment would take approximately four days.
The combined total hydrogen gas on site, considering the temporary supply in conjunction with the existing supply which was " pumped down" before the changeout, did not exceed the normal full capacity of the existing storage facility plus the capacity of one tank truck used for recharging the tube banks. The proposed work did not impact the control room habitability study performed in response to Question III.D.3.4, Control Room Habitability, of NUREG-0737.
The temporary hydrogen gas supply consisted of trailer-mounted tube banks which were similar to, if not exactly the same as, the permanent storage arrangement that was installed by the gas vendor. The temporary facility was located approximately 50' to 37 1
l l
~
19 75'- north of the existing: storage _ system which is east of'the Unit 2l turbine building /, It was installed in the area between the building,and the road with a' L~
minimum clearance from the building of 25'.
The trailer-mounted design was acceptable from'a structural missile protection, leakage _ protection, explosion consequences, and Appendix R standpoint per NSEAS evaluation documented via NEPB 88-026.
i The change does not pose an-unreviewed safety question.
The probability of occurrence or.the consequences of.an accident or malfunction'of equipment important to. safety is not increased. The change does'not create the' possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as l
defined in the Technical Specifications is not reduced.
3.2.41 SMP #927: 2X04 Outage for 13.8 kV Modifications.
Summary of-Safety Evaluations: 'The activities performed via this special maintenance procedure constituted a potential change to the facility or its operation as described in the FSAR. The 4160 V power supply configuration during certain phases of the U2R14 refueling a.nd maintenance outage was as described in the special maintenance procedure. This configuration was evaluated (with and without'a' fast bus transfer, on Unit 1) for its effect on the FSAR,'the Technical Specifications, and the NRC commitments as found in the safety evaluation reports.
Section 8.2.2 of the FSAR states that auxiliary power-required during unit shutdown is supplied from the unit's X03 and X04 transformers. This was the case for Unit 2 as its auxiliary power was supplied from IX04. However, Section 8.2.2 also states that buses 2A03 and 2A04 can be supplied through crossties.to 1A03 and 1A04, respectively, if 2X04 is removed from service. This was the case for Unit 2, and the FSAR was not significantly affected.
The configuration is controlled by Technical Specification 15.3.7 on the auxiliary electrical systems required for plant operation. Technical Specification 15.3.7.A.1 includes the requirement that the XO3 and X04 transformers associated with the reactor to be taken critical are in-service. Since 1X03 and 1X04 were in service while Unit 1 was operating, and Unit 2 was at cold shutdown, the Technical Specification is met.
A loss of all AC power to the auxiliaries for either unit, as presented in the FSAR Chapter 14 safety analyses, was not affected by this SMP. An initial assumption in the analysis of this event is a loss of 38
all power to the. station auxiliaries. A loss of power to Unit 2 would have no effect because 2X03 and 2X04 were out of service. A loss of power to Unit 1 would have created the situation analyzed in Chapter-14 of-the FSAR. As this is an analyzed condition, the SMP did not adversely affect the safety analysis.
Plant safeguards electrical loads for both. units and nonsafeguards 480 V loads for Unit 2 were carried by the 1X04 transformer while the fast bus transfer-,
capability for Unit 1 was maintained.
In the event of affast. bus transfer, IXO4 would have also' carried the Unit I nonsafeguards loads. This configuration differed slightly from the normal outage lineup.and, therefore, was evaluated. Unit I was assumed to be at power and Unit 2 was assumed to be in cold shutdown during the performance of this SMP.
Maintaining the 4160 V fast bus transfer' capability for.
Unit I while. all Unit 2 buses and all Unit 1 safeguards buses were supplied by the 1X04 transformer,. required -
bypassing an interlock for breakers 1A52-37 (1A01 to 1A03 tie breaker) and 1A52-55 (1A02 to 1A04 tie breaker). This is an abnormal configuration and as such-
-also required: evaluation; As the fast bus transfer capability itself-is not required.to maintain safety-related functions, its operation does not require analysis. However, the potential for overload of the
.1X04 transformer and its related hardware was analyzed to ensure that the preferred safeguards power source is not adversely affected.
A conservative load analysis assuming a safeguards actuation and a fast bus transfer on Unit-1 while at full power and Unit 2 at cold' shutdown was performed to support the safety evaluation performed for RMP'#47.
The 4160 V configuration in RMP #47 is identical to that in this SMP. The applicability of the load values used in the analysis assumed the following: Normal full power 4160 V loads for Unit 1, Unit 14160 V safeguards loads (i.e., two safety injection pumps), and full-rated loads on all Unit 1 and Unit 2 480 V buses (safeguards and nonsafeguards). This analysis excluded the Unit 2 l
4160 V safeguards loads due to the cold shutdown condition.
Results of the analysis indicate a total load on IXO4 and its related hardware of ~35.7 MVA. Since 1X04 is rated for 37.7 MVA with forced ventilation and having forced ventilation is a condition of the SMP, the performance of 1X04 will not be compromised. Using the load analysis, the loads on each of the bus supply breakers (lA52-36 and 1A52-56) can be shown to be well within their 3000 ampere continuous ratings.
39 l
m__.
p
'At one pointiin_the procedure, 1X04'and 2XO4 were both' i!
in service.and supplied by.1X03. ~0peration of a Unit-2 reactor coolant pump (RCP) was. periodically needed while-this configuration was in place. LTo avoid
' overloading IXO3,'the procedure stated that'the gas
~
turbine generator (G05) must~be in operation and supplying power to bus.H01 while the RCP was running.
Possible failure of G05 while~a' Unit 2 RCP was running.
did not cause a safety concern for the following reasons.
Should this have, occurred, the loading on'lX03 would have been greater than rated; however,. ANSI /IEEE C57.92-1981 states.that for the conservative conditions of an ambient a
temperature of 30 C (86 F) and.an equivalent load before peak load of 50% of maximum nameplat_e rating, the IXO3 transformer can withstand a peak load of 191% of rated' load.
~for one-half hour, or 167% of. rated load for one hour, without a sacrifice of transformer life. Since the-procedure required that the Unit 2.RCP be secured upon a shutdown of G05' and a. fast bus transfer and reactor trip on Unit 1, the overload condition (~108% of rated:16ad with one Unit 2 RCP running) on 1X03 would be relieved within minutes through operator action. The worst case loading on IXO3 fell well within the capabilities of the transformer.
Therefore, its operation would not have_been compromised.
The analysis shows that the worst case loading of'a Unit I safeguards actuation, a Unit 1 fast bus transfer, and Unit 2 at cold shutdown would not have been detrimental to the preferred safeguards. power _ source..A spurious Unit 2 safety injection signal would not affect' the load analysis because it already assumed maximum loading on the Unit 2 480 V safeguards buses, and because.the two Unit 2 safety injection pumps would not operate automatically.. Even with a fault in the 1X04 transformer, the backup safeguards power source (i.e.,
emergency diesel generators) would be available to.
supply the power requirement of one complete set of safeguards equipment for Unit 1 and to provide sufficient power to allow Unit 2 to remain'in cold shutdown. This worst case scenario is bounded by the analyzed loss of all AC power to the auxiliaries event presented in Section 14.1.11 of the FSAR. The reliability of the electrical system was not compromised because the system would be operated within its rated capacity.
Since existing plant hardware was used to carry less than rated loads, and the configuration described in this SMP was not affecting the Technical Specifications or the inputs to the FSAR analyses, this SMP did not involve a significant increase in the probability or consequences of an accident previously evaluated. No significant hardware changes are associated with this SMP; therefore, it does not create the possibility of a 40
new or different kind of accident from any accident previously evaluated.
Finally, this SMP did not involve a significant reduction in a margin of safety.
because the loads on the electrical system will not exceed the rated capacity for which it is designed.
This SMP, therefore, did not involve a significant hazards consideration.
3.2.42 SMP #937: CC-00747D Replacement and CC-00130 Repair.
Summary of Safety Evaluation: The work to be done per SMP #937, changeout of valve CC-00747D via MWR #870873 required freeze sealing the unisolable discharge line or removal of the component cooling loop from service.
Freeze sealing was selected. A freeze seal was installed on the piping downstream of CC-00747D. An evaluation was required because the proposed evolution could affect the function or method of a system, structure or component which is described in the FSAR.
Prior to freeze sealing, the discharge piping was checked for integrity via MT inspection. Wall thickness checks were also performed. After the freeze was complete (melted), an MT inspection was again performed.
Loss of the freeze seal or fracture of the 1" pipe could have resulted in a component cooling water leak.
The line is at approximately El. 12' of the primary auxiliary building and is on the suction side of the component cooling pumps, thus seeing the head of the component cooling surge tanks. The end of the pipe by the valve is flanged.
If the freeze melts, flanges could have been made up.
The evolution was a direct changeout with a replacement valve. The duration of the job was short.
If the pipe would have fractured, the short-term fix would have been to thread the open end of the pipe and cap it.
In either event, the supply of component cooling water to the RHR heat exchangers would not have been impaired since reactor makeup water could have supplied component cooling losses. Other component cooling loads during the refueling outage, other than RHR, were not considered to have been needed.
The generic acceptance of freeze sealing carbon steel pipe was performed under SER 88-106.
3.2.43 WMTP 11.50: Processing Waste Holdup Tank with Chem-Nuclear Demineralizers.
Summary of Safety Evaluation: The procedure was developed to process waste holdup tank liquid waste via 41
n 1
a demineralized train to the waste' distillate. tanks prior.
l(:
to' sampling and discharge..This method of waste liquid processing is utilized when the blowdown-evaporator is not available and storage of waste liquid poses a problem-.
An evaluation was required because:the proposed evolution constituted a change to the facility as described in the FSAR and also constituted a change to the. procedures-described in the FSAR. The probability of an occurrence of an accident was not increased because.the temporary connections-and. installation is equal to'or better than the normally installed configuration. The hose and-~
camlock fittings'to be used are' rated to 225 psig. The' temporary demineralizers are rated at only 100 psig, however WMTP 11.50.contains precautions to maintain
<100 psig on the pump discharge.
The consequences of any inadvertent release were equal'to those which would have occurred.if the installed,. normal x
- system experienced a failure. The temporary equipment is housed in the auxiliary building, thereby affording the.
radioactive liquid and radioactive gas handling capabilities normally available.
l 42 Q_--_----_,
3.3' Tests or Experiments
)
The following tests or' experiments performed at Point. Beach Nuclear Plant ~during 1988 required a 10 CFR.50.59 review:
i 3.3.1 SMP #895: Residual Heat Removal Pump Cubicle Ambient-Air Temperature Measurement During Simulated Loss of HVAC and Recirculation Phase After a LOCA.
l Summary of Safety Evaluation: An evaluation was i
required because the procedure constituted the conduct.
{
of a test or experiment not~ described in the FSAR.
Environmental qualification of safety-related
)
electrical equipment located in mild environments-within the auxiliary building is based upon a design temperature range of.65-85*F.
This range may be exceeded.for short periods of time due to anticipated operational occurrences without. degrading the. equipment (reference 10 CFR 50.59 definition of mild environments). The justification for'this' range notes that the.suxiliary building HVAC is powered from ori capable of being powered from the emergency diesel generators (EDGs). The auxiliary building HVAC equipment is not included in the EDG load analysis.
The concern is that ambient air. temperatures in mild areas may exceed the design range significantly during operation without'HVAC.
SMP #895 conducted a test.to measure ambient air-temperature within an RHR pump cubicle during simulated accident conditions. The simulation of accident conditions involved operating the RHR pump during RCS heatup with two area HVAC supply vents' secured.
The ambient air temperature was monitored and recorded every five minutes until either the temperature stabilized or increased to 104 F.
If the temperature reached 104*F,.the HVAC supply.was'to be restored as pump operation with ambient air temperature in excess of 104*F would reduce the qualified life of the pump motor.
If the ambient air temperature did not stabilize at a temperature lower than 104 F, further action would be necessary to resolve the finding.
During this test, RHR pump 1P10A and steam generators A and B were available to provide additional residual heat removal. During the test, the pump temperature rise above ambient was not allowed to exceed the nameplate rating (90*C rise).
The test does not pose an unreviewed safety question.
l The probability of occurrence or the consequerm of an accident or malfunction of equipment important 43
f
.I J l'
safety.is not increased. -The. test doesinot create the possibility for an accident or malfunction which has not been previously evaluated. -The margin'of safety as
~
defined in the Technical Specifications is not reduced.
3.3.2-WMTP 9.22: Emergency Diesel Generator Room Ventilation Test.
- Suminary of Safety ^ Evaluation: An evaluation was
-required as the procedure described the conduct of a test or experiment not described in the FSAR. The EDG room ventilation test was used to determine the adequacy'of the EDG room ventilation system under.
..various configurations, including a' simulated failure of the automatic intake louvers on the east wall of the room. The test ran coincidentally'with an EDG biweekly-exerci'se with a generator loading of approximately 2500 kW.
The EDGs are discussed in Sections 8.1-8.3 of the FSAR and in Technical Specification 15.3.7 and 15.4.6,'and are considered equipment important to safety. The consequences of this test'did not affect the operability of this equipment since precautions were taken during the test to monitor and control the operating environment of the EDG within that specified by the EDG manufacturer.
Execution of the test does not pose an unreviewed safety question..The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The test does not create the possibility for an accident of malfunction which has not been previously evaluated.
The margin of_ safety as defined in the Technical Specifications is not reduced.
44 1
1 F
1 I
(
]
3.4 Design Changes There were no plant modifications made during 1988, beyond those authorized with license amendments as noted above, which required Nuclear Regulatory Commission approval.
The following modifications made at Point Beach Nuclear Plant during 1988 required a 10 CFR 50.59 r.eview:
3.4.1 E-256 (Unit 1), Electric Generator. The modification I
installed a synchroverifier relay on'the main generator, having a contact in the breaker control scheme such that the generator breaker cannot be closed if the generator is out of synchronization greater than the setting of the relay.
Summary of Safety Evaluation: Non-nuclear safety related.
3.4.2 82-067*A, Fire Protection. This modification installed a fusible link on each fire damper for the emergency diesel generator exhaust. ventilation fans. A total of 16 fusible links were required.
Summary of Safety Evaluation:
This work package was a potential change to the facility or its operation as described in the FSAR.
Installation of fusible links on the EDG exhaust ventilation dampers, located on El. 26' of the turbine building, is an Appendix R requirement and the modification provides a 3-hour barrier around the EDG rooms.
Operation of these fusible links will not have a detrimental effect on diesel operation for the following reasons:
a.
Isolating the primary exhaust ventilation of the EDG room will not impair short-term EDG operation.
The outside supply air path to these rooms will not be affected and this path will provide adequate EDG combustion air. The intake air flow will also present some EDG room cooling, although minimal. A short period of time would be available prior to EDG overheating concerns to accommodate immediate identification, extinguishing of the fire and manual opening / blocking of the fire dampers to allow recovery / continued operation of the EDG.
b.
A fire in this area is unlikely. A significant fire coupled with a need for the EDGs is even more remote.
If a large enough fire occurred in the area around c.
the fire dampers which caused all of the fire dampers for a room to close, securing of the EDG room from the fire would be preferable to the risk of having the fire spread to that room.
45
t' d.
If a'fireLshould occur in the. vicinity of these dampers,-the dampers can be manually reset once.the fire has been extinguished in a manner similar.to-the current configuration.
3.4.3 MR 83-021:(Common), Post-Accident Containment Ventilation System. The modification designed and installed the electrical and piping interfaces necessary to easily install a. hydrogen recombiner following a i
postulated accident..The modification connects to the post-accident containment ventilation system' (PACVS)..The electrical requirements of'the.recombiner-are provided from nonsafeguards' supplies. The electrical installation is non-seismic'and does not require redundant power supplies. The piping and valves which extend the containment pressure boundary are seismic Class I.
The supply from containment to the hydrogen recombiner branches off the PACVS containment vent'line after the first valve downstream of the containment penetration'to the primary auxiliary-building exhaustifilters. Two locked shut manual diaphragm valves are provided in series for containment isolation per FSAR Table 5.2.1.
Downstream of the valves is a test connection followed by a blank flange. The blank flange can be removed to permit installation of the hydrogen recombiner when required.
The test connection uses the same type of valve.as the existing containment penetration test connections with a capped connection after the valve. This test connection is intended to permit leak testing of the containment' isolation valves, the recombiner, as well as being a purge supply connection for the recombiner.
The return to containment from the hydrogen recombiner branches off just before the second valve 'after the containment penetration which normally supplies compressed air to the containment as part of the PACVS. A manually-operated, normally open diaphragm valve is installed just upstream of the branch connection and the existing gate valve has been removed. This configuration permits testing of the new valves. The return line from the recombiner has a test connection upstream of the two locked shut, manually-operated diaphragm valves located in series. The test connection uses the same type of valve as the existing containment penetration test connections with a capped connection after the valve. A second test connection and blank flange follows the manual isolation valves.
The second test connection permits testing of the'new valves without removing the flange. The blank flange can be removed to permit connection with the recombiner when required.
46
______________m_.
n.
Summary of Safety Evaluation: The modification constituted a change to the facility as described in the FSAR. The modification extends the containment boundary and provides a potentia 1' post-accident containment atmosphere recirculation leak path after the hydrogen recombiner is placed into service. The extension of the containment boundary has been designed and fabricated to the same standardsfas the existing PACVS extension so the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. Test connections are provided which enable leak testing to.
be performed prior to connecting the hydrogen recombiner equipment into the containment boundary.
The change does not pose an unreviewed safety question.
The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.4 83-87-01 (Unit 1) and 83-88-01 (Unit 2), Containment Atmosphere Grab Sampling (original modifications approved on June 20, 1985). The addenda to these modifications relocated post-accident containment atmosphere. sample septum isolation valves 3200J and 3200L to be more accessible to personnel wearing Bio-Paks.
Summary of Safety Evaluation: The addenda lengthened the tubing to the sample septum and relocated the isolation valves to the front of the RE-211/212 cubicle. The new tubing and fittings meet or exceed the specifications of the original. The changes do not degrade the seismic integrity of the containment isolation valves. No Technical Specification changes are required.
3.4.5 84-016, Steam Generator Channelhead Ventilation System Ducting. This modification replaced the soft " elephant" trunk ducting on the discharge side of the fan with rigid sheet metal ducting.
The original modification package, including safety evaluation, was misplaced and is not retrievable. When the package was recreated, it was noted that reapproval by the MSS and manager was not required since this modification is very similar in concept and design to the Unit 2 installation. The Unit 2 safety evaluation (from MR 83-083) is given below.
Summary of Safety Evaluation: The steam generator channelhead ventilation system is not safety-related.
The duct has been routed so it and its support do not 47
l adversely. affect' safety-related instrumentation, structures or_ components. This modification involves no unreviewed_ safety questions, and does not require a change;to the Technical Specifications.
3.4.6 84-108, Hangers and Supports. 'The modification removed the orifice plugs installed.in the hydraulic snubber control blocks when the snubbers were removed for regular.
maintenance. The plugs.were susceptible to contamination; they limiteu-the bleed rate adjustment to less then 4"/ minute; and caused the bleed _ rate test results to be nonuniform.
~
Summary of Safety Evaluation: The increased bleed rate will not affect the piping analysis and it will have little"or no effect upon snubber function. The-activation level and bleed rate was tested after completion of the modification and tested per Technical-Specification surveillance requirements to verify acceptability.
3.4.7 85-006, Waste Disposal. The modification installed a sightglass in the spent resin cask fill line directly before'the-armored hose flange and header isolation valves.
Summary of Safety Evaluation:
It_was observed by the Staff-that this modification was the result of an ALARA suggestion by operating personnel to provide indication of resin flow. The modification eliminated multiple entries into a demineralized cubicle to verify that the resin bed is actually dropped. From the 10 CFR 50.59 perspective, this modification does not affect the safe operation of the plant because the sightglass was selected such that the pressure, temperature, and environmental conditions were within its design limits. The work was performed in accordance with B31.1-1967. No Technical Specification changes are required.
3.4.8 85-242 (Unit 2), Handling Equipment / Fixtures. The modification changed the method of attaching the reactor vessel flange protective ring to the internals lifting rig by installing a quarter turn lock to replace the existing threaded adapters.
Summary of Safety Evaluation: The change does not structurally degrade the integrity of the protective ring pick-points. However, the pickpoint connection can be disconnected by a quarter turn of the lifting rods, as opposed to several turns required per the old design. A dropped protective ring (due to an inadvertent rotation of the lifting rods) is not-able to descend into an open vessel due to its large j.
outside diameter (as opposed to the vessel inside diameter). Thus, fuel damage is prevented.
l 1
48
3.4.9 85-243A, Instrument Air System.
Swnmary of Safety Evaluation: The work package constituted a potential change to the facility or its.
operation as described in the FSAR. The three tie-ins made to the instrument air system do not affect
-the safe operation of the plant. This conclusion was based upon the following considerations:
a.
The tie-ins will be made during a refueling outage and when containment integrity is not needed. The tie-ins that involve the Unit 2 containment will also be done when fuel is not being moved, b.
The components that will not be supplied by instrument air during the tie-ins will either be supplied by service air, or not be required for
- service, c.
A loss of service air during the tie-ins will npt adversely affect the plant because the component fails to a safe position or a backup source of air could be arranged for by the operators.
d.
The service air that will be provided as an alternative supply will be the same as instrument air except that the air will be non-filtered and dried. Provisions to filter (and remove some moisture) have been incorporated into the special maintenance procedures for the tie-ins.
e.
The tie-ins will be made to the requirements of B31.1, meeting original system design criteria.
f.
The tie-ins will be administratively controlled and monitored by special maintenance procedure during the tie-in process.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The changes do not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.4.10 85-254, Radiation Monitoring System. The modification relocated the Model V-6 Flotech flow switch in a horizontal pipe per the vendor's recommendations from its previously installed vertical configuration.
The change was proposed because spurious " fail ext" alarms which originated from the flow switches could interrupt the control signal, thus defeating the 49
i-
)
.J
[/ '
{.
t L
control function. The radiation ' monitors.affected by j
.this modification are RE-218, RE-220, RE-223, RE-229 1
-and RE-230.' ~ Monitors 1&2RE-219 are being addressed 1
- via MR 85-068.
Summary of Safety Evaluation: No evaluation is required. The change does not affect the flow switches as they are described in the FSAR.
3.4.11 85-279-01 (Unit-1), Penetrations. The modification
-addendum installed one module in electrical penetration IQ28 for use with.the containment integrated leak rate test (CILRT). system instrumentation.
Summary of Safety Evaluation: The' modification constituted a change _ to the facility as: described in the FSAR. This penetration is located in the area near penetrations of nonessential services above Pipeway #1.
This is consistent with the guidance provided in Section'5.1.2.6 of the.FSAR, " Electrical Penetrations."
This section. states that in containment.there is typically 2-5' vertical clearance between penetrations
-that supply essential services'(safeguards and reactor protection)~and nonessential services, with the nonessential services on top. Outside containment, essential service penetrations are-located within the concrete pipe tunnels while nonessential service penetrations are located above the concrete tunnel ceiling.
~
Separation of redundant trains of safeguards and reactor protection are not jeopardized by this modification addendum.
Integrity of the containment structure was-not affected by this modification. Design, fabrication, j
installation, and post-installation testing was j
conducted in accordance with the appropriate codes, standards, PBNP Technical Specifications, and FSAR.
The penetration was installed while the unit is in a cold shutdown or refueling shutdown condition.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased.
The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
50
l
]
b[4 3.4.12-86-104'(Unit'l), Instrument Buses. The modification
' installed two banana jacks in each of the instrument bus control panels.1-(2)Y01/101, 1(2)Y02/102 in order-i that the normal and' alternate instrument bus power sources'are in' synchronization before.a manual transfer is made. The modifications did not apply to the remaining plant instrument buses because these buses incorporate an automatic-synchronization feature.
Summary of Safety Evaluation: The modification constituted a change to the facility as described inl the FSAR. The banana jacks installed in each panel indicated " normal supply hot," and " alternate supply hot."
A voltmeter indicates if the two sources are in synchronization and the transfer can be m ie.
r The test jacks.were fused to provide isolation-between the two incoming sources.
In addition, fu'ses in the test device ensure that isolation' occurs when this desice.is in use and a fault is present.
-The chrage does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment:
important to safety is not increased. The change does not create the possibility for an accident or malfunction which.has.not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.4.13 87-017 (Unit 1) and 87-018 (Unit 2),' Safety Injection-System.. The modifications add a test connection for valve SI-D8.
This includes a tee and valve (to maintain the low point drain) and tubing below the floor deck plate and a pressure gauge above the deck plate by the wall.
In order to access this valve in the past for the purpose of performing the SI system leakage test, the-floor decking had to be moved and an RWP issued to perform the work. The gauge is used for the IST program only and can be isolated from the system during normal operation.
The test connection is used to accommodate the low leakage testing program.
Summary of Safety Evaluation: The modification was a change to the facility as described in the FSAR. This addition is outside the post-accident recire flow path and therefore is not an extension of containment boundary. Therefore, tubing valves, and tubing fittings were used for the addition and run along the floor, under the deck plate, to the wall. The size and locations of the deck plates, coupled with the supporting configuration, prevent the deck plates from damaging the tubing during a seismic event.
51
It. should be noted; that a' failure of the tubi.ng could
. result-in air binding both high head SI pumps-while running with a low RWST level (NPSH below atmospheric).. The' potential for this occurring'is no
- f
' greater than that for the previous drain connection since the tubing meets or exceeds original design-criteria, is seismic, and is not affected by the deck plate. The extension also has redundant isolation capabilities by using the existing drain valve. Two leakage barriers.are maintained. Thus, in view of the above, this modification has no'effect on the-functionality of the SI system.
The addition was designed per B31.1-1967 and' material selection was per FSAR and piping class specifications.
Supports comply with the Seismic Class 1 considerations of the SI system. A leak check'was performed to ensure leak-tightness'of the addition.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety-is not.-increased. The change does not create the-possibility for an accident or malfunction which has not been previously evaluated.
'The margin of safety as defined in the Technical Specifications is not reduced.
3.4.14 87-059 (Unit 1) and 87-060 (Unit'2), CVCS. The modifications installed a new pressure regulator in the volume control tank (VCT) gas makeup system in each unit and replaced the existing gas makeup line check valve, 1(2)CV-363. The changes provided for finer control of hydrogen addition to the VCT.
Summary of Safety Evaluation: The modifications constituy d a change.to the facility as described in the FSAR. Per Appendix "A" of the FSAR, the VCT and associated piping is Seismic Class 1, while the hydrogen and nitrogen systems are Class 3.
Appendix "A" also states, "The interface between a Class 1 system and lower class system is at a normally closed valve which is capable of remote operation from the control room."
For the VCT gas makeup system, a normally closed valve would be impractical and a remotely-operated valve was not originally installed.
Regulatory Guide 1.29, " Seismic Design Classification,"
for Seismic Category 1 describes the interface as, "The system boundary includes those portions of the system required to accomplish the specified safety function and connected piping up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure when the 52
4 safety function is required." For the VCT gas makeup
_q system, the safety function that is_ required is the
]
isolation of the piping leading to the VCT to prevent the release'of fission gases from the VCT to the primary j
auxiliary building.
Based upon Regulatory Guide.11.29, the'VCT gas makeup.
)
line check valve constitutes the separation between' the Seismic Class 1 and Class'3 piping. The new piping / tubing was designed for Seismic Class I service up.to and including the first' support after the new check valve. 'All material used was compatible with the.
existing arrangement and was in accordance with Westinghouse Piping Class.1 SIR.
A failure of the VCT gas makeup piping is bounded by the safety analysis for a VCT rupture (FSAR Section 14.2.3).
The modifications will not increase the likelihood of a failure of the VCT gas makeup system piping.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident'or malfunction which has not been previously evaluated._ The margin of safety as defined in the Technical Specifications is not reduced.
3.4.15 87-066 (Unit 1), Safety Injection System.. The modification revised the SI test / recirculation line by eliminating a tap and valve 883 and by modifying the configuration of valve 882. The purpose of the modification was to' eliminate a leakage point and to minimize future leakage.
Summary of Safety Evaluation: The modification constituted a change to the facility as described in Sections 5.2 (containment isolation) and 6.2 (safety injection). The configuration changes were made in accordance with the Code and Westinghouse,e system requirements. These minor configuration changes do not increase the probability of an SI system leak from the test / recirculation line nor does it decrease the reliability of the SI system. Removal of valve 883 removes a potential recirculation boundary leak path, although of minor significance.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment l-important to safety is not increased. The change does l
not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
53
3.4.16 87-075 (Un_. 1), Various Systems. Design work packages for each of these modifications were' reviewed by the staff as documented in MSSM 87-11.
Work Package "A" addressed changes for SI system MOVs 826A, B & C.
These " base" modifications changed the closing circuits for MOVs 1(2)-112C (volume control tank to charging pump suction), 1(2)-313 (seal water return containment isolation), and 1(2)-427 (letdown isolation). They also changed the motor operator worm gearing for valves 1(2)-112C and 1(2)-313 to make these valves self-locking.
1(2)-427 is already self-locking.
Summary of Safety Evaluation: No evaluation is required. The closing circuits for these valves were previously modified to provide for simultaneous actuation of both the shut torque switch and a valve closed limit' switch upon valve closure. This was done to eliminate the hammering problems associated with auto closure of MOVs with nonself-locking gear ratior.
Plant experience has shown that this simultaneous' '
l actuation is difficult to consistently achieve and that these switches can actuate independently.
If valve closure is stopped by the limit switch without torquing-
[
shut, valve leakage can occur.
Conversely, if the 1
valve torques out prior to actuating the limit switch, the potential for valve hammering is created and the torque open switch bypass switch would not be made up, which could prevent valve opening.
The modification removed this simultaneous switch actuation and installed self-locking gearing. These-valves now operate in a manner similar to the majority of plant MOVs.
The modification replaced the gearing of valve 1(2)-l' 2B (reactor makeup water to charging pump suction) to make them self-locking..This was done to prevent valve leakage due to the valve operator relaxing after the motor is deenergized.
The modifications do not significantly change the stroke time of any of the valves and improve valve reliability.
The largest increase is less than one second. The valves close within the required times.
l A safety evaluation is not required for these l
modifications because they do not change the facility j
or its operation as described in the FSAR or other i
licensing bases; do not require a Technical Specification change, and do not change a commitment to the NRC.
Documentation of this status is being issued as addenda to the original safety evaluations conducted for Work Package "A" of each of the respective modifications.
54 I
3.4.17 87-089, Fuel Handling System. The modification proposed uce of purchased UT equipment to test for-leaking fuel rods in the same manner as the services i
provided by BBR during U2R12 and U1R14. The only difference is the location of the testing.
It is desirable to perform UT inspections in the reactor cavity during fuel movement when the core is not unloaded. This saves time and wear on the fuel transfer system as opposed to the UT rig being located in the spent fuel pit.
A stand is used to elevate the rig 3' off the lower cavity floor under the normal path route of the manipulator just north of the upender.
Raise / lower bypass switches are used to allow lowering of a fuel assembly about l' away from the rig to preclude any chances of hitting the rig. Then the assembly is moved in a north / south direction by hand cranking the manipulator until it is in position in the rig. A temporary index mark is placed on the indexing dial.
for the rig position. The full fuel assembly weight is held by-the manipulator at all times.
Summary of Safety Evaluation: The original review concluded that no evaluatien was required in accordance with the provisions of 10 CFR 50.59.
It stated that the chances of a fuel handling accident would not have been increased when using the system in the cavity versus the spent fuel pit because the only parts of the rig that touch the assembly are the UT probes and the clamp to fix the position of the fuel assembly.
The UT probes are made of 0.5 mm thick by 20 mm high and 250 mm long fingers that travel in and out of the fuel assembly between fuel rods.
If the fingers strike a fuel rod, the rig immediately stops movement. Any displacement would be with the fingers and not the fuel rod.
If the operator lifted a fuel assembly with the fingers inserted, a grid may sustain some minor damage but the fuel rods would remain intact. The clamp is a cylinder that weights about 30 lbs and rests against the fuel assembly with about 10 lbs of force. If the fuel assembly was raised or lowered with the clamp in place, the clamp would just ride along the face of the fuel assembly without damaging it.
If one tried to lift a fuel assembly with the UT device engaged and it was securely attached (to support its own weight), the fuel handling machine would terminate the attempted lif t via the overweight interlock / cutout prior to risking damage to the manipulator or dropping the fuel assembly. The worst case damage that the UT rig could inflict on a fuel assembly is less than that of the par sipper that was previously used at PBNP. This is because the par device has no interlock to prevent shutting the spring can lid on a fuel assembly.
l 55
a.
)' I
- 1..
b
- Bypsssing.of raise / lower interlocks over an open area-of the: lower cavity.cannot damage a fuel assembly because if the assembly struck the floor of: the ' cavity, s
1 the no-load cutoff would stop' motion.' The liner would be subjected.to an impact of 1700.'lbs'over about V
4 in,
z
['
Since the underwater rig. weighs about 1500 lbs, it is not classified as a heavy. load. The consequences of dropping it into.the spent--fuel pit or cavity'are less
~
severe than those analyzed for a dropped fuel'
- assembly, b
Since the system has been and will be in service for only.short periods of time and then removed ~from the spent fuel pit or cavity, there are no long-term'or
-permanent effects on water chemistry'or reactions with the storage racks or liner plate.
Since it-requires 3 empty spent fuel pit rack locations to use the rig on a spent fuel pit rack, the rack cannot become overloaded when it is used in..the spent fuel' pit.
There are no unusual fuel handling techniques needed to perform the UT inspection. No prior training of operators is required. The use of the purchased UT system in containment does not adversely affect fuel handling in the cavity or other systems, and does lessen the chances of a fuel. handling accident by eliminating the need to transfer fuel to the spent fuel pit to be UT-inspected.
During a seismic event there are no positive methods to prevent movement of the UT rig or1its support table.
The fuel gripper is not designed to accommodate the added weight or sideways motion that would be impurted upon it.
Therefore, it is possible that the fuel assembly could disengage and become damaged if the fuel leakage detection ystem was in use during a seismic event. Fuel assembly' damage.would be likely even if it remained engaged to the gripper since the fuel assembly would be acting to hold the equipment in place.
The modification was approved with a design change which would include provisions for installing the UT rig to the lower core barrel support stand and securing it to the stand to preclude the possibility of increasing the likelihood of an accident or decreasing the margin of safety of the plant.
3.4.18 87-097 (Unit 1), Auxiliary Feedwater System. The modification involved two changes to the turbine drive for auxiliary feedwater pump 1P29:
(1) Replacement of the coupling (gear type) with a coupling (spring 56
r q
1 l
I disc type); and (2) Replacement of the existing i
governor end bearing with an improved version. The improved bearing utilizes an anti-friction type thrust bearing in place of the existing babbitt face thrust bearing. The radial bearing design of this combination bearing was not changed.
Summary of Safety Evaluation: The modification constituted a potential change to the facility or its operation as described in the FSAR. A spring disc Series 54, Size 262, dynamically balanced coupling was chosen to replace the existing gear Model B, Size 2-1/2 coupling. The spring disc coupling transmits a l
predictable thrust force in proportion to the axial shaft displacement. The maximum thrust force developed in the. spring disc coupling is 150 lbs at 0.043" displacement. Testing on 2P29 showed a thrust force of approximately 620 lbs developed in the gear coupling.
The thrust force is caused by the resistance of the coupling to accommodate thermal expansion of the turbine or pump shaft following startup. The spring disc coupling allows for axial growth by the use of disc-shaped spring packs. The gear coupling accommodates axial growth with sliding friction on the coupling gear teeth. The reduced thrust force developed in the spring disc coupling reduces the load on the turbine and pump thrust bearings to a valve within the bearings' design limits, thereby improving thrust bearing life.
The spring disc coupling weighs 26 lbs; the gear coupling weighs 31 lbs, 7 oz (with grease).
Shaft stresses due to the overhung coupling mass are reduced for both seismic and gravity loading with the sping disc coupling installed.
The dimensional configuration and mass distribution of the spring disc coupling is similar to the gear coupling. The rotational mass moment of inertia of the spring disc coupling is slightly lower than that of the gear coupling since the spring disc has a lower weight and a smaller overall diameter (6-5/8" versus 7")
Because the rotational inertia of either coupling is minor when compared to the inertia of the turbine-pump rotating assembly, no significant change in the time to acclerate to full speed (on a fast start) has been seen following installation of the spring disc coupling.
The spring disc coupling is rated to carry 19.4 hp per 100 rpm. Thus at 3600 rpm, the coupling is rated for 698 hp.
The turbine is rated at 415 bhp, which gives a 1.7 service factor (698/415). The spring disc vendor recommends a service factor of 1.0 for boiler feed pumps coupled to smooth prime movers such as electric motors or turbines. The spring disc allows a 200%
l 57
[.
overload (non-continuous) to handle shocks or unusual overloads. The pump manufacturer recommends a service factor of 1.5.for a spring disc-type coupling on~their i.
pumps.
The potential for coupling overloads were considered on a fast turbine start, a turbine overspeed or during pump runout.
Coupling loads during a fast start can be significant due to the high angular acceleration and the rotational j
inertia of the pump. The manufacturers of.the turbine, I
pump and coupling were consulted with regard to'the suitability of the spring disc coupling (Series 54 Size 262) for this service. All parties indicated that the coupling would be adequate given'the 1.7 service factor.
It was generally felt that a fast turbine start would be no more severe than a normal start on an electric motor.
The overspeed trip setting of the turbine is 125% of rated speed.
From the pump laws, pump horsepower increases with the cube of the speed. Therefore, at 4500' rpm (125% of rated), the horsepower loading on the coupling would be:
hp2 = hp1 (1.25)3 hp2 = 400 (1.25)8 hp1 = horsepower required for 400 gpm
= 400 hp (from pump curve) hp2 = horsepower required at 4500 rpm (with the same discharge valve setting as in hpi case).
At 4500 rpm, the capacity of the coupling is:
(45)(19.4) = 784 hp The coupling load at overspeed is less than the capability of the coupling. The 4500 rpm trip setting is well below the 13,000 rpm maximum allowable rpm.
The axact horsepower loading which would occur during a severe pump runout is difficult to predict. An upper limit on the pump horsepower requirements can be determined by extrapolating the pump curves.
This has been done, resulting in a conservative value of 700 hp.
The pump could never reach this condition since the flow rate would stabilize at a point where the pump discharge head equals the discharge piping head loss. Also, with an overload of this magnitude, the turbine speed would drop, causing the pump horsepower requirements to drop in proportion to the 58
b b
cube'of the' speed. Therefore, both of these effects.
were ignored. At 700 hp, the load 'on the coupling is essentially the same as the rated capacity. Therefore,.
l l
it is concluded that the coupling could withstand pump runout.
It should be noted this approach is ultraconservative; thus 700 hp estimated versus-the 698 rated-is not a concern.
The spring disc coupling misalignment limits for continuous operation are *0.025" parallel offset, 10.043 end float,.and-11/3* angular misalignment.
Typical cold alignment readings:at 1(2)P29 are 0.008 parallel offset and 0.04 ' angular misalignment. Total shaft growth-(measued at the coupling, including both the turbine and pump shaft growth) has been measued to be approximately 0.025". -The. cold alignment settings are adjusted to compensate for. thermal growth. The hot (operating) misalignment.is less than the above values for parallel and angular misalignment. Typical shaft misalignment and axial growth is within the capability of the spring disc coupling.
The existing governor end bearing will be. replaced with an improved design bearing. The engineering evaluation for this modification dated September 4, 1987, provides these details. The bearing thermocouple was installed
.on the inboard' side of the new bearing similar to the thermocouple installation on the existing bearing.
This location was chosen because:
a.
The inboard side of the bearing will see higher temperatures due to heat conduction along the shaft. The journal bearing lubricating film and babbitt is more sensitive to high temperatures than the anti-friction thrust bearing.
b.
The anti-friction bearing will not generate as much heat as the journal bearing. Therefore, monitoring the temperature at the anti-friction bearing would give a lower indicated bearing temperature.
The lubrication and lubrication cooling requirements (lubrication type, amount and delivery) for the new bidirectional anti-friction bearing are the same as the original unidirectional anti-friction bearing. Thus, the change is compatible with the lubricating system in place.
The manufacturer certifies that the original design criterion, including seismic qualification, is maintained following installation of the new bearing.
59
p The change does not pose _an unreviewed safety question. The probability of occurrence or the consequences of an_ accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or l
malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
l 3.4.19 87-132 (Unit 1), Crossover Steam Dump System. This modification installed a microprocessor-based control system in the electrical controls for the Unit I turbine crossover steam dump (TCSD) motor-operated valves, MOV-1 and MOV-2.
Summary of Safety Evaluation: The modification was a Ftentialchangetothefacilityoritsoperationas described in the FSAR. The new microprocessor system has been added to the present control circuit to operate, or add the new controls as desired through,an auto / manual switch and control relay.
The new microprocessor controls add additicnal and
.more precise torque / limit settings to the stroke of the valves to test the feasibility of having the microprocessor replace the existing torque and limit switches. The existing torque and limit switches have been adjusted to allow the microprocessor controls to operate first, but if they do not operate successfully, the existing torque and limit switches will operate to prevent 1/alve or operator damage and stop the valve travel.
The possibility of an inadvertent actuation of the microprocessor system is virtually eliminated since the actuating device is electrically isolated from the microprocessor through its optically isolated triac.
Also, each valve has its own microprocessor system for train separation.
The greatest credible failure in this cystem (TCSD) is to have the MOV shut when an actuating signal is generated for the dump valves to open to prevent an exce.ssive turbine overspeed. This could occur through a short in the close circuit of the valve motor operator. However, the possibility of a short in the wiring occurring in the close circuit of the microprocessor system, at a time when the dump valves are required to operate, is no greater than with the existing wiring configuration. Therefore, the new configuration does not increase the previously analyzed accident (i.e., turbine overspeed) or create a new accident.
Since this is a test program, no updates to the FSAR will be performed (they are also not necessary).
Technical Specifications do not apply to this system.
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3.4.20 87-134, Fire Protection System.
Summary of Safety Evaluation: The modification replaced the existing pneumatic heat-activated devices (HADs) on the fire protection deluge systems. The new detectors are electric, heat-only detectors. A local contro.1 unit was installed to provide full supervision per NFPA 70 and NFPA 72 requirements. A 24-hour backup battery was installed in each supervisory panel to provide full supervision and suppression actuation functions upon loss of normal AC power. Battery condition (voltage) is monitored and annunciated. Thus, redundant power is available to ensure protection.
The deluge valves are not described in the FSAR.
However, the supervisory air to the deluge valves is shown on Detail A of Figure 9.6-1.
The supervisory air to the deluge valves has been removed as part of this modification. The new detection, supervisory and release system are all electric. It should be noted that the electric actuation system was originally used (solenoid) for manual remote deluge valve operation.
The third paragraph of Fire Protection Evaluation Report Section 2.4.2 indicates that power for the detection system is from the normal lighting system.
The new detection system for the deluge system is powered from 1Y05.
The change has no significant impacts upon instrument air.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.4.21 87-141, Service Water System. The modification installed retractable pitot tubes on the north and south service water headers.
Summary of Safety Evaluation: The modification constituted a change to the facility as described in the FSAR. For initial installation, the affected header was drained and the plant was in a 24-hour Limiting condition for Operation per Technical Specification 15.3.3.D.2.b.
The meters took three to six hours to instal), so the header was returned to service within the time allowed. The pitot tubes can now be inserted or removed from the header without draining the header.
61
Inctallatfon of the flow meters does not affect the supplying ability of the service water system. A drawing of the service water system is included in the FSAR. This drawing will need to be changed to reflect the installation of the flow meters.
This instrumentation was installed in an effort to obtain measurements required by Section XI of the ASME Codc. However, because of the system alignment, repeatability of the same data point may not be possible.
Pump performance may be evaluated by comparison of test data with the pump design curve and considering the service water loads upstream of the meter.
The connection were made per B31.1-1967 requirements.
The connections were designed ar.d installed to meet seismic criteria. The probability of flooding in the emergency diesel generator rooms as a result of adding these connections is not increased.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.22 87-204 (Unit 1) and 87-205 (Unit ?), 4160 V Electrical Distribution. The modifications made permanent the temporary modifications (TMs 88-02 and 88-03, respectively), which installed jumper wires in the A05/A06 bus tie breaker auxiliary switch contact scheme.
Additional administr0.ve controls have also been imposed on the A05/A06 bus t.u breaker use.
In fact, the modification goes beyond the temporary modifications to install independent wiring paths, to be used for each diesel generator breaker scheme, which allows separation from the single auxiliary switch in the A05/A06 tie breaker.
Summary of Safety Evaluation: The modifications constituted a potential change to the facility or its operation as described in the FSAR. Since an auxiliary cell switch in the A05/A06 bus tie breaker has contacts and wiring attached which is used in each emergency diesel breaker automatic close circuit (both Trains A and B), the potential exists that a single active failure of the cell switch will prevent automatic loading of the emergency diesels under safety injection and loss of offsite AC power. As such, a single failure potential is deemed to exist with the A05/A06 bus tie breaker cell switch.
62
Additionally, when the bus tie breaker is in use, the possibility of a single failure exists with its use due to potential mechanical failure of the breaker or loss of DC control tripping power. As such, the breaker itself presents a single failure potential.
Therefore, use of the breaker should be strictly administratively controlled and its use should be limited to times when ESF power is absolutely not-required (e.g., fuel is fully unloaded from the core or cavity flooded) or under emergency conditions where it is necessary to mitigate the consequences of an accident when its use is authorized pursuant to invoking 10 CFR 50.54 (x) and (y).
To remove the single failure potential, the breaker has been racked out and the cell switch contacts removed from-the diesel circuits. Under conditions of use as-previously specified, the interlock contacts have been administratively reestablished to be connected in the diesel circuits to provide original design protection.
Use of the breaker at any other time would constitute an unreviewed safety question and would require prior NRC approval before use.
A change to the FSAR will be required to address the description of the breaker and its authorized use. No change to the Technical Specifications is required.
The change as described does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.23 88-001 (Unit 1) and 88-002 (Unit 2), Auxiliary Coolant System. The modifications changed the power supply for instrument loop 1(2)PC-629. This instrument loop provides the B RHR pump discharge pressure indication and also the RHR high pressure interlock for the B containment spray pump crossconnect valve 1(2)MOV-871B.
The original arrangement had the instrument loops for the high pressure interlocks for both the A and B spray pump crossconnect valves powered from the red instrument bus, 1(2)Y01. Therefore, a single failure (the loss of the red instrument bus) could have resulted in the loss of ability to go on containment spray recirculation. To eliminate this potential, the power supply for one of the two loops [1(2)PC-629) was changed from 1(2)Y01 to 1(2)YO2, the blue instrument bus. This modification 63 e
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l required moving the Foxboro power supply unit, 1PQ-629, and alarm unit,'1(2)PC-629. Both compu.ents. are located in feedwater' control panel 1(2)C127' This. control panel has plug molds powered from both 1(2)Y01 and 1(2)Y02.
To maintain consistency in power supply locations,
-1(2)PC-629 and 1(2)PQ-629 were relocated internal to 1(2)C127 to' place them in the same area as the other blue-powered instrument.
i i
As a result of. powering the alarm unit'from the blue instrument bus, the high pressure. interlocking relay, 1(2)PC-629XB, was also powered from the blue instrument bus. This relay is located in the' rear section of safeguards Train B panel l'(2)C167. With the exception i
of this relay, the remainder of the relays in the rear of 1(2)C167 are powered from D02, the Train B DC bus.
Since 1(2)YO2 also receives its power from D02, this 1
modification made the power supply for 1(2)PC-629XB more consistent with the remainder of-the relays in the
- rear of 1(2)C167.
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The modification did not affect instrument channel-separation in either panel 1(2)C127 or 1(2)C167 Summary of Safety Evaluation. An evaluation in
]
accordance with 10 CFR 50.59 was not required because the 1
power supplies for 1(2)PC-629 and 1(2)PQ-629 are not
'3 discussed in the FSAR or Technical Specifications.
These modifications did not create an unreviewed safety-question.
3.4.24 88-005 (Unit 1) and 88-006 (Unit 2), Reactor Protection H
System. The purpose of the modifications was to ensure j
that reactor trip bypass breaker cell switch contacts used in the turbine trip circuits do not malfunction for j
an undetermined period of time.
It has been determined l
that the most reliable way to verify cell switch contact i
position is to use test points on the actual contacts of concern. These test points enable'the verification 1
of the bypass breaker cell switch contact position during'the rack-out and rack-in of BYA and BYB by means of a continuity check between the test points.
I Summary of Safety Evaluation. The modifications constituted a change to the facility as described in the FSAR.
Installation and testing of the test points were I
controlled by means of a special maintenance procedure.
These test points are part of the turbine trip circuitry, which is non-safety grade and non-Class IE.
l Installation and operation of these test points i
neither cause nor prevent the protective action of the normal trip or bypass breakers or the turbine trip 1
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L1_u__ __ _ _
, _ - _ _ _ _ =, _
c t
circuits. ' Train separation is maintained consistent ~
with existin'g configuration by running-the new: test 1
point' wires'along;the same routes as"the, existing-4 turbine tripfand alarm cabling.
The change'does not pose an unreviewed safety question.
The probability of. occurrence-or the consequences of an I
accident or malfunction:of-equipment important to safety.'
'is not increased. The change.does not create'the possibility for an accident or malfunction which has not.-
-been previously evaluated.' The margin.of safety as defined in the Technical Specifications is not reduced.
'l 3.'4. 25 ' '88-011-01-(Unit'2), lleating,; Ventilation and Air Conditioning Modification 88-011 replaced the volume:
s dampers which were components of the containment:
ventilation system'ast originally' designed,' but which.
.were removed due to maintenance ~ problems;in the latter 1970s. The addendum replaced / repaired backdraft dampers which are components of the containment _ ventilation'.'
i system as originally designed.'
Summary of Safety Evaluation: The modification'and.
l addendum constituted a potential change (to the_ facility 1
or its operation as described in the,FSAR. The replacement dampers. meet or exceed' original ~ desiga '
specifications. :The modifications' restore the-
=
system'to its original' design which is described in the FSAR. The modifications do not adversely affect.
the operation-of the plant since the installations will d
be performed under administrative controls stipulated ~in j
SMP #923 and were performed during a refueling outage.
Following installation, -the system was f adjusted 'and test'ed'for operability using test procedure WMTP 9.24.
This testing is described in FSAR Section~6.3.1 under the heading of'" Testing of Containment Pressure-Reducing i
Systems Components."
j l
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an I
accident or malfunction ~of equipment important to safety I
is not increased. The change does not create the l
possibility for an accident or malfunction which has not I
been previously evaluated. The margin of safety as.
defined in the Technical Specifications is not reduced.
l 3.4.26 88-011-02, Heating, Ventilation and Air Conditioning.
1 Summary of Safety Evaluation: A system, structure, or i
l component, as described in the FSAR, was changed.
The FSAR describes the pitot grids in a single sentence on Page 6.3-7.
This sentence will be removed from the l
65 u____________..__
)
f h
FSAR following removal of the grids from both plant r
units. Using equivalent sheet metal construction,-the
. ducts were returned.to an. original configuration.
,The change does not pose. an unreviewed safety h;
question. The probability of occurrence or the h
consequences of an accident or; malfunction of' equipment; important to safety.is not increased. The change'does not create the possibility for an accident.or malfunction ~which has not been previously evaluated, a
The margin of safety as defined in'the Technical Specifications is not reduced.
3.4.27
'88-025, Fire Protect 1'on.
Summary of Safety Evaluation: The modification constituted a potential change to the facility or'its operation as described in the FSAR. The content of the.
safety evaluation is classified as safeguards
.information.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.
ia not increased. LThe change does not create the-possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as 1
defined in the Technical Specifications is not reduced.
3.4.28 88-026 (Unit 1), Containment Liner Inspection. The modification consisted of drilling numerous small holes
(%" diameter) in the containment floor slab expansion joints; replacing the floor slab expansion joint seal-in its entirety with a configuration slightly different-than the original configuration; and core drilling 6" diameter holes in the floor slab. All of the holes will be down to the containment liner plate. The-containment liner plate was not disturbed.
This effort was pursued to accommodate investigation (s) into the effect on the containment liner from the trapped stagnant water. The small holes in the expansion joints will allow for potential measurements to identify-galvanic corrosion cells.
The larger holes will provide access to a sample for analysis to determine potential corrosion rates and provide visual inspection access to the containment liner.
Summary of Safety Evaluation: The small holes in the E
expansion joints and the repair of the expansion joint j
seal does not present a 10 CFR 50.59 concern. The expansion joint design is not described in the licensing basis. The expansion joint seal repair material exposed l
l 66
to containment atmosphere was silicone, similar to the original material. The new sealant is slightly.
different in that it remains more pliable than the original. The new seal was not made as thick as the original seal to ensure that the joint can accommodate expansion / contraction cycles without losing a watertight seal.
The barrier provided by the floor slab and associated joint seals is not a containment boundary. Thus, the j
resulting joint configuration is equivalent to, if not
{
better than, the original expansion joint configuration.
The 6" diameter holes are being addressed by k
10 CFR 50.59 because of the FSAR implication that the containment liner is either coated with a protective coating or is in contact with concrete. The liner at the bottom of the 6" diameter hole was left in the as-found condition. -Accelerated corrosion of the liner at this location due to the removal of the concrete. core is not expected since the hole will not be exposed to post-DBA fluid.
Corrosion from the environment within the core hole will be minimal in view of the minimal amount of air and potential stagnant fluid. A bladder.
device or a concrete plug was replaced in the core hole (s) to ensure a minimal amount of air. The modified configuration has been reviewed and approved by the original containment liner designer. An acrylic plate was used on the bottom of the concrete plug to be able to adjust the height of the plug. Acceptability of the acrylic was approved by the architect-engineer. Thus, the resulting configuration meets the intent of the FSAR in that the liner at this location will be protected from accelerated corrosion.
Two of the 6" holes have corrosometer CK-3 installed to allow monitoring of the potential corrosion cell on an ongoing basis. The liner at these locations will be inspected on a periodic interval to ensure that accelerated corrosion is not occurring. This will be performed during the pre-1LRT containment inspection.
The performance of the ILRT will also ensure that a containment boundary problem does not go undetected.
The 6" diameter core holes have been placed in a location which has negligible impact on the strength of the floor slab or tied-in structures. The location is also such that internal containment missiles cannot damage the liner, or missile protection will be prcvided. The modified containment configuration has been reviewed and approved by the original containment liner designer.
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SMP #873 controlled the work for this modification request. The SMP stated that is it acceptable to cut the ends of rebar in the floor slabs, but it contains reviews to ensure that rebar which ties any column or equipment to the floor. slab will not be cut. SMP #873 also contained precautions and a drilling procedure to ensure that the liner plate is not damaged.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.29 88-027, DC Power System. One of the original trouble alarms for battery chargers D107, D108 or D109 was always in an alarm state on control room panel 2C20 since at j
any time, one of the chargers is out of service.
Modification 88-027 allowed for the disabling of the trouble alarm for the charger which is removed from service.
Summary of Safety Evaluation: No evaluation is required. All switch contacts and associated wiring was of sufficient size to support normal alarm current. The logic change did not add any load to the alarm circuit power sources, nor did it degrade them in any fashion. There are no Appendix R concerns associated with the modification due to the fact that all of the wiring changes were made inside 2C20.
Administrative control of this switch can be accomplished by revision of 01-33.
Since the battery chargers are not normally rotated, and OI-33 would be used for a charger exchanger, administrative control of the switch appears adequate.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.30 88-031, Hydrogen Gas System. This modification changed out the existing hydrogen storage system provided by the vendor and replaced it with a new storage system provided by a gas vendor. The existing system is described in i
Section 11.1 (Page 11.1-16) of the PBNP FSAR. The FSAR states:
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" Hydrogen is supplied to the volume control tanks and each generator from a central storage fecility located outside the east wall of the turbine building. The facility includes 50 rack-mounted horizontal cylinders, having a total capacity of 75,000 cu. ft. (STP), a pressure reducing station,.and a low-pressure liquid.
i The pressure reducing station controls hydrogen evaporation and maintains header gas pressure.
Individual pressure controllers at each generator and the volume control tanks maintain required hydrogen pressure."
The new storage system consists of one tandem axle trailer with a capacity of 62,391 scf at 2400 psig in 38 tubes. The trailer is located in the exact same location as the original units.
The structural, missile protection, leakage protection, explosion consequences and Appendix R concerns posed by this change have been evaluated and have been found to be acceptable.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety-is not increased. The change does not create the possibility for an accident or malfunction which'has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.31 88-036, Industrial Safety. The modification provided three permanent noise proof. telephone boxes (Hear-Here boxes). The locations of the booths are: The G01 diesel generator room's south wall near C34A and C78; the north wall of the Unit 1 charging pump cubicles near ILCV-112B; and south of the Unit 2 charging pump cubicles near 2B32.
The purpose of these telephone booths is to enhance the use of new PBX extensions in noisy (or potentially noisy) areas in an Appendix R situation. This is not an Appendix R requirement; therefore, we do not take credit for the booths in any of our procedures.
Summary of Safety Evaluation: No evaluation is required. The exact location of the telephone booths is such that safety-related or seismic category structures or equipment (as described in FSAR Chapters 8, 9, and 11, and Appendix "A" of FSAR Chapter 2) could be i
affected. Therefore, the booths will be seismically mounted. All required attachments were performed using approved procedures. The booths do not affect any NRC IE Bulletin 80-11 masonry walls.
69
1 p
V Other: issues that were addressed in the installation of the telephone booths include:. Installation will be only in areas of relatively low background radiation
-)
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effects upon. ingress / egress and personnel evacuation routes..shall be. minimized; existing equipment accessibility.shall be maintained to the. maximum extent' possible; and the; installations..shall have no effect.
upon any vital or instrumentation buses.
The final design minimizes required maintenance.
t Materials selected were compatible with existing plant-components. -The vital-area's combustibleLloading is not unacceptably increased.
Appendix.R compliance is addressed and the resulting configuration does not impact our Appendix R. commitments.
']
There is no effect upon the plant description as.found.
in Chapters 8, 9, 11 and' Appendix "A" of the FSAR;.
j the plant Technical Specifications, ' and prior commitments
-l
~to the NRC as found in the SERs, or any plant procedure j
as described in these documents.
The change does not pose an unreviewed safety. question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety-is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
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1 3.4.32 88-040, Sampling System.
I Summary of Safety Evaluation: ~The modification constituted a change to the facility as described in the FSAR. Since this line is not normally isolated from the y
l RHR system, it is considered as part of'thelRHR system
-l I
and serves as part of the containment boundary. This modification supports the new sample line such that it 1
can be considered seismically supported to meet the requirements of the FSAR. This is based'upon:
The loop step valve 2SC-958A being supported per the a.
architect-engineer small bore pipe guidelines; b.
The stainless steel tubing being supported based l,
upon the guidelines used'for' seismically supported 3/8" SS tubing for post-TMI modifications.
c.
The material used was in accordance with the l
pipe specifications for this line location.
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1
l The original RHR loop sample lines are included in the High Energy Line Break Analysis (HELBA) in Appendix E of the FSAR. Following the installation of the supports, the sample line that was added by modification M-676 does not affect the HELBA based on the line being seismically supported; the material used having a design rating equal to or greater than the original sample line and the new sample line not being closer to any equipment of concern, namely the refueling water storage tank to RHR pump suction valves (2-856A&B). The resultant configuration is adequate for containment boundary purposes (temperature, pressure and seismic).
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.33 88-042 (Unit 2), Reactor Vessel. The modification installed dosimetry in Unit 2 in the reactor cavity adjacent to the vessel beltline. The dosimetry is intended to quantify the azimuthal, axial and radial gradients of neutron exposure within the vessel.
Measured neutron exposure data from excore dosimetry are necessary to verify that planned fuel management schemes achieve the desired vessel flux reductions.
Summary of Safety Evaluation: The modifications constituted a potential change to the facility or its operation as described in the FSAR. A safety evaluation for this modification has been performed The following evaluation is based upon that evaluation.
The dosimetry is supported by a stainless steel frame which was installed around one of the RV outlet nozzle support shoes. The frame itself rests on the ring girder. The support frame is installed with a nominal 1/8" clearance around the support shoe and is restrained by the two frame braces which have 1/2" bolts tightened against the underside of the ring girder. The total weight of the support frame, bead chains, dosimetry, and miscellaneous hardware is $33 pounds. The bottom termination of the bead chains consists of chain clamps connected to stainless steel eye nuts threaded onto studs in the sump wall.
71
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The specific aspects of the mechanical evaluation are as follows:
)
a.
Loose Parts: The generic safety evaluation determined that there were no adverse effects on the RHR system pumps or valves due to the ingestion of stainless steel dosimetry bead chain.
As described on Page 6.2-9 of the FSAR, the sump "B"
screens prevent the entry of objects larger than 1/8".
This effectively eliminates any chance of the-bead chains entering the system since the nominal diameter of the beads is 3/16".
The loose parts evaluation also considered the effects on the Sump A drain line's containment isolation valves (CIVs).
It was concluded that the potential loose parts did not present a significant safety concern in regard to the Sump A drain line CIVs. This conclusion was primarily based upon the fact that the subject valves are normally closed and would be cicsed immediately following a containment isolation signal. This would be before any debris could migrate to the sump suction line CIVs.
Post-accident use of this line could possibly result in leaking ' valves subsequent to use due to debris (note leakage only in view of the valve type and Sump A grating). Use of this line in this fashion is not anticipated as the need is questionable and access (unless a mild LOCA, low radioactivity in coolant) would be prohibitive. If use was pursued, it certainly would be after the containment pressure subsided; thus little driving force would be present.
In any event, the RHR system would provide for necessary sampling for Sump B recirculation.
b.
Missile Hazards: The design of the support frame prevents the dosimetry system from constituting a missile hazard by two means.
First, the physical size of the frame assembly is too large to permit passage through the narrow annular air gap between the RV insulation and biological shield liner plate.
Second, the support frame is physically restrained by virtue of the fact that it is assembled around one of the RV nozzle support shoes.
In the event of chain failure, the dosimetry capsules and gradient chains are prevented from becoming missiles above the elevation of the nozzles by the geometry of the air gap and the presence of the RV insulation support channel.
72
The dosa<try capsules and gradient chain do not constitutt a missile hazard to the incore detector thimble conduit tubes for two reasons. First, the dosimetry is not installed above the incore tubes and second the mass of a capsule is approximately-one-third of the most conservative acceptable impact mass.
The dosimetry capsules and gradient chain do not constitute a missile hazard to the containment liner plate. The liner plate is protected by 18" of concrete in the region below the RV, and by a minimum of 12" of concrete on the floor of the incore tunnel. The exposed liner plate in the incore tunnel is not at risk because of the geometry of the tunnel and due to the small mass of the dosimetry capsules.
c.
Effect of Added Mass: The installed hardware does not affect the capacity or operation of the RV.
support system or the RV insulation during normal operation or during seismic events.
d.
NIS Excore Detectors: The reactor cavity neutron dosimetry system is totally passive and does not interact either neutronically or mechanically with the active excore neutron detectors. This is due to the fact that the reactor cavity neutron dosimetry system is designed to produce free field measurements with little or no' local perturbation of the' neutron flux and that the excore detectors are physically isolated by the steel positioning tubes and the reactor cavity liner plate.
e.
Materials Used: With the exception of the aluminum chain stops and dosimetry capsules (43.5 pounds),
the reactor cavity neutron dosimetry system is constructed of stainless steel.
The installation was also reviewed with respect to the mass of aluminum in containment. The amount of aluminum in containment is restricted since in a post-LOCA environment the aluminum corrodes, forming hydrogen gas which is a flammability concern. As indicated in FSAR Table 5. i.2-2, the present mass of aluminum in the containment is 566 pounds (of which 100 pounds is identified as contingency). The total mass of aluminum that would be added to the containment by the addition of the dosimetry is about 3.5 pounds Not only is this a small fraction of the contingency value that has been taken into account, it is also a very small increase in the total mass of aluminum. Additionally, per FSAR Appendix B, Table 9, aluminum corrosion contributes only about 10% of the total hydrogen production. Thus, the assumptions regarding the times of hydrogen purge to 73
r_
control hydrogen buildup remain unchanged and the calculated doses due to containment purge that are reported in Appendix D of the FSAR are also unchanged.
-Table 5.6.2-2 will be updated to include the addition of 3.5 pounds of aluminum to the containment inventory.
The change does not pose an unreviewed safety question.
The probability of' occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.34 88-044 (Unit 2), Feedwater System. The Unit 2 No.-4 feedwater heater tube side relief valves (2RV-2406 and 2RV-2408) originally discharged through drain funnels to the floor drain system. When these relief valves lift, the funnel and drain capacity is insufficient to handle the flow. A significant amount of splashing and spraying occurs which poses a burn / scald hazard to personnel and'a flooding hazard to equipment, most notably the Unit 2 secondary sample panel. The modification will eliminate the discharge to the existing funnels and drain system by piping 2RV-2406 and 2RV-2408 to the atmospheric blowoff tank (2T71).
Summary of Safety Evaluation: The modification constituted a change to the facility as described by the FSAR. The relief valves and the existing discharge lines are shown on the feedwater vent and relief drawing in the FSAR (Figure 10.7-7A).
The valves and their discharge piping are not rentioned or described anywhere else in the FSAR. No credit for the relief valves is taken in the safety analysis for the reduction in feedwater enthalpy incident (section 14.1.6); the excessive load increase incident (Section 14.1.7); the loss of normal feedwater accident (Section 14.1.10); or any of the other events analyzed in Chapter 14 of the FSAR.
The relief valves and piping satisfy Wisconsin Administrative Code and FSAR Section 10.1.5 requirements by providing overpressure protection in accordance with ASME Code Section VIII and USAS-B31.1.
The new piping was designed and installed to maintain compliance with these code requirements.
The relief valve discharge lines are non-seismic and not safety-related. The small additional flow to the atmospheric blowoff tank will not significantly affect its pressure relieving capability. Since the new discharge lines are open to the atmosphere via the blowoff tank, B31.1-1986 exempts the discharge piping 74 l
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.i from_hydu tatic1 testing. ~The_resulting,configurat' ion?
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.'does notipose a high' energy:line break problem since-the-line is'normally:att atmospheric pressure 1and-in view of' tits location.
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!The change'does not pose an unreviewed safety question.
The probability of' occurrence ~or.the consequences _of ani R
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Jaccident or malfunction of equipment'. important. to safety
~
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is not increased. The change does not create the' possibility for a'n' accident or malfunction'which has-not been previously evaluated. The margin ofLs'afety_as'
. defined.in the Technical Specifi'ations is'not. reduced.
c 3.4.35 88-047 (Unit 1), Incore Instrumentation' System.x The-modification removed the thermocouple-(TC)) connectors'on incore TCs F12; I7, H6, and H10J(or'L10 pending.the outcome of-the post refueling TC stalk _ inspection) and Einstalls a 1/8" 316 SS tubing cap to seal the TC assembly.
i These TCs had developed leaks in their~ sheaths-which.
y
- result in reactor coolant leaking out at their
- electrical connectors.
. Summary'of Safety Evaluation: The modification is'.a potential change to the_ facility.or its operation'as described in the:FSAR, From an instrumentation standpoint, this' modification will not create problems because:.(1);
g.
Suitable replacements for H6 and H10, which are. inputs to the reactor vessel level instrumentation system and subcooling system, have been' identified, and (2) Only,4 out of 39 thermocouple will be.out:of service,and will not affect-the ability to meet'the-TS requirements of Table 15.3.5-5 to have four.TCs per core quadrant.
From a pressure-retaining standpoint, all the tubing caps used were compatible with the TC sheaths. TheJ installation of the tubing caps is also similar to the 7
existing electrical connectors. All material-in contact:
L4 with primary coolant is stainless steel:and'will not be W
affected by contact with boric acid.
If a leak through a tubing cap develops, it will not. exceed the capacity of one charging pump-(FSAR Section 14.3.1).
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
m 1
1 75 1
____.._m_____.-__.m_
e i
3.4.36'88-061 (Unit 1) and 88-062 (Unit 21, Turbine Crossover Steam Dump System. The modifications add _a means to monitor the internal valve pressure above.the disc of.
the turbine crossover steam dump valves. This is being
'done to aid in maintenance troubleshooting of the recurrent events.in which these valves fail to close following actuation during testing.
Summary of Safety Evaluation:.No evaluation is required.
The modifications added a tubing fitting and a~ isolation-valve to original bonnet ports.which'are presently tapped had plugged. A tubing cap was installed over the outlet end of the valve. -A pressure gauge-was provided to allow for pressure monitoring, when appropriate.
All components either meet or exceed system pressure and
~
temperature requirements. The changes were made in accordance with the requirements of the original installation code, B31.1.
Additional' components did not-significantly impact the main valves' seismic capabilities (which appear to be Seismic Class III).
Failure of the gauge or pressure taps will not impact the ability of the. system to perform its function of ensuring that the turbine rotating elements remain below design rates speed (132%) in conjunction with IOPS. A failure would only pose the potential for opening of a turbine crossover steam dump valve.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipoment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety ar defined in the
-Technical Specifications is not reduced.
3.4.37 88-073 (Unit 2), Main Control Boards. The modifications relocated the LTOP switches and indication lights on 2004. The switches were moved from the vertical panel to the bench panel so they will be near the switches for the respective PORVs and isolation valves-. The modification also required that the switches for the PORVs and isolation valves be rearranged.
Summary of Safety Evaluation: No evaluation is required. The modification maintained and slightly improved the separation between the two trains.
Rearranging the switches also slightly improved the separation over the presently existing configuration.
The key-operated switch for the LTOP system is described in FSAR Section 4.2.3 and the NRC safety evaluation report dated May 20, 1980. These modifications do not affect either of these descriptions.
76
3.4.38'88-080, Radwaste Steam System. Modification 88-080 addressed a recurring problem of water hammer that existed during startup of the blowdown evaporator reboiler. The problem developed when radwaste steam supply valve RS-21 was first opened. The problem was t
that the condensate level in the reboiler was maintained to almost completely full on the shell of the heat exchanger and the piping / valve configuration provides a water trap just downstream of RS-21.
When RS-21 was
~
opened, water flowed down back into the radwaste: steam line. This steam / condensate mixture caused water hammer. This modification revised the piping to prevent drainage back down the radwaste steam piping. In addition, a small bypass line included to provide finer control when bringing in steam. When the condensate is now below the level of the 6" steam supply line, the RS-21 valve can be opened to provide full flow. This should minimize backleakage, thus reducing the chances for water hammcr.
The FSAR provides a description of the blowdown evaporator in Section 11.1, but modification 88-080 did not change i
this description or the role described therein. The modification was designed per the requirements of the code of record and specification requirements. The original flow characteristics will remain unchanged.
The bypass line will not affect the original function of the heat exchanger, but will only provide finer cot, trol of steam into the heat exchanger during startups.
The modified line is not within the area of concern for a high energy line break so this issue was not evaluated.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment-important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.4.39 88-107 (Unit 2), Safety Injection System. The modification installs an isolation valve and calibration tee in each of the sensing lines for the accumulator level indication system. This allows for calibration of the level element /, transmitter without isolating the associated accumulator pressure transmitter.
Summary of Safety Evaluation: The modification constituted a potential change to the facility or its operation as described in the FSAR. The original sensing line is 3/8" tubing and was part of the accumulator pressure boundary. Original configurations on the sensing line employ tubing fittings and valves.
77
3 Y
- Pressure and-temperature' requirements of the'systen were much less than the' design ratings of the fittings and.
valves. Therefore,'use of tubing tees and valves for.
these modifications'is-justified. Material types were.
compatible with original types. Valve positioning
. ill be administrative 1y controlled.
w The. accumulator' level indication system is Seismic Class l'.
Installation was' based upon TMI tubing installation criteria for seismic qualification.
The change does not affect the functionality of-the system.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction-of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as, defined in the Technical Specifications is not reduced.
3.4.40 88-110, Radwaste System.
Summary of Safety Evaluation: The modification.
constituted a change to'the facility or its operation as-described in the FSAR..The FSAR provides a description of the blowdown evaporator in Section 11.1, but.this modification does not change this description or.the role described therein. The modification was designed per the requirements of the code of record and the specification requirements.
The removal of this line does not affect the operation of the blowdown evaporator and the capped drain can be used with temporary hoses for drainage and sampling when the BDE is' shut down, cooled down and deptessurized to-within the capacity of the hose to be used. The existing valve on the drain line and the capped connection (threaded) provides the necessary leakage protection. This new configuration returns the unit to the originally designed configuration.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the p
possibility for an accident or malfunction which has not l
been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
i 1
3.4.41 88-124, C01 SI Block. Switch Replacement.
Summary of Safety Evaluation: This modification replaces the existing SI block switch with two train-specific switches to meet the intent of IEEE-384 for train sepera-tion. This change requires mounting a new selector switch on C01 and moving the contact blocks from the existing SI blocks switch to the newfone. The wiring design was not changed; however, as found wiring was upgrade to safeguards system design practices. An evaluation is required as the change will affect the function of a system, structure, or component as described in the FSAR.
The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for.an accident or malfunction which has not been previously evaluated. The margin of safety as,
defined in the Technical Specifications is not reduced.
3.4.42 88-132 (Unit 2), Secondary Piping.
Summary of Safety Evaluation: The proposed MR constituted a change to the facility as described in the FSAR. The only significant change to the drain lines was in the piping class designation (from EB-6 to ED-2).
The 'ED' class has the same pressure and temperature ratings as the existing 'EB' class. The replacement pipe was the same size and schedule as the existing, and was routed in the same field locations. Since the sizes and weights of piping did not change, the existing supports and support locations were reused.
Stainless steel has a larger coefficient of thermal expansion than carbon steel, but existing routing and support configurations will allow the additional thermal growth without creating excessive additional stresses.
The replacement did not change the functionality of the system nor affect any of its failure modes. This material change has no impact on the plant high energy line break analysis.
i The change does not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
79 l
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F ys 9
B 3.5'
3.5.1 88-002 (Unit 1) and 88-003 (Unit 2), 4160 V Switchgear.
The temporary modification installed jumper wires on the auxiliary switch.within the A05/A06 bus tie breaker cubicles.
(Installed 01-19-8S; restored 06-25-88.)
Summary of Safety Evaluation: The temporary modification constituted.a potential change to-the
. facility or its operation as described in the FSAR.
Due to the potential for a single, failure in the auxiliary switch of the A05/A06 bus tie breaker, the
~
possibility exists'for disabling both diesel generator output breaker closures. As an interim measure, it is recommended that the'A05/A06 bus tie _ breaker _beLracked out and the auxiliary switch contacts-that are in the diesel breaker closing scheme (contacts 13-14 and 17-18) be jumpered to ensure continuity even under a single failure condition. This will ensure that the single failure will not affect the diesel breakers,.
provided the A05/A06 bus tie breaker is kept racked out by red tag or administrative procedures.
If the A05/A06 bus tie breaker is to be racked in for.
emergency use or for maintenance purposes, then it will
=be required that the jumper wires be-removed and cell-switch contacts be verified closed by electrical continuity testing to ensure diesel and safety-related power reliability.
The FSAR states that-the breaker is to be used for I
i maintenance purposes only. This change does not change-the safety analysis report description.
Strict administrative controls will be applied to ensure the breaker is not closed onto the safeguards buses without the interlock being operable (i.e., removal of the temporary modification).
The change did not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.5.2 88-016 (Common), Plant Process Computer System (PPCS).
The temporary modification installed four jumpers from the PPCS input MUX to record Bank D position signals during rod drops using the programmable chart recorder.
(Installed 04-04-88; restored 05-20-88.)
80
j
~
't Summary of Safety Evaluation: The temporary modification is a potential change to the facility or I
its operation as described in the FSAR. During the installation /use of the jumpers, the worst possible j
occurrence is the grounding of the MUX input signals.
l This will cause the computer-indice,ted rod position to
]
be zero inches (fully inserted).
In Technical Specification Table 15.3'.5-5, Item 6, control rod misalignment as monitored by the on-line computer is an " instrument operating condition for indication." Thus, if an indicated rod position went to zero inches, the computer would alarm. Remedial action is then specified by Tatle 15.3.5-5.
It requires the, " logging of rod position once/ hour, after a load change of >10% or after >30 inches of control rod motion." In actuality, the grounding would be momentary and so the remedial action would not be-required.
Thus, since the grounding of the MUX input signal would not inhibit the operator's ability to determine rod position from the RPI board indication and the loss of computer indicated rod position, remedial action is already specified in Table 15.3.5-5, and no margin of -
safety as specified in the Technical Specifications.
Hookup of the recording instrument will be controlled via MWR work plan. RPI runback will be temporarily
(<20 minutes) bypassed when making connections to eliminate the potential for a spurious runback if a lead is accidentally shorted. Bypassing RPI runback is acceptable because dropped rod runback protection will still be provided by NIS..
1 The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment j
important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.5.3 88-018 (Unit 2), Feedwater System. This temporary modification clamped valve 2MOV-2189 ("B" cteam generator feed pump discharge MOV) in the closed position. The valve disc and stem separated, as evidenced by open indication without flow. Therefore, valve position could not be confirmed.
(Installed 04-08-88; restored 04-10-88.)
81
- - --. ;, 7 -
Summary of Safety Evaluation: The temporary modification constituted a change to the facility as described in the FSAR. 'To regain control of the valve in the closed position, it is necessary.to drive the stem against the disc and the disc against the seat; then clamp it in that configuration. The B steam generator feed pump can then be secured and Unit 2 reduced in power in order to make repairs.
Although this valve shuts upon a safety injection, it serves no safeguards feature.
It does, however, provide backup feedwater isolation.~ It receives a shut signal when the steam generator feed pump receives a safety injection signal and its breaker opens.
Clamping this valve in the closed position, however, will ensure there is no opposite train recirculation if the B discharge check valve fails, thereby providing backup feedwater isolation.
With the valve clamped, the unit will be limited to. one feed train. Power level will be administratively controlled within the capacity of one train.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of-equipment important to safety is not increased. The change does not create the possiblity for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduceo.
3.5.4 88-024,88-025 and 88-026, Circulating Water Temporary Chlorination System. The temporary modification added hypochlorite to the circulating water pump suction by installing a piping and hose arrangement from the sodium hypochlorite tank located on the north end of the circulating water pumphouse. There are three temporary modifications involved in this process; a PVC fill line for the hypochlorite tank; a PVC fill line for the bisulfite tank; and a connection to the Nos. 2 and 3 waterboxes.
Both the hypochlorite and bisulfite systems are composed of PVC, polypropylene and rubber hose which are compatible with the chemicals used. These are inert materials and will not affect any existing plant components.
The temporary modification added a very small electrical load (120 V AC) to a non-vital bus to run the flow meters (only a few watts). These will only be run during chlorination (one hour per day).
82
+
l i
The PVC fill line (2") for the hypochlorite tank'is.
attached to the fish hatchery pipes. The weight of the PVC pipe, clamps.and even hypochlorite when the
{
tanks are being filled will be negligible for these large pipes. The PVC fill line for the bisulfite tank has been attached to a foundation beam. The combined weight of the PVC pipe, clamps and bisulfite when-the.
tank is being filled.will be negligible for this support beam.
Both sodium hypochlorite and sodium bisulfite are somewhat hazardous. Both have a rather strong odor
.which will increase with increasing temperature. During chlorination, normal ventilation will maintain the concentrations very low.
In the case of a spill, respirators would be required for those personnel working near.either chemical. Normal ventilation should clear the fumes in a short period of time. The CHES system sheet will be posted by both tanks should a question arise. Eyewash stations have been placed in the area, along with.all required safety equipment.
Flow meters are used for both the hypochlorite and bisulfite injection.
Instrument scales are 0-10 gpm,.
which are more than adequate for the flow rates which are required.
The only portion of the hypochlorite system which could fail is the tank itself; either at the outlet PVC pipe connection or a tank rupture. The bisulfite system can fail by either a tank rupture (similar to the hypochlorite system) or a break in the hose between the tank and the waterboxes. The PVC pipe connection could be broken off as a result of a vehicle or even a person applying pressure to the pipe. The bisulfite hose could be broken by applied pressure or could be disconnected from the tank or the waterbox.
I If a hypochlorite system failure occurred, the hypochlorite spilled in the circulating water pump pit would be contained while that portion passing through the manhole covers would enter the circulating water, and to a lesser extent, service water and (if the pump l
was running at the time) through the jockey pump into I
the fire water system.
j '
The result of a hypochlorite spill will be high chlorine levels in the circulating water for a short period of time which will not cause any long-term damage to the condenser tubes.
Service water and fire water will also experience high chlorine levels for a short period of time, but components will not be adversely affected.
83
n y.
If condenser tube leaks are present, chlorine will Lenter the secondary system. This excursion may cause
.us to enter Action Level l'.
PBNP 8.4.1' allows one week-for recovery from this situation. 'No significant steam generator damage will occur. A large portion of-the
~
hypochlorite.will be released to the~1ake'via the.
circulating water system. We would violate our DNR-permit' limit for total residual' chlorine (0.2 ppm).
Levels as high as 1000 ppm could be. released for a:
short period of time.
Due to the size of the circulating water pumphouse floor, 1500' gallons'(a fullLtank) would not be enough.
to endanger any electrical.or safety-related systems-due.to flooding. Tanker. trucks,.which deliver both chemicals, will not contain amounts greater than the capacity of:the storage tank.
If a fire occurred in the service water pump area coincident with a hypochlorite tank rupture,'the fire
~ brigade would have to use respiratory protection due to the chlorine gas which would be present. Since the fire brigade has SCBA with them when called upon, this should not present a problem. Due to the limited amount of hypochlorite in the tank and the physical separation from the control' room, chlorine will not enter the control room atmosphere.
A rupture of the bisulfite tank could send up to 1500 gallons of bisulfite into the turbine building sumps and condenser pump pit.
Sulfur dioxide fumes would result from the spill. This may~ affect the air in the control' room. Control room ventilation could be placed in a different mode to maintain air purity by using outside air.
If the bisulfite hose was broken or disconnected, air would be pulled into the waterboxes.
If the waterbox priming system could not keep up with the air inleakage, the control room would receive a waterbox low level alarm.
If no action was taken, condenser vacuum would decrease and a power reduction.or trip would occur.
(Note this is only possible during the one hour per day when chlorination is in progress.
During all other times, the waterbox isolation valve will'be maintained closed per procedure CAMP-004.)
Warning signs are used to make personnel aware of the presence of concentrated chemicals. Safety equipment is used during the off-loading of chemicals and safety equipment is kept in the area l3 at all times. Portable eyewash stations have been placed in the area.
84
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c i
b l
y h
j The changes does not pose an unreviewed' safety 1
question..The probabilityLof-occurrence or the
. consequences of an accident or. malfunction of equipment 1.:
.'important'to safety is.not increased. The change does R
l-not create.the possibility for an accident or.
malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications'is not reduced.
(Installed 04-22-88; restored 12-01-88.):
3.5.5 88-030 (Common), Health Physics. :The temporary 4
modification installed three (3) lead-shielded frisking j
booths to meet an NRC commitment. 'This is being done as a temporary measure until permanent, portable frisking l
booths can be designed and built. The proposed locations for these boot'.s are the Unit 1 and Unit.2 El. 66' fan
>{
rooms, and the spent fuel pit general area.
(Installed j
04-26-88.)
i Summary of Safety Evaluation: The temporary l
modification constituted a change to the
)
facility or its operation as described in the FSAR.
- j Each of the proposed locations contains seismically 1
anchored and/or supported equipment and/or piping. The
)
booths must not affect the qualified equipment and j
p2 ping.
l i
Due to the temporary nature of this installation (lead
)
blankets hung on metal' scaffolding.with nylon cable ties), it would be quite difficult to qualify the booths j
seismically. For this reason, the booths.have been j
located away from seismic equipment / piping and are j
chained to nearby structural steel members. These l
members will ensure that the booths will not fall onto, l
5 or strike up against, any qualified equipment or. piping' during a DBE.
~
The booths do not exceed floor loading limits, per the structural design criteria. They also do not pose a concern from a transient combustibles standpoint ~
(r ference PBNP 3.4.13), and they will not interfere with existing fire lanes.
The temporary modification does not pose an unreviewed safety question. The probability of occurrence or the 1
consequences of an accident or malfunction of equipment important to safety is not increased. The temporary modification does not create'the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
85
3.5.6 88-035, Fire Detection System. The temporary modification installed a temporary power supply to fire detection master control panel D400 to support work controlled via SMP #880.
(Installed 05-06-88; restored 05-20-88.)
Summary of Safety Evaluation: The temporary modification constituted a change to the facility or its operation as described in the FSAR. Neither D400 nor the power supply to'D400 is described in the FSAR.
The FPER references the " master control panel" but does not make reference to the specific power supply.
Technical Specification 15.3.14, Paragraph D, indicates that fire detection instrumentation for each zone shown in Table 15.3.14-1 shall be operable. The battery backup will provide eight (8) hours of operation for D400. SMP #880 requires the temporary power aupply be installed before the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are expired. Work on D400 is controlled by SMP #880.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.5.7 88-037 (common), Water Treatment System. The temporary modification installed two 2\\" hoses from valve WT-200A through the Unit 2 turbine hall into the Unit 2 seal well to provide a path for high flow rates (>250 gpm) which was required for startup testing of the new water treatment retreatment system.
(Installed 05-21-88; restored 10-07-88.)
Summary of Safety Evaluation: The temporary modification constituted a change to the facility or its operation as described in the FSAR. The normal discharge for this water, used for makeup water treatment startup tests, is through the retention pond.
l The retention pond discharge pipe is unable to handle I
the required flow.
By discharging directly to the seal well (where the retention pond also discharges), the water enters the lake at the same point. This is upstream of the normal sample point for outfall 002, approved by the DNR as a plant discharge point.
I l
i l
l 86 l
No automatic monitoring will be provided. This is acceptable in view of service and is similar to other discharge paths without automatic monitoring, e.g., _
circulating water.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.5.8 88-043 (Common), Radiation Monitorir.g System. The temporary modification installed a 2" strainer and fittings in the supply line to the RE-230 pallet in parallel with the existing 1" strainer.
(Installed 06-06-88; restored 06-27-88.)
Summary of Safety Evaluation: The temporary modification constituted a change to the facility as described in the FSAR. The temporary supply is non-seismic. The possibility of an accidental radioactive release is not significant because the retention pond effluent is normally not radioactive.
Recent samples have shown less than detectable contamination.
In the past, the only radioactive isotope present has been tritium.
The possibility of leakage from the hose connections and/or failure are minimal because:
a.
Good quality hardware of proper size, along with good quality hose, was employed using accepted techniques to implement the temporary modification, b.
The overboard valve is constructed so it is always open, thereby preventing the hose and fitting from experiencing a P55A&B deadhead pressure of-63 feet. Normal neutralizing tank releases are about 100 gpm, which give a pressure of 40 feed less line loss. This pressure is easily accommodated by the construction employed (the hose is rated for 315 psig).
Assurance of flow to RE-230 is provided by the low flow alarm associated with RE-230 itself.
If for some reason this feature failed, the location of RE-230 near water treatment ensures frequent observation by operating personnel.
87
The change-does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not' increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.5.9 88-049 (Unit 1), Feedwater System. The temporary modification accomplished a temporary repair of valve ICS-466AA. The proposed fix for this valve includes seal welding the hinge pin bonnet flange to the valve body; drilling a vent hole in the bonnet flange, and welding pipe caps over the studs / nuts.
(Installed 07-12-88.)
Summary of Safety Evaluation: The evolution
/
constituted a change to the facility or its I
operation as described in the FSAR. Additional stresses due to changing the pressure boundary are addressed in Calculation #P-88-022. Material strengths and inservice (and design) pressure are compared to Code allowables in-the calculation. The calculation / evaluation does not reveal a condition in which the integrity of the valve is endangered.
The additional weight on the valve due to the pipe caps and vent plug is insignificant and is not considered to impact the seismic capability. The volume created between the seal weld and pin gasket will form a small vessel, but pressure excursions due to a heatup of the trapped volume should be eliminated by leakage into the main valve. Loss of normal feed and high energy pipe failure are both considered in the FSAR.
The extension in the cover pressure boundary as a result of the seal weld raises the bolt stress to above that which is allowable for pressure loads per Section III and Section VIII. The valve was not l
designed and built to these Codes so they are being used as more of a reference than as absolute limits.
l Sections III and VIII limits for the bolting is 25,000 psi.
Calculated stress at a design pressure of 1310 psig is 42,730 psi. The specified minimum yield is 105,000 psi.
i 88
If some credit is taken for the seal weld, it can be shown that the bolting pressure stress would be within allowable.
Seal welds are not typically considered as providing any structural strength, however, and a portion of the veld will have to be made wi.th water present.
An adequate margin of safety is considered to be maintained for this temporary repair based upon the following:
a.
The design pressure of 1310 psig is not expected to be seen.
b.
An adequate margin to yield (2.5) is still available.
c.
The seal weld will provide some strength.
d.
Section III allows going to two times the allowable for special service conditions.
The bolting has already been torqued to 60,000 psi, e.
so in actuality, the change in pressure boundary will not increase the stud loads from what they are noW.
The repair does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident of malfunction of equipment important to safety is not increased. The repair does not create the possibility for an accident or malfunction which has not been previously Ivaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.5.10 88-051, Feedwater System. The temporary modification accomplished a temporary repair of valve ICS-476AA The fix for this valve included seal welding the hinge pin bonnet flange to the valve body; drilling a vent hole in the bonnet flange, and vtiding pipe caps over the studs / nuts.
(Installed 07-14-88.)
The evolution constituted a change to the facility or its operation as described in the FSAR. Additional stresses due to changing the pressure boundary are addressed in Calculation #P-88-022.
Material strengths and inservice (and design) pressure are compared to Code allowables in the calculation.
The calculation /
evaluation does not reveal a cor.dition in which the integrity of the valve is endangered.
89
I The additional weight on the valve due to the pipe caps and vent plug is insignificant and is not considered to impact the seismic capability. The volume created between the seal weld and pin gasket will form a small vessel, but pressure excursions due to a heatup of the trapped volume should be eliminated by leakage into the main valve. Loss of normal feed and high energy pipe failure are both considered in the FSAR.
The extension in the cover pressure boundary as a result of the seal weld raises the bolt stress to above that which is allowable for pressure loads per Section III and Section VIII. The valve was not designed and built to these codes so they are being used as more of a reference than as absolute limits.
Sections III and VIII limits for the bolting is 25,000 psi calculated stress at a design pressure of 1310 psig is 42,730 psi. The specified minimum yield is 105,000 psi.
If some credit is taken for the seal weld, it can be shown the bolting pressure stress would be within allowable. Seal welds are not typically considered as providing any structural strength, however, and a portion of the weld will have to be made with water present.
An adequate margin of safety is considered to be maintained for this temporary repair based upon the following:
The design pressure of 1310 psig is not expected to a.
be seen.
b.
An adequate margin to yield (2.5) is still available.
c.
The seal weld will provide some strength.
d.
Section III allows going to two times the allowable for special service conditions.
The bolting will have been torqued to 60,000 psi in e.
an attempt to stop the leak before performing the temporary modification.
The repair does not pose an unreviewed safety l
question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The repair does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
90
w 3.5.11-88-055, Radwaste Steam System. The temporary.
modification takes condensate-samples from the blowdown evaporator via valve RS-25 immediatelyLfollowing installation of the new reboiler. 'The. condensate samples must'be taken.to check if contamination levels are low lenough so water can be sent to the condenser. A tee fitting was installed from valve-RS-25 so a hose-could be run to a floor drain inside the blowdown evaporator building with a second hose being run to a water sampler.
(InstalledLO9-02-88; restored 09-09-88.)'
Summary of Safety Evaluation: The temporary' modification constituted a change to the facility or its; operation as described in-the FSAR.
The water sampler was used specifically as a heat exchanger to cool the condensate to a temperature at which samples can be handled. 'This effort was a precautionary measure, as the condensate should be radiologically clean. Wasting and sampling the condensate'from the startup of the new reboiler r
also ensured that chemical contaminants (manufacturing and installation) do not' result in an operating unit chemistry excursion.
Coolant water was supplied from the west service water header and no electricity was necessary to operate the sampler.
l The blowdown was. initially routed to the blowdown evaporator building floor drain. Cooling water',
blowdown (when it is determined that no activity exists in the blowdown) and samples were routed to floor drains outside of the blowdown evaporator building. It should be noted that the blowdown evaporator drain goes
.to the sump tank and the facade drain goes to the facade sump. Both resulting paths are monitored or processed prior to release to the environment.
3.5.12 88-058 (Common), Fuel Oil System. The temporary modification removed check valve FO-0023 from the fuel oil fill line. Manual isolation valves FO-0022 were red tagged shut and administratively controlled via SMP #919.
(Installed 09-22-88; restored 09-28-88.)
Summary of Safety Evaluation: The temporary E
modification constituted a change to the facility or its operation as described in the FSAR. If the line would have been left open while the check valve was removed, the vacuum breaker on the tank would have prevented any significant amount of fuel oil leaking from the tank.
The valve was reinstalled after one tank has been pumped out for inspection. The red tag control was designed to prevent inadvertent spillage of fuel oil.
91
__--_______2
The changeLdoes not pose an_unreviewed safety.
question. The probability'of occurrence or the
?
consequences.of an accident or malfunction of equipment important.to safety is not increased. The change'does not createLthe possibility.for'an accident or malfunction which has not been previously evaluated.
.The margin of safety as defined in the Technical Specifications'is not reduced.
3.5.13 88-061 (Unit 2), Plant Process Computer System. Eleven inputs'to the plant process computer system-(PPCS) wereH
. paralleled off the MUX terminal-strips to provide
. signals to a fatigue monitoring minicomputer.
(Installed 10-27-88,)
The following input signals will be'used:
FT-414,.RCS flow rate; LT-428,~ pressurizer water. level; N41, nuclear power; PCV-431A&B, pressurizer spray control valve position; PT-420B&C, RCS wide range pressure
.(flow taps); TE-421, pressurizer water temperature;.
TE-424,. pressurizer surge line water. temperature;.
'TE-451B&D, RCS hot leg wide range temperature.
Summary of Safety Evaluation: The temporary modification
~
represented a change to the facility or:its operation as described in the FSAR. The analog inputs from the MUX.
are isolated from the corresponding instrument loops;
.therefore, installation of twisted shielded pairs (hereafter jumpers) will not affect operation. of the '
reactor control and protection system. However, during installation and ure. of the jumpers, if the MUX input voltage signals are shorted or if input impedance of the minicomputer drops to a low level, the voltage signals to the PPCS would be reduced significantly. This reduces the value of the channels / parameters indicated by the PPCs and further affects the subcooling margin calculation and reactor thermal output (RTO) calculation.
Technical Specification 15.6.3.5-5, Item 4, requires that the RCS subcooling monitor or backup be functional. The PPCS and SAS are the relied upon backup displays of subcooling margin. The RTO calculation is used by operations to precisely control reactor power, although it is not included in the Technical Specifications.
Grounding or shorting of a MUX analog input signal would cause the indication in the PPCS to go to "O."
The computer would alarm in this condition. Typically, shorting or grounding would occur during the installation process. Hookup of the jumpers were controlled via a maintenance work request work plan to eliminate the potential for spurious signals.
92
]
t
(
An' isolation amplifier with 8 megohm input' impedance-
.(both on and off) was utilized to separate;the PPCS and Westinghouse fatigue monitoring' minicomputer. With an 8 megohm impedance, the_ MUX input signals to the PPCS
- were unchanged, and hence, the'PPCS indications, subcooling margin indication, and reactor thermal output calculation were not affected. In case problems occur i
with PPCS displays, control board' displays can be utilized j
to control plant operation.
The heat load associated with:the fatigue monitoring-minicomputer was reviewed to~ ascertain its impact on the environment in the north computer room.
It was concluded '
that air conditioning capacity was. sufficient'and that
~!
the ' diffusers will satisfactorily.' circulate the air inL
.)
the north computer room even with the installation of 1
the minicomputer. In:the area between the Spec-200 i
racks'and adjacent to the. auxiliary racks (where the minicomputer will be' located),.there'are two diffusers j
where air flow is throttled. Additional air flow circulation can be obtained by opening.the dampers,'if l
needed. Note, since the review was performed, it has j
been determined that the-vendor. mistakenly double-.
1 counted some heat loads'from their minicomputer. These n
loads are reduced from 14,785 BTU /hr to 10,275 BTU /hr, which.further ensures adequate installed air conditioning
.l capacity.
]
The last item considered was theieffect of the fatigue:
L monitoring'minic'omputer.on equipment during a seismic.
l event.
Basically,.the minicomputer is contained in a i
2' high x 2' wide x 6' long box and is otherwise very j
similar.to a PC with peripherals. The fatigue monitoring l
minicomputer will be installed on the floor with the i
keyboard, CRT and printer / plotter.on a table above it.
This is like current PC and equipment installations in
.the computer space and poses no problem.
Power panel 1YO7 is stripped via IB31. Therefore, the additional load on 1YO76 does not result in additional diesel generator loading.
The change does not pose an unreviewed safety j
question. The probability of occurrence or the j
consequences of an accident or malfunction of equipment important to safety is not increased. The change does l
not create the possibility for an accident or malfunction which has not been previously evaluated.
l The margin of safety as defined in the Technical Specifications is not reduced.
l 1
l f
93 l
-i I
i 3.5.14 88-067 (Unit 2),88-068 (Unit 2) and 88-076 (Unit 2),
1 Primary and Auxiliary Systems. The temporary l
modifications installed a total of 44 temperature q
sensors (K-TCs); 25 on the pressurizer surge line;;11 1
on the auxiliary charging line; and 8 on the auxiliary spray line. This necessitated the installation of 44 thermocouple cables in containment to one box of remote multiplexer; require the temporary use of 117 V AC power in containment for the multiplexer; the temporary use of 4 spare twisted pair cables and 5 other wires from containment to the north computer room; and installation of a computer, terminal, printer, and 9600 baud modem in the north computer room, with parallel connection and use of the 11 inputs being used by the transient and fatigue monitoring pilot project (TM-88-061) via taps made downstream of the isolation amps. This will require the temporary use of 117 V AC power and one tie-line telephone connection in the north computer rocm.
This temporary modification is required to fulfill-specific piping temperature monitoring requirements contained in NRC Bulletin 88-08.
(Installed 11-19-88.)
Summary of Safety Evaluation: The temporary modifications constituted a change to the facility as described in the FSAR. Eleven inputs to the plant process computer system (PPCS) will be paralleled off the MUX terminal strips to provide signals to the fatigue monitoring minicomputer. The CompuDos computer (data acquisition system) being used for this project will also parallel off the 11 PPCS inputs.
The input signals utilized are:
2FT-414, RCS flow rate; 2LT-428, pressurizer water level; 2N41, nuclear power range detector; 2PCV-431A and 2PCV-431B, pressurizer spray control valve position; 2PT-420B and 2PT-420C, RCS wide range pressure flow taps; 2TE-421, pressurizer water temperature; 2RE-424, pressurizer surge line water temperature; and 2TE-451B and 2TE-451D, RCS hot leg wide range temperature.
The analog inputs from the MUX are isolated from the corresponding instrument loops. Therefore, installation of twisted shielded pairs (hereaf ter called jumpers) will not affect operation of the reactor control and protection system. However, during installation and use of the jumpers, if the MUX input voltage signals are shorted or if input impedance of the minicomputer and the CompuDos computer drops to a low level, the voltage signals to the PPCS would be significantly reduced. This reduces the value of the channels / parameters indicated by the PPCS and further 94
affects the-subcooling margin calculation and' reactor
~ thermal output (RTO)' calculation. ~ Care will be taken to:
inform the Operations staff when input termination work will start ~and finish. Note that these connections will~
be'made downsteam of the' isolation amplifiers.
Technical Specification-Table 3 5.3.5-4, Item 4, requires that the RCS subcooling monitor or backup be functional. The PPCS'and SAS'are thefrelied'upon backup displays of subcooling margin.'.The RTO-
. calculation is used by Operations to precisely control reactor power, although'it is not included in the Technical Specifications.
Grounding or shorting of a MUX analog input signal would cause the indication to the PPCS to go to zero (0)..The computer would alarm in this condition.
Typically, shorting or grounding would occur during the installation process. Hookup of t the jumpers will be controlled via a maintenance work request work plan.to eliminate the potential for spurious signals.
An isolation amplifier with 8 megohm inputLimpedance (both on and off) was utilized-to separate the PPCS and-fatigue monitoring minicomputer. The input of the CompuDos computer has 10 megohm inputs-(on and off).
With'an'8 megohm impedance, the MUX input' signals.to the PPCS were unchanged, and hence, the PPCS indications, subcooling margin indication, and RTO calculation were not.affected.
In the event problems occur with PPCS 7
' displays, control board displays can'be utilized in controlling plant operation.
Additional considerations are as follows:
During accident a.
Blockage of Containment Sump.
conditions there is a possibility of ordinary cable ties coming loose and accumulating on the Sump "B" intake structures. For this installation, blue-colored cable ties left over from TMI projects or stainless steel cable ties will l>e used. The blue cable ties are made of Tefzel, a teflon material capable of withstanding high temperatures.
b.
Impact on Existing Seismic Structures. The thermocouple cables will be tied off to existing structures in the containment such as grating, pipe supports and conduit supports. The mass of the cables is small compared to the mass of the pipe or conduit already supported so this small increase in weight will not affect the seismic qualification of l
l 95 b
L.-..-__.--__.___:_-_________________
I the pipe or conduit supports. Any force from the thermocouple cables to the supports would be transmitted through the cable ties and if the force becomes significant, the ties would break.
c.
Concrete Anchor Bolts.
Concrete anchor bolts will be used to attach the multiplexer to the El. 46' floor level in containment. Some concrete anchor bolts may also be used to support the thermocouple cables. These concrete anchor bolts will be installed in accordance with Maintenance Instruction 7.1.
This will ensure that they will not interfere with existing concrete anchor bolts or
- rebar, d.
Radiation Levels in the Regenerative Heat Exchanger Room.
It was initially planned to install three thermocouple at two locations on the auxiliary spray line in the regenerative heat exchanger room. Due to the high radiation levels in this, j
room, the locations were reduced to one; the number of thermocouple. reduced to two; and the method of mounting the thermocouple changed to reduce the amount of insulation to be temoved. Nondestructive testing of the pipe in the area was determined not to be necessary at this time. These deletions and changes significantly reduced the radiation exposure expected to be received by the people doing the installation.
The temporary modifications do not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The temporary modifications do not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.5.15 88-074, Water Treatment System. The temporary modification allows water from the condensate storage tank (CST) to be supplied to the 3" DI water header.
Summary of Safety Evaluation: The temporary modification constituted a change to the facility or its operation as described in the FSAR.
The temporary modification was evaluated with respect to interaction with the safety-related function of the CSTs. Since the CSTs serve as the source of emergency auxiliary feedwater, Technical Specification 15.3.4. A.3 requires that a minimum reserve of 10,000 gallons per operating unit be available in the CSTs.
96
A conceptual' evaluation of!this temporary modification demonstrated the worst case situation'would result from the failureLof the temporary pump suction line. This-temporary connection is made at_the CST drain and failure would result in;draindown of a CST under its own hydraulic head.
An evaluation has been performed assuming a worst case CST configuration. The configuration evaluated was for a single CST in service as the' source of auxiliary feedwater for two operating units. ' Calculations demonstrate that for an instantaneous failure of the 2" temporary connection,'approximately 22 minutes is available from the CST low level alarm (10' relative elevation) to a level corresponding to 20,000 gallons 1 of feedwater reserve. Consequently, approximately one-third hour is available for operator intervention.
to prevent violation of,the applicable Technical ~
Specification. The draindown can be secured by closing the CST drain valve. No other mechanism;related-to-installation, operation or removal of this temporary modification has been identified as a potential safety issue under'the terms of 10 CFR 50.59.
Safety-related equipment will not be impacted by a.
failure of the pump hoses.(due to flooding).
Flow rates from a hose failure would be within the capability of existing' floor drains.
Operation of the pump will be controlled,such that suction flow will be supplied from the water treatment trailers rather~than from the CSTs. This will
-eliminate the potential for introducing contaminants from the CSTs to the RCS.
The change does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.is not increased. The change does not create the possibility for an accident or malfunction which has not been previously evaluated.
The margin of safety as defined in the Technical Specifications is not reduced.
3.5.16 88-077 (Common), Water Treatment System. The temporary modification routes hose (s) from the water retreatment plant at valve WT-200A and connects them to the portable water treatment equipment. The effluent from the equipment was routed to the condensate storage tanks via valve AF-13.
(Installed 10-07-88; restored 10-22-88.)
97 1
lL___--_--
Summary of Safety Evaluation: 'The temporary modification constituted a change to the facility or its operation as described in the FSAR. An accidental loss of two hoses to the condensate storage tank (CST) provides an open drain path for the CSTs.
Calculations show that with one CST in service, approximately 7 minutes elapse'from the time the low level alarm is received until the level required by the Technical Specifications (10,000 gallons /
unit) is reached. This is adequate time for an operator to close the drain valve and isolate the leak path.
A loss of the hoses in the turbine hall will not flood any safety-related equipment. The water will drain via normal drain paths.
The effluent from temporary water treatment is sampled periodically and is analyzed by Chemistry. This should minimize or prevent contamination of the CSTs by the temporary water treatment.
Hydrazine is used by temporary water treatment for oxygen control. The hydrazine, stored outside or in the Unit 2 turbine hall, will not be near any safety-related equipment. Normal plant procedures for storage and handling of hydrazine should be followed to prevent personnel injury.
Propane is used by temporary water treatment for heating the trailers. The trailer locations, east of the emergency diesel generator exhaust dampers,.are outside and have good ventilation from all sides. The trailers are an adequate distance from the hydrogen storage area also. Normal plant procedures for transient combustible control will be followed. Smoking will be prohibited near the propune tanks, piping and controls.
The hose to the CST should be routed at least 10' from the controls on El. 8' for the emergency diesel generator fire dampers in order to minimize damage caused by hose leaks.
The temporary modification does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The temporary modification does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.5.17 88-078 (Unit 2), Heating, Ventilation and Air Conditioning. The temporary modification removes the feeders for 2W3A-M and 2W3B-M, control rod drive mechanism cooling fans, and install one 60 amp i
l l
98 I
?!F temporary supply to each fan's motor power supply to provide a. power supply for. steam generator sludge lancing equipment.
(Installed 10-29-88;. restored 11-12-88.)
Summary of Safety Evaluation: The evolution:
constituted a change to a system,' structure or component as described in the FSAR. Unit 2 will be in the cold shutdown condition. The W3 fans are used to cool the control rod drive mechanisms during operation. Since the mechanisms are=not operated during cold shutdown, the fans are unnecessary during a cold.
shutdown. This temporary modification was returned to the normal configuration and operationally. tested prior to the beginning of reactor coolant system heatup.
The temporary modification does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of. equipment
'important to safety is not increased. The temporary modification does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
3.5.18 88-086 (Common), Service Water System. The temporary modification routed service water to the controlled side drycleaners for cooling purposes.
(Installed 10-17-88; restored 10-26-88.)
Summary of Safety Evaluation: The temporary modification constituted a change to the facility or its operation as described in the FSAR. Flow rates required for this temporary modification were well within the capabilities of the service water system and should not cause any deleterious side effects.
Since an extremely remote possibility of the service water coming into contact with the contaminated cleaning solvent does exist, the return flow path for the service water was monitored. The drainage point for the return service water is the Unit 1 turbine building sump. This return flow path has monitoring capabilities via the l
retention pond discharge monitor.
The return hose was routed through the door between the controlled side and Unit 1 turbine hall (Door #260). The door is red locked, but it is not a security boundary. Health Physics re-red locked the i
L door after the hose was routed. The red rubber hose I
for this temporary modification has adequate temperature and pressure ratings to withstand normal service water operating conditions.
J i
1 99 f
_----._.n_.._
The. temporary configuration was non-seismic and--
non-QA..If a leak. developed, the line could have been isolated at the hose station and did not poseLany possibility for flooding the auxiliary building..The' isolation valve.was shut when the unit was not in operation.
The door in question is an industrial fire door, and therefore, a. fire watch was employed on one. side of.the door if it was to remain open for extended periods of time.
The temporary modification does not pose an unreviewed:
safety question. The probability of occurrence or the consequences of an accident of malfunction of equipment important to safety is not increased. The temporary modification does not create the possibility for an accident or malfunction which has not been'previously evaluated. The margin of safety as defined.in the Technical Specifications is not reduced.
3.5.19 88-087 (Unit 2),. Reactor Coolant System. The temporary modification placed a strongback across.the coupling of 2PIB reactor coolant pump to hold the pump and rotor-in the axial position.
(Installed 10-31-88;; restored 11-14-88.)
Summary of Safety Evaluation: An evaluation is required since this temporary modification altered a.
system, structure or component as described in the-FSAR and changed the faility or its operation as described in-
.the FSAR. The strongback was installed to allow removal of the reactor. coolant pump flywheel with the. motor' installed. The flywheel holds the thrust bearing runner in place; thus, a means of holding the pump rotating element and motor rotor in the axial position when removing the flywheel was required.
If the reactor coolant pump was not held in the axial position, the seal faces would open up, creating a leak. The strongback was designed with adequate margin to support the pump rotating assembly, motor rotor and flywheel. Should the strongback fail and the pump shaft drop, the leakage out through the pump seals would still be restricted. This leakage would be relatively small and would be detected by the seal leakoff alarm or containment sump alarm. Action could have then been taken to provide makeup to the reactor coolant system and to place a fuel assembly, if one was in the manipulator, back into the core.
If the shaft would drop, the shaft has a step machined on it that rests on a ledge in the thermal barrier.
This was designed into the pump to be a leak-limiting seal and is used by some plants to allow seal work with 1
100
7 p
6' l'
the cavity flooded. A leak from the pump-seal area would not have uncovered the reactor core,-and in fact.-
.would not' drain the reactor coolant' system to below the top of;the loops. Thus, residual-heat removal will:
l remain operable.
A temporary modification similar to this.was evaluated and performed previously via temporary modification 86-074.
I The temporary modification does not pose an unreviewed safety question. The probabili+.y of occurrence-or the consequences of an accident or. malfunction _of equipment
~
importa..t to safety is not increased. The temporary.
modification does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety:as defined in'the-Technical Specifications is not reduced.
3.5.20 88-089,- PACVS. The temporary modification added a cap to the_2" post-accident containment ventilation system (PACVS) inside of the-Unit'2 containment 2 The cap was added to facilitate testing required for modification 83-021,' Work Package B and SMP #917, and provided a barrier for containment as required by Section 2.3 of SMP #917. The installation was controlled by the-temporary modification and SMP #917.
(Installed 10-21-88; restored 11-15-88.)
Summary of Safety Evaluation:. The' temporary modification constituted a change to tlie facility or its operation as described in the FSAR.
The PACVS is described in the FSAR but the addition of the cap will only be done while the unit is'in cold-shutdown or refueling shutdown when the PACVS is not needed. The system was restored prior to containment integrity being established. The cap also served as a refueling containment barrier in place of valve 3 when the valve was being changed out'per modification 83-021 Work Package'B.~ The socket-welded cap met the design requirements of Class 151 piping.
The temporary modification does'not pose an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The temporary-modification does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.
101 l
- I 4
3.5.21 88-090, Fuel Manipulator..The temporary modification replaced the' existing auxiliary hoist on the. Unit.2 fuel manipulator crane with another similar hoist. nThe existing hoist does not work and replacement parts-cannot be obtained, so the use of the other hoist is needed..(Installed. 10-21-88; restored 11-15-88.)
Summary of Safety Evaluation: An evaluation is required because the temporary modification altered a L system, structure or component as described in the FSAR..The auxiliary hoist on the-fuel manipulator is described in Section 9.5.2 of the.PBNP FSAR. The description is as follows:
"A hoist is attached which~
is used for rod' latching tool, plug device tool, and other tools used in the refueling cavity."
- Both hoists are rated at 2000 pounds.- The new hoist weighs'approximately the same as the old hoist and has the'same trolley as the existing hoist. These two considerations assure that.the hoist will not fall-Jnto the reactor cavity during an earthquake. The new hoist used the same power supply as the existing hoist and was electrically. compatible. The new hoist functioned in a manner comparable with the existing hoist.
The temporary modification does not pose-an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of' equipment important to safety is not increased..The temporary modification does not create the possibility for.an accident or malfunction which has not been previously evaluated..The margin of safety as defined in the Technical Specifications is not reduced.
3.5.22 88-091, Service Water System. The temporary modification provided a service water supply to temporary water treatment via ISW-404'and an effluent from the temporary water treatment system to the condensate storage tanks (CSTs) via valve AF-13'(south CST drain valve).
(Installed 10-22-88;- restored 11-15-88.)
Summary of Safety Evaluation: An evaluation is required because the evolution constituted a change to the facility or its operation as described in the FSAR.
An accidental loss of two 2-1/2" hoses to the CST provides an open drain path for the CSTs. Calculations.
show that with one CST in service, approximately seven minutes would elapse from the time the low level alarm is received until the level required by Technical specifications (10,000 gallons per unit) is reached.
This is considered adequate time for an operator to close the drain valve to isolate the leak.
4 102
A loss of hoses in the turbine hall will not flood any safety-related equipment. The water will drain via normal drain paths.
l.
The effluent from temporary water treatment is sampled periodically and analyzed by Chemistry. This'should L
prevent or minimize contamination of the CSTs by temporary water treatment. A new truck is ordered when the next to last bed depletes on the demineralized trailer. This provides plenty of time for a replacement. truck to arrive.
Hydrazine is used by temporary water treatment for oxygen control. The hydrazine is stored outside and will not be near any safety-related equipment. Normal plant procedures for storage and handling of hydrazine should be followed to prevent personnel injury.
Propane is used by temporary water treatment for heating the trailers. The trailer location east of the emergency diesel generator exhaust dampers is outside and has good ventilation from all sides. The trailers are an adequate diitance from the hydrogen storage area. The area around the trailers is labeled "no smoking" near,the propane, piping and controls.
The hoses to the CST should be routed at least 10' from the controls'on El. 8' for the emergency diesel generator fire dampers to minimize damage due to hose leaks. 'A temporary water treatment operator is on duty 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day and maintains radio communications with the control room. This will prevent unnoticed problems with temporary water treatment affecting the plant. Unit I turbine hall isolation will occur if less than four service water pumps start upon a Unit-1 safety injection.
The temporary modification does not pose an unreviewcd safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The temporary modification does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined'in the Technical Specifications is not reduced.
3.5.23 88-092, Heating and Ventilation. The temporary j.
modification disabled the purge exhaust fan to steam generator channelhead blower interlock. The purpose of the interlock is to secure the channelhead blowers if the purge exhaust fans are not operating to prevent discharging iodine from the channelheads into the containment atmosphere.
(Installed 10-21-88; restored 10-22-88.)
103
_-_..__a
c__z,__----
- Summary of Safety Evaluation: An evaluation is-required because the interlock was committed to in j
response:to NRC open item 04 contained in inspection report 86-015. The interlock was completed under modification 86-115/116.
i Additionally, to improve the performance of the channelhead ventilation system, new blowers were installed with roughing filters, HEPA filters and charcoal bed adsorbers to reduce the activity levels discharged by the blowers.
An' evaluation performed on October 21, 1988, showed that without the purge system operating and the channelbead blowers running, the containment atmosphere would reach an equilibrium level of 64.6% MPC for iodine.
Removal of the purge exhaust interlock was discussed with the NRC, who concurred with the temporary removal of the interlock provided it is properly evaluated and documented.
During the time period that the interlock was inoperable, Health Physics increased surveillance of the containment atmosphere.
In addition to the continuously operating low volume air samplers located throughout containment, an additional sampler was installed near the purge exhaust duct where Maintenance was working and at the containment hatches. Health Physics also collected grab samples at the maintenance work area, at the valve, El. 668-and El. 8' areas and at the steam generator manway platforms.
These grab samples initially were collected once per hour.
Based upon containment activity levels, the frequency was relaxed by the Health Physicist, however, the minimum frequency was at least once every two hours.
Health Physics procedure HP 4.4, " Respiratory Protective Device Use Criteria," was used for guidance and action levels during the evolution.
Additionally, if containment iodine levels reached 50% of the iodine MPC, Health Physics evaluated the need for additional protective actions, including the need for securing the channelhead blowers. At 100% iodine MFC, the channelhead blowers would have been secured.
As discusseC in Section II of the detailed safety evaluation report, the installation of the jumper did not affect the operation of the purge exhaust system.
The temporary modification does not pose an unreviewed safety question. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased. The temporary 104
i-tz r.u m
modification does not create the possibility for:a'n i
~
accident'or malfunction which has not been previously.
' evaluated. : The. margin of safety as idefined in' the L
4 Technical SpecificationsLis.not reduced.
~
3.5.24
~88-097'(Common), Water Treatment System. The temporary modification routed hoses from the water
.' treatment' plant at, valve WT-316C-and connect them to the' temporary. water treatment (TWT) portable water treatment equipment. The effluent from the TWT equipment will be routed'to the condensate storage tanks, (CSTs) via valve AF-13..The hoses are' rated for-at-least 350 psi and the pressure inside the hoses will be r
less than 100 psi.
(Installed 11-15-88.)-
Summary of Safety Evaluation: The temporary.
modification constituted a change to,the. facility.or its-operation as described.in the FSAR. An accidental loss of two hoses to the. CST provides'anLopen drain path for.
the CSTs. Calculations show that with one CST in service, approximately 7 minutes elapse'from the-time-the low level alarm is' received until the level required by the Technical Specifications (10,000 gallons / unit) is reached..This is adequate time for an' operator to close
~
the drain valve and isolate the leak path.
A loss of the hoses in the turbine hall will'not' flood.
any safety-related equipment. The water will drain'via normal drain paths.
.The effluent from TWT is continuously monitored with in-line instrumentation'for purity.and is sampledL periodically and analyzed by Chemistry..This should minimize or prevent contamination of the CSTs by TWT water.
Hydrazine is used by TWT for oxygen control. The hydrazine, stored outside or in the Unit 2 turbine hall, will not be near any safety-related equipment.
Normal plant procedures _for storage and handling of hydrazine are adequate to prevent personnel injury.
Propane is used by TWT for heating the trailers. The trailers are an adequate distance from the hydrogen storage area. Normal plant procedures for transient combustible control will be followed. Smoking will.be prohibited near the propane tanks, piping and controls, and no smoking signs will be installed as necessary.
If the trucks are stored inside, the propane tanks will be isolated.
The hoses to the CST should be routed at least 10' from the controls on El. 8' for the emergency diesel generator fire dampers in order to minimize damage caused by hose leaks. A TWT operator is on duty 105
p 1
N 24' hours per day and maintains radio communications
.with the control room. This will prevent unnoticed-
-problems with TWT affecting the plant. Unit 2' 7
. turbine hall isolation sill occur if;1ess than four service water pumps start upon a. Unit 2-safety o
s injection.
The temporary modification does not pose an unreviewed safety question. The probability of occurrence or the consequences.of an accident or malfunction of equipment.
important to safety is not increased. 'The temporary.
modification does not create the possibility for an accident or malfunction which has not been previously evaluated. The margin of safety as defined in the Technical Specifications is not reduced.'
3.5.25-88-097-01, Water Treatment System.
In' addition to the items stated in temporary modification 88-097 and associated safety evaluation SER 88-133, an additional hose will be-run from WT-300C to the temporary wate,r-treatment (TWT): trailers and two. additional hoses will be routed from the effluent of the TWT trailers to the
- drain line.on the north condensate storage tank-(AF-14).
(Installed 11-15-88.)
Summary of Safety Evaluation: The calculation regarding the draining of-the CST still applies since.
no more than two hoses are attached to each CST.
A booster: pump will provide the additional head to support the higher flow rates. The pump will be supplied from a local welding receptacle.
All other provisions of originally approved temporary modification and safety evaluation report remain applicable to this addendum.
1 3.6 Core Reloads 3.6.1 Unit 2 Cycle 15 Reload, The Unit 2 Cycle 15 reload contains 12 fresh Region 17A optimized fuel assemblies (OFA) at 3.4 w/o, 16 fresh Region 17B 0FA at 3.8 w/o, 32 once-burned Region 16 0FA, 31 twice-burned Region 15 0FA, 24 thruce-burned Region 14.0FA, and 6 Region 9, 12B, and 13B standard design assemblies. The Cycle 15-reload is the fifth reload containing a full region of 0FA fuel for Unit 2.
Summary of Safety Evaluation: No additional Technical Specification changes are required beyond those already covering 0FA transition cores. No special environmental considerations are involved and, therefore, evaluation of environmental effects by the NRC staff and issuance of environmental impact or assessment statements are not involved.
106
-The Unit:2_ Cycle ~15 reload design and safety analysis is acceptable and falls within the scope of earlier-
. analyses discussed in the'OFA RTSR, and indicates that operation of the Cycle 15 core does not involve a significant increase in the probability or consequences of accidents previously considered,-does not' involve a significant decrease in safety margin, and'does not; involve a significant hazard consideration. Therefore, provided the startup physics testing does not result in any discrepancies with the analysis assumptions,: the operation of Cycle 15 is> acceptable based on its reload.
design.and safety analysis.
1 3.6.2 Unit 1 Cycle'16 Reload, The' Unit 1 Cycle 16' reload contains 12 fresh Region 18A optimized fuel assemblies.
(OFA) at 3.4 w/o, 16 fresh Region 18B 0FA at 3.8 w/o,-
32-once-burned Region 17 OFA, 28 twice-burned Region 16 0FA,~~20 thrice-burned Region 15 0FA, and 13 Region 10A,-
11, and 12B standard' design assemblies. The four Region 10A and 8 Region 11 assemblies were initially loaded into Unit 2. ~The Cycle 16 reload is.the fourth reload containing a full region of 0FA fuel for PBNP Unit 1.
Summary of Safety Evaluation: No additional' Technical Specification changes are. required beyond those already covering 0FA transition cores. No special environmental considerations are involved and, therefore, evaluation of environmental' effects by the NRC staff and issuance of environmental impact or assessment statements are not involved.
The Unit 1 Cycle 16 reload design _and safety analysis is acceptable and falls within the scope.of earlier analyses discussed in the OFA RTSR, and indicates'that operation of the Cycle 16 core does-not' involve a significant increase in the probability or consequences of accidents previously considered, does not involve a significant decrease in safety margin, and does not involve a significant hazard consideration. Therefore, provided that the startup physics testing does not result in any discrepancies with the analysis assumptions, the operation of Cycle 16 is acceptable based on its reload design and safety analysis.
107 a _ - - __ __ -.
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5.0 STEAM GENERATOR TUBE INSERVICE INSPECTION The=results of the findings from steam generator tube inspections are as'follows:
E 5.1 Unit-1 5.1.1 Inspection Plan During the Unit 1 Refueling 15 " outage, eddy current -
testing was performed from April 10,1 1988 to May 6,
. This inspection was in.two phases.as a result of 1988.
inadvertent damage to two tubes and a potential loose i
parts concern resulting from a secondary side modifica-tion.,The extent tested in each steam generatorLis as follows:
Eddy Current Inspection Plan Extent of Inspection Number of Tubes Inspected -
A SG B SG Hot' Leg (Cold Leg) Hot Leg (Cold Leg)'
Full Length 129
'112 U-Bend-18
- 1 TSP
- 179*
(178)'
1
_ (4)
'#2 TSP-1 BP (1)
TOTALS 308 (178).
132 (5)
BP - Baffle Plate TSP - Tube Support Plate U-Bend - Through U-Bend From Referenced Leg
- Phase 2 Inspections - For potential loose parts damage to peripheral tubes 5.1.2 Inspection Results The results of the initial sample showed no indications
'in either A or B steam generator. There was an additional five-tube sample taken in the B steam generator cold leg as a result of two damaged tubes during the tube lane blocking device removal project. The two tubes which had received minor damage were preventively plugged.
The Phase 2 inspection was performed on A steam generator as a result of a loose parts concern raised during the closecut inspection following the tube lane blocking device removal project. This inspection found no indications on any of the tubes.
l 109
)
1 s
l 5.1.3
' Repaired or Plugged Tube's The:following'is a' list of.the tubes mechanically-plugged'as'a result-of damage incurred during the-tube lane blocking device removal project.
Plugged Tubes in the B Steam Generator Row - Column Indication Location LOrigin 1-1 18" ATS CL OD 2-1 18"~ATS CL OD ATS - Above Tubesheet E
5 '. 2 Unit 2 5.2.1
. Inspection-Plan During.the Unit 2 Refueling 14 outage, eddy current -
testing was performed from.0ctober 12 to October 31, 1988. The extent of_the' inspection program in each steam generator is as follows:
Eddy' Current Inspection Plan Extent of Inspection Number of Tubes' Inspected A SG B SG Hot Leg (Cold Leg). Hot Leg (Cold Leg)
Full Length 56 58 Sleeves
- 53' 53 To Top of Sleeve *
(53)
(53)
U-Bend 18 19
- 1 TSP 1522 (3042) 1611 (2898)
L Sleeved at l
Both Ends (3)
TOTALS 1649 (3095) 1741 (2954)
- Exams were selected so that the full length of 53 tubes was examined in two parts.
i 110
_--______.__________,._x
.I T
.M
' ~
5.2.2 Inspection Results-
-E c
'The following is'a summary of the results;offeddy 11 5 current inspection showing the 'rundaer of tubes with indications in the ranges listed:
Eddy Current Inspection Resul'ts Hot Leg (Cold Leg).
<20%
.18(28)-
'10(142) 20-29%
9(60) 8(276) 30-39%'
7(22) 6(239) 40-49%
1(1) 1(22)
~'
50-59%
1
'2(13) 60-69%-
'(1) 70-79%
2 80-89%
1 1-90-100%~
3
- UDI 1
- DI-15 (8) 3 (30)
TOTALS 58(119) 31(723)
- UDI' indications are those whose quantitative analysis has not been possible but in previous instances have necessitated repair.
- DI - indications whose: quantitative analysis has.not~
been possible but in previous instances'have not-necessitated repair.
5.2.3
' Repaired or Plugged Tubes The following is-a' list of tubes which were mechanically plugged as-a' result of. indications found during eddy current inspection performed in 1988.
Plugged Tubes in the A Steam Generator Row - Column Indication %'
Location-Origin 33-25 86
.9.1" ATE HL.
OD 33-30 DI 7.1" ATE HL OD 37-36 DI
~1.7" ATE HL OD.
44-41 51 8.0" ATE HL OD 39-49 DI 3.7" ATE HL OD 34-62 DI 6.0" ATE HL OD 35-64 94 6.7" ATE HL OD 35-65 UDI 6.5" ATE HL OD 4-75 73 10.6" ATE HL OD l
6-76 94 13.7" ATE HL OD 3-79 DI 10.5" ATE HL OD 7-79 DI 8.6" ATE HL OD 20-83 DI 6.1" ATE HL OD 111
i::,
o 1
h
- c.,
s l
. Row - Column.
- Indication %-
Location.
OEigin bl' 8-86 73 7.;2" ATE HL
'OD-F
,18 DI 5.9" ATE HL' OD 10-87 43 7.iB" ATE HL OD 8-88 90
-3.0" ATE HL-CD 12-88 DI
.5.4" ATE HL OD
~
ATE -.Above' Tube End Distorted Indication DI
' HL Hot' Leg UDI - Undefined Indication Pluqqed Tubes in the B Steam Generator Row - Column 1 Indication %
Location Origin:
--15-70 40 0.4" ATS CL OD 15-71 42
~9.9" ATE HL-
.0D 10 35 9.9" ATE HL 10D '
~
5-18
g
'l-67 87; 3.4" ATE HL OD t
' 33-77' 42
- 1 TSP HL OD:
-' ATE -
Above' Tube'End' CL - Cold Leg Hot. Leg ATS - Above Tubesheet HL UDI. '
Undefined Indication
- 1TSPL-First Support Plate.
5.2.4 Tubes with In'dications Not Plugged The following is a list of. tubes which'had indications but were not plugged or sleeved as a ' result. of eddy t
current 1 inspections performed in 1988.
A Steam Generator Row - Column Indication %
-Location Origin 7-1 DI
- 1 TSP CL OD-18-5 35
- 1 TSP HL OD 3-18 DI 0.4" ATS HL OD 3-19 21 0.0" ATS HL OD 4-19 25 0.0" ATS HL OD 40-26 25 5.4" ATS HL OD 33-48 DI 0.2" ATS HL OD 38-52 DI 0.0" ATS HL OD 41-55 22 2.2" ATS HL OD 39-58 31 3.8" ATS HL OD 112
m -
p 7; >
.w,t li
.;;; R L i
9
~,
t.
I,-
.o,,
. Row
'Co umn
- Indication %
Location
, 0r'iqin l
36-59.-
'32
'2.'1" ATS HL.
'OD-28 3.6" ATS HL OD-35 34; 8.4" ATS HL OD-35-67 20 9.0" ATS HL'
'0D 4-76 DI 0.2" ATS HL.
OD 4 DI 0.1" ATS HL OD 8-77 31
~0.0".ATS HL OD 7-78 34 0' 5" ATS HL 0D 10 25 0.0" ATS HL' OD 33-78 20-21.7" ATS HL OD 10-79 271 0.0" ATS HL OD 6-80' 20 0.0" ATS HL OD.
L B Steam Generator
- Row
- Column Indication %
Location
' Origin-3-1 29
- 1 T5P CL.
OD 6-1 39
- 1. TSP CL'-
OD 12-2 31
- 10-3.
DI
13-4 DI
- 1 TSP'HL OD 28-23 21 0.5" ATS HL OD-11-24 26 0.7" ATS CL OD-28-24 DI 0.3" ATS HL OD 14-25 21 0.6" ATS CL OD 6-26 26 1.3" ATS CL OD 9-26 22 0.3" ATS'CL OD 11-27 26 0.5" ATS CL 10D-16-27 22 0.3" ATS CL OD 5-28 24 1.0" ATS CL
-0D 7-28 25 0.3" ATS CL.
OD 24 1.7" ATS CL OD 17-28 27 0.4" ATS CL OD 8-29 26 1.1" ATS CL OD 13-29 24 0.9" ATS CL OD 23-29 33
- 1 TSP CL OD 8-30 26 0 5" ATS CL OD 12-30 21 0.9" ATS CL' OD 15-30 26 0.9" ATS CL OD 16-30 21 1.2" ATS CL OD 113 t
-_i_--_.--... _ _ - _ - _ _ - - - - - - - - -
I i
1 Row - Column Indication %
Location Origin l
17-30 25 0.3" ATS CL OD 19-30 DI 0.1" ATS CL OD L
4-32 26 0.5" ATS CL OD 8-32 27 0.8" ATS CL OD 10-32
~20 1.1" ATS CL OD 13-32 25 1.5" ATS CL OD 26 2.5" ATS CL OD 21 0.9" ATS CL OD-16-32 DI 0.4" ATS CL OD 17-32 24 1.0"'ATS CL 0D 21-32 DI
- 1 TSP CL OD 8-33 27 0.9" ATS CL OD 11-33 24 0.7" ATS'CL OD 20 1.1" ATS CL OD 16-33 32 0.9" ATS CL OD 20 2.0" ATS CL OD' 11-34 24 0.6"'ATS CL.
OL 22 1.0" ATS CL OD 12-34 21 1.0" ATS CL OD 20 1.5" ATS CL OD.
16-34 21 0.8" ATS CL OD 11-35 25 0.5" ATS CL OD 20 1.1" ATS CL' OD 13-35 27 0.6" ATS CL OD 27 1.2" ATS CL OD 21 1.6" ATS CL OD 16-35 30 0.9" ATS CL OD 6-36 24 0.5" ATS CL OD 8-36 24 0.8" ATS CL OD 15-36 21 1.0" ATS CL OD 16-36 33 0.8" ATS CL OD-19-36 21 0.7" ATS CL OD 4-37 24 0.7" ATS CL OD 6-37 21 0.5" ATS CL OD 24 0.5" ATS CL.
CD 9-37 21 0.8" ATS CL OD 15-37 20 0.8" ATS CL OD 16-37 22 0.6" ATS CL OD 17-37 21 0.7" ATS CL OD 18-37 25 0.6" ATS CL OD 7-38 25 0.5" ATS CL OD 8-38 21 2.1" ATS CL OD 25 0.5" ATS CL OD 11-38 20 0.8" ATS CL OD 22 1.2" ATS CL OD 12-38 27 0.6" ATS CL OD 23 1.3" ATS CL OD 13-38 26 1.0" ATS CL OD 16-38 26 0.4" ATS CL OD-17-38 24 0.6" ATS CL OD 20-38 27 1.2" ATS CL OD 114
Row - Column Indication %
Location Origin 20-38 35 0.6" ATS CL OD l
24-38 21 0.4" ATS CL OD 6-39 25 0.5" ATS CL OD 9-39 25 0.8" ATS CL OD 10-39 23 0.5" ATS CL OD 15-39 23 0.4" ATS CL OD 17-39 27 0.6" ATS CL OD 19-39 33 0.5" ATS CL OD 22-39 26 0.4" ATS CL OD 11-40 26 0.6" ATS CL OD 14-40 20 1.1" ATS CL OD 22-40 25 0.5" ATS CL OD 8-41 27 0.5" ATS CL OD 41-41 20 23.9" ATS HL OD 13-41 23 0.9" ATS CL OD 14-41 25 0.5" ATS CL OD 21 1.2" ATS CL OD 15-41 23 0.5" ATS CL OD 21 0.5" ATS CL OD 10-41 21 0.8" ATS CL OD 17-41 20 0.6" ATS CL OD 9-42 20 0.7" ATS CL OD 6-43 27 0.6" ATS CL OD 23-43 26 0.9" ATS CL OD 3-44 25 0.5" ATS CL OD 1-44 33 1.2" ATS HL OD 12-44 26 0.7" ATS CL OD 1-45 DI 1.1" ATS HL OD DI 1.2" ATS HL OD 3-45 26 0.5" ATS CL OD 24-45 21 0.5" ATS CL OD 21-46 26 1.8" ATS CL OD 33-46 31
- 1 TSP CL OD 36-48 29 42.6" ATS HL OD 26 36.3" ATS HL OD 12-49 20 1.1" ATS CL OD 14-49 27 0.8" ATS CL OD 7-50 22 1.5" ATS CL OD 27 0.7" ATS CL OD 8-50 21 1.0" ATS CL OD 25 0.5" ATS CL OD 10-50 26 1.1" ATS CL OD 11-50 22 1.2" ATS CL OD 12-50 21 0.9" ATS CL OD 14-50 26 0.5" ATS CL OD 15-50 23 0.6" ATS CL OD 24-50 27 0.6" ATS CL OD 115
q 1
l
'i i
Row - Column Indication %
Location Origin I
4-51 22 0.6" ATS CL OD 5-51 24 0.8" ATS CL OD 8-51 24 1.1" ATS CL OD 9-51 27 0.8" ATS CL OD 15-51 27 0.8" ATS CL OD 16-51 34 0.8" ATS CL OD 17-51 20 0.7" ATS CL OD' 6-52 27 0.8" ATS CL OD 11-52 21 0.7" ATS CL OD 12-52 22 0.6" ATS CL OD~
14-52 22 0.6" ATS CL OD 18-52 25 0.9" ATS CL OD 20-52 20 0.8" ATS CL OD 3-53 27 0.6" ATS CL OD 5-53 20 0.7" ATS CL OD 14-53 26 1.0" ATS CL OD 17-53 27 0.7" ATS CL OD 20-53 22 1.1" ATS CL OD 19-54 22 1.1" ATS CL OD 6-55 22 0.7" ATS CL OD 7-55 25 0.8" ATS CL OD 22 0.2" ATS CL OD 9-55 22 0.4" ATS CL OD 10-55 27 1.4" ATS CL OD 15-55 25 1.1" ATS CL OD 14-55 27 1.0" ATS CL OD 16-55 25 1.0" ATS CL
'0D 19-55 27 1.1" ATS CL OD 11-56 23 1.5" ATS CL OD 13-56 25 1.1" ATS CL OD 11-57 20 1.4" ATS CL OD 14-57 20 1.2" ATS CL OD 15-57 25 1.3" ATS CL OD 16-57 25 1.1" ATS CL OD 21-57 26 1.0" ATS CL OD 25 0.6" ATS CL OD 15-58 22 0.9" ??S CL OD 17-58 26 0.9" ATS CL OD 33-58 28
- 1 TSP CL OD 8-59 26 1.6" ATS CL OD 12-59 26 1.2" ATS CL OD 16-59 29 0.7" ATS CL OD 21-59 26 0.6" ATS CL OD 22-59 23 0.6" ATS CL OD 3-60 22 0.7" ATS CL OD 10-60 26 0.8" ATS CL OD 22 0.4" ATS CL OD 11-60 20 1.1" ATS CL OD 21 0.7" ATS CL OD 15-60 24 1.1" ATS CL OD 16-60 24 1.0" ATS CL OD 42-60 29 5.9" ATS HL OD 116
7.
I Row'- Column Indication %
Location Origin 6-61 21 1.0" ATS CL OD 20 0.4" ATS CL OD 7-61 20 0.4" ATS CL OD 8-61 20 0.3" ATS CL OD 21 1.3" ATS CL OD 10-61 21 0.9" ATS CL OD 15-61 25 1.3" ATS CL OD 16-61 27 1.0" ATS CL OD h
17-61 25 0.9" ATS CL OD 19-61 24 0.6" ATS CL OD 21-61 23 0.9" ATS CL OD 10-62 26 1.8" ATS CL OD 16-62 20 0.8" ATS CL OD 19-62 22 0.5" ATS CL OD 10-63 26 1.9" ATS CL OD "24 1.1" ATS.CL
'0D 14-63 23 1.1" ATS CL OD 17-63 24 0.5" ATS CL OD i
21-63 25 0.7" ATS CL OD 10-64 26 1.3" ATS CL
.0D 11-64 22 2.3" ATS CL OD 22 1.5" ATS CL OD 21-64 26 0.7" ATS CL OD 20 1.0" ATS CL OD 20 1.6" ATS CL OD 36-64 26
- 1 TSP CL OD 12-67 24 0.5" ATS CL OD 14-67 25 0.7" ATS CL OD 17-67 21 0.8" ATS CL OD 20-67 25 0.8" ATS CL OD 21-67 22 0.6" ATS CL OD 5-68 21 0.7" ATS CL OD 8-68 25 0.8" ATS CL OD 13-69 22 0.8" ATS CL OD 36-69 DI
- 1 TSP CL OD 6-70 21 0.5" ATS CL OD 11-70 25 1.0" ATS CL OD 7-71 21 0.6" ATS CL OD 20 1.2" ATS CL OD 8-71 20 0.7" ATS CL OD 11-71 22 0.8" ATS CL OD 10-72 20 0.6" ATS CL OD 25 1.,3" ATS CL OD 5-74 35 0.2" ATS HL OD 5-75 DI 0.0" ATS HL OD 117
I Row - Column Indication %
Location Origin 26-75 25 49.6" ATS HL OD 23 50.3" ATS HL OD i
10-76 25 0.0" ATS HL OD 33-77 22
- 1 TSP CL OD 7-78 DI 0.4" ATS HL OD 27-83 35 1.6" ATS HL OD 5.2.5 Preventive Sleeving Program During the fall outage, Unit 2 steam generator cold leg tube ends were sleeved using the Westinghouse ROSA system. This project was performed due to cold leg wastage progression which was observed during eddy current examinations in prior years. 'A total of 509 tubes were selected based on eddy current results. The criteria used for selection of these tubes were:
- first, the location of the tube; second, the location of the defect; and third, the severity of the degradation. -
A list of the successfully sleeved cold leg tube end is as follows:
A Steam Generator Row-Column Row-Column Row-Column Row-Column 9-24 10-29 8-32 10-35 13-24 13-29 9-32 11-35 12-24 14-29 13-32 12-35 11-25 15-29 15-32 17-35 12-25 16-29 17-32 18-35 9-26 17-29 18-32 19-35 11-26 18-29 19-32 21-35 12-26 6-30 20-32 9-36 13-26 8-30 8-33 10-36 5-27 10-30 9-33 11-36 7-27 13-30 10-33 12-36 11-27 14-30 11-33 18-36 12-27 15-30 12-33 19-36 13-27 17-30 16-33 20-36 14-27 18-30 18-33 21-36 15-27 6-31 19-33 10-37 6-28 7-31 20-33 14-37 8-28 8-31 10-34 17-37 l
11-28
'0*?-
11-34 19-37 l
13-28 21 al 12-34 10-38 14-28 17-31 17-34 11-39 15-28 18-31 18-34 13-39 16-28 19-31 19-34 6-40 7-29 6-32 20-34 10-40 9-29 7-32 21-34 3-41 118
E--
I:
l Row-Column Row-Column Row-Column Row-Column 10-41 10-46 16-46 25-56 9-42 11-46 10-47 5-63 i
10-44 12-46 12-47 12-68 j
9-45 14-46 16-47 9-46
.15-46 15-48 B Steam Generator Row-Column Row-Column Row-Column Row-Column 9-24 19-32 23-36 15-42 10-24 4-33 5-37 16-42 13-24 5-33 7-37 17-42 8-25 6-33 8-37 18 9-25 7-33 10-37 21-42 11-25 9-33 11-37 8-43 12-25 10-33 12-37 12-43 7-26 12-33 13-37 13-43
- 11-26 13-33 21-37 15-43 13-26 18-33 22-37 16-43 14-26 19-33 5-38 17-43 15-26 20-33 9-38 18-43 5-27 21-33 10-38 21 12-27 4-34 18-38 4-44 14-27 5-34 21-38 8-44 15-27 6-34 23-38 10-44 6-28 7-34 4-39 11-44 10-28 8-34 7-39 13-44 12-28 9-34 8-39 14-44 15-28 10-34 11-39 15 18-28 13-34 12-39 16-44 4-29 18-34 13-39 18-44 5-29 19-34 21-39 20-44 7-29 20-34 23-39 22-44 9-29 5-35 8-40 12-45 14-29 6-35 9-40 13 15-29 7-35 10-40 14-45 16-29 8-35 12-40 15-45 1
4-30 9-35 13-40 20-45 5-30 10-35 15-40 20-45 6-30 12-35 16-40 21-45 9-30 18-35 5-41 22-45 18-30 20-35 7-41 23-45 5-31 21-35 16-41 3-46 7-31 22-35 22-41 4-46 9-31 5-36 5-42 13-46 10-31 10-36 6-42 15-46 13-31 11-36 7-42 20-46 l
15-31 12-36 8-42 22-46 17-31 13-36 10-42 13-47 18-31 14-36 11-42 14 47 6-32 20-36 12-42 17-47 7-32 21-36 13-42 18-47 l
k 12-32 22-36 14-42 20-47 119
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t 6;0 IREACTOR COOLANT SYSTEM RELIEF VALVE CHALLENGES-
?
6.1 Overpressure Protection During Normal Pressure & Temperature Operation
'l There were no challenges to the Unit 1 or Unit 2 reactor coolant p
system power-operated relief valves or safety. valves ~at no..nal operating pressure and' temperature in 1988.
'6.2 Overpressure Protection During Low Pressure'& Temperature operation During low pressure and temperature operation, the low temperature overpressure protection system circuit momentarily
. opened a Unit 1 power-operated relief valve on May 10,1988, at 1724 hours0.02 days <br />0.479 hours <br />0.00285 weeks <br />6.55982e-4 months <br /> following the start.of the B reactor coolant pump.
The slightly higher temperature on the secondary side'of the steam generators resulted in an increase of the temperature of the water in the reactor coolant ~ system causing an increase'in l
j.
the pressure of the reactor coolant. system.
There were no challenges to the Unit 2 power-operated relief valves'during low pressure and temperature operation, j
7.0 REACTOR COOLANT ACTIVITY ANALYSIS There were no indications during operation of Unit I and Unit 2.in
-l i
1988 where reactor coolant activity exceeded ~that allowed by Technical Specifications.
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y Wisconsin:
Electnc
- POWER COMPANY f-231 W Michigor., RO, Box 2046, Milwaukee,WI 53201
. [414)221-2345
[
VPNPD-89-112 NRC-89-26
~ February 27,.1989 U.S. NUCLEAR REGULATORY COMMISSION 10CFR50.59(b)
L.
Document' Control Desk
' Mail Station P1-137 Washington, D.C. 20555 L-Gentlemen:
DOCKET NOS. 50-266 AND 50-301 ANNUAL RESULTS AND DATA REPORT POINT-BEACH NUCLEAR PLANT, UNITS 1 AND 2 Enclosed-is the Annual Results and Data Report for the Point Beach Nuclear Plant, Units 1 and 2, for the year 1988. ~ This
~
report is submitted in accordance with Technical Specification 15.6 9.1.B. pursuant to the requirements of 10CFR50.59(b).
The report contains information regarding operational highlights of the Point Beach' Nuclear Plant operations during 1988 and includes description of facility changes, tests and experiments, personnel-occupational exposures, results of steam generator inservice inspections and listingsof reactor coolant system relief valve challenges. -Ten bound copies of this. report are also being provided to you under a seperate cover.
Very truly yours, e
y!
C. W. Fay Vice President Nuclear Power Enclosure copies to NRC Regional Administrator Region III NRC Resident Inspector II t
A subsidhy ofIIIsavish Dway Cupratkn l