ML20090D125

Annual Rept of Changes,Tests & Experiments,Calvert Cliffs Nuclear Power Plant 1991

ML20090D125
Person / Time
Site:	Calvert Cliffs
Issue date:	12/31/1991
From:	Creel G, Denton R BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9203060039
Download: ML20090D125 (25)

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/ BALTIMORC GAS-AND ELECTRIC 1650 CALVERT CUFFS PARKWAY

• LUSBY, MARYl.AND 20657-4702 Gromot C, CREEL Vict PRE 54QENT nucle An LNEnov (eio)teo-ests February 28,1992 U.S. Nuclear Regulatory Commission Washington, DC 20555 J ATTENTION: Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2: Docket Nos. 50-317 & 50-318 B_enort of Chances. Tests. and Experiments REFERENCE- (a) 10 CFR 50, Paragraph 50.59(b)

Gentlemen:

As required by the abeve reference, please find enclosed our annual report of changes, tests, and experitcents completed on Calvert Cliffs Unit I and/or 2 under the provisions of 10 CFR 50.59(a),

including a summary of the safety evaluation for each. This report covers the period from January 1, 1991 through December 31,1991.

Items in the repa t are referred to by Facility Change Request (FCR), Field Engineering Change (FEC), Temporary Modification or Miscellaneous Activity number.

Should you have any questions regarding the contents of this report, we will be pleased to discuss them with you.

-Very truly yours, ,

/

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T' N~h${ -

for G. C, Creel Vice President . Nuclear Energy

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GCC/JBB/ alm /rej %0

j. - Enclosure 1: Annual Report of Changes. Tests and Experiments (22 pages) l '

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Document Control Desk February 28,1992 Page 2 cc: D. A. Brune,lhquire J.11. Silberg, lhquite R. A. Capra, NRC D. O. McDonato Jr., NRC T. T. Martin, NRC P. R. Wilson, NRC

.R.1. Mclean. DNR J.11. Welter, PSC

ENCLOSURE (1) ll ANNUAL REPORT OF CHANGES, TESTS, AND EXPERIMENTS CALVERT CLIFFS NUCLEAR POWER PLANT 1991 l

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Pagd

• 10CFR50.59 Anmmi Report CCl 117 S/N 1-91 1 This temporary modification was to gag open flow control valve 1-SRW 1627CV in present operatmg position while the associated flow control Pump Seal Water Coolers.

The flow controller 1 PIC 1627 had malfunctioned and was there was no unreviewed safety question or change in the Technical specifications.

CCI 117 S/N 1915 The temporary modification authorized the removal of the inboard and outboard j bearing temperature elements from the No.11 Componen Computer).

The The originally installed motor for #11 CCPP failed and was removed for repairs.

currently installed replacement motor (per FEC 90-03120) installed. This temporary modification supports plant operation during ALL MODES.

This 50.59 was required because the temperature . No. 60elements 235-E). The (1TE3813A temperature& B) were shown o FSAR Fig. 9 6 (M 51, Sh.1 of 3 Rev. 28, BG&E Dwplant Computer are not otherwise elements and their respective inputs to the DAS/l discussed in the FGAR or the Technical Specifications, llowev Cooling Water System.

The temporary modification was processed to support con safety question or change in the Technical Specifications.

CCI-l 17 S/N 191-2ji 1-TE 156 is a temperature element which monitors the upper guide bearing temperatu a motor. The RTD is spring loaded within the motor housing No.11 A Reactor Coolant Pum ie point of contact. Tne lube oil was leaking, dat approximately 4 inches from t the the springs where it was accumulating. Temporary Modification (CCl-ll7) S/N 1-91-25 A similarsevere RTD outside the motor housing and sealine the sheath with a 1987.

' motor. The The leaking RTD cannot be replaced without partial disass TechnicalSpecifications.

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4 10CFRS6.59 Annual Report Pagel CCI-117 S/N ?-91-6 This Temporary Modification substituted flasks of nitrogen gas in place of instrument air to operate _ the ' safety-related pressurizer auxiliary spray line control valve, 2 CV 517.

Instrument air needed to be secured in order to conduct the repair / replacement of +

instrument air check valve,2 IA 175, which was leaking.

Operability of the auxiliary s, ray line control valve 2-CV 517 was maintained to provide pressurizer spray cooling of t ie llCS during plant heatup or if the reactor cooling pumps were sect. red. 'Ihe Safety Evaluation concluded there was no unreviewed safety question or change in the Technical specifications.

CCI 117 ' S/N 2 91-016 e This Temporary Modification isolated LPSI loop check valve,2-SI 144, from the line that

~ taps into the cold leg of reactor coolant loop 22A by freeze scaling a section of the injection

- line downstream of the valve. The freeze seal resulted in isolating both the llPSI and LPSI-

- headers from reactor coolant loop 22A.

p A freeze seal was required to isolate reactor coolant loop 22A from 2 SI-144 and to n maintain reactor coolant pressure boundary so that the valve may be repaired in place, o

Valve 2 SI 144 required repair due to improper seating of the disc resulting in backleakage.

The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

DCR 90-694

~his activity revised drawing M 65 Sh.4, Rev. 7, Ventilation Systems Control Room HVAC to reflect the as built condition.

The DCR was a result of walkdowns performed to verify "as built" vent drain and

- instrumentation configurations versus existmg OM and P&lD drawings.

The 1/4" test connections were shown on the chilled water pipin The test connections were not shown on P&lD M-65 Sh, 4.The g drawing M 807,60-564-E.

test connections were shcwn incorrectly on OM 65 Sh. 4. These drawings were corrected per DCR 90 693. The Safety Evaluation concluded there was no unreviewed safety question or change in the ,

TechnicalSpecifications.

FCR 84-1072

' This modification -- to the Containment Vent / Hydrogen -Purge System involved the L replacement and relocation of the existing non-safety-related Foxboro FE-6901 Flowmeters with non-safety-related air flow monitors (e.b, Kurz Model #455 08) with an indication range from 0 - 750 SCFM. The new flow tota.izers added to the system as a part of the new air flow monitors were not safety-related. The location where the existing Foxboro flowmeters were installed was replaced with a pipe spoolpiece.

The non-safety-related Foxboro FE-6901 Flowmeters were replaced as they do not meet the sensitivity / accuracy requirements for air flow values described in the FSAR post-accident analysis. The Foxboro Flowmeters were replaced with non safety related air flow monitors installing t_he new flow transmitters in the vertical section of the Ilydrogen Purge System

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_10CFR50.59 Annual Report - Page 3 3ipe downstream of the outboard MOV and installing the flow indicator / totalizer in the Cryogenics Room. Installing the new Dowmeters in the vertical section of pipe meets the

-vendor's requirement of havmg a straight pipe length of 10 pipe diameters upstream and 3 -

pipe diameters downstream from the flow transmitter.

A 50.59 Evaluation was needed because the activity constituted a revision to FSAR figures 6 9,9 20A and Section 6.8.3.3. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

FCR 851023 Letter L91-090-documented licensing concerns recarding the capability of instrument channel FT-212 to satisfy licensing criteria from NRC Regulatory Guide 1.97 regarding post. accident monitoring of charging now. During the past three years efforts have been made to compensate for system now pulsations caused by the positive displacement charging pumps and thus increased the accuracy of channel F1 212. Although some success has been achieved and further testing is planned, evaluations can be undertaken to demonstrate that the existing plant configuration is technically adequate.

The technical evaluations provided herein demonstrate the capability of utilizing existing plant instrumentation to ascertain _ proper post accident operation of the char ing system.

These evaluations show that the Calvert Clius Nuclear Power Plant (CCNPI meets the intent of Regulatory Guide 1.97 given the current existing plant configuration, pecifically, it is proposed that the charging pump breaker lights be utilized as Category 2 indication of chargmg system performance while FT-212 be utilized for Category 3 backup indication.

The charging pump electric current meters, which are alreadl 3 utilized as Catego ' 2 indication, can b: utilized as an additional backup to the chargmg pump breaker il ts.

These evaluationi demonstrate that plant safety is not compromised given existing avai ible status indicators 1or charging flow. Charging flow is tested using STP 0-73D 1(2). Test data has shown that a char;ing pump in degraded packing wil. dehver. gpm, 414.With goodthiscondition test data, will deliver use of 44 45 pump the charging gpm, and a pum breaker lights, (backed up by the electric current indications and FT-212), to demonstrate charging system performance is justified.

Licensing commitments currently,specify that control room indicated charging Dow rate from channel FT-212 can be designated as a type D, Category 2 variable as defined in Regulatory Guide 1.97. A type D designation means that the prescribed variable is utilized to ascertam the operation of a safety system. Category 2 is the recommended designation

-for this parameter as per Regulatory Guide 1.97 and thus, the as licensed plant design satisfies the licensing criteria therein via direct compliance. However, channel inaccuracies

-due to pulsating flow from the positive displacement charging aumps have called into question the reliability of FT 212 and thus, its stand alone capabi'ity to satisfy Category 2 criteria from Regulatory Guide 1.97. -

-A review of the maintenance history associated with channel FT-212 showed that the channel had consistently produced high flow readings over the life of the plant. Efforts to-correct this problem had yielded only partial success. Additionally, a review of applicable CCNPP procedures indicated that operators utilize charging pump breaker lights as the preferrec, charging system performance indicator (vice FT 212), lentting further credence to this proposed change.

The change is technically suitable and meets the intent of the NRC Regulatory Guide 1.97.

The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

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- ICCFR50.59 Annual Reoort Pace 4 l

FCR 851052

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in order to moet 10CFR 50.62, this FCR installed an additional means of scramming the reactor on hig pressurizer pressure. The design used PT 102ABCD to develop a EhFAS l 2/4 actuation s nal.

The ESFAS Am logic cabinet will open the load contactor of the M/G set 27'(1 & 2 G206).

The ESFAS TL logic cabinet will open the load contactor of the M/G set on 45' (1 & 2 G306).

During normel operation, a single channel trip would not de-energize the RTS bus and cause rod drop. Rod drop will occur on a single channel trip only if the M/G set load sharing circuit docs not operate properly. The load sharing feature has operated successfully in the past to catch the load of a M/G set trip.

A new by-pass contactor was installed in parallel with the existing load contactor to allow testing of the load contactor while maintaining generator output. The load contactor and by-pass contactor are controlled by different ESFAS channels so that while testing the channel 'A' load contactor, the by pass contactor is contrrIlin the Channel 'B' and vise-versa. Using this scheme,if the ESI AS channel controlling the pass contactor receives a high pressunzer pressure signal, the output of both M/G sets will e lost.

This scheme uses a set of auxiliary contacts on the by-passed condition at control room panel C05. The by-pass contactor is fully rated for the M/G set output and there are no regulatory aass mode, a restrictions single DSS trip against signal of thecontinued operation proper channel wil in the byl trip the reactor. Eac four channel histables is annunciated on 1(2) COS if a histable tnps. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

FCR 85-1052

- This modification installed a Diverse Scram System (DSS) in compliance with 10 CFR 50.62, Reduction of Risk from Anticir'*ed Transients without Scram events. The DSS uses existing Pressurizer Pressure Transmitters PT102A, B, C, D, to provide a high pressurizer aressure signal to "high" bistables added to the pressure loops circuitry. Unused isolators, ,

aistables, and logic modules in the ESFAS ,anels are used as well as existing CEDM Motor Generator controls. The addition of a new aypass contactor in p_arallcl with the existing load contractor for the CEDM Motor-Generator allows testing of ATWS at power.

.The four pressure channels provide pressure signals to four high bistables in the ESFAS sensor cabinets. Each bistable provides channel trip annunciation, and input to two isolators. One isolator provides an input to a two-out-of four logic module in channel"A" of

-the ESFAS logic cabinets while the other isolator inputs to channel"B". The logic module supply annunciation and data logger input for " Diverse Scram System' Trip", each also energize a trip relay in the ESFAS relay cabinets to open a contact in the " load oft" portion -

of the MG set control circuit thereby causing the load contactor (3M) to open. Both Motor-Generator set load contactors must open to cause a reactor trn. The Safety Evaluation concluded there was no unreviewed safety question or change in _- th'e Technical Specifications.

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10CFR50.59 Annual Report ., Pagej FCR 87-0073 Supplement 9 to FCR 87-0073 was issued to facilitan the processing of DCR 89-826. This DLR was released to revise P&lD hi 56, Sh. 3 of 6 Plant Fire Protection System, Turbine, and Service Bldgs. & Intake Struct., Units No.1 & 2() to reflect the addition of a non sa related fire protection sprinkler system in the North Service Building. P&ID hi-56, Sh. 3 o'f 6, is FSAR Figure 9 228. Supplement 9 was also issued to update FSAR Table 9 20 to re0cet new information associated with the sprinkler system. The FSAR P&lD does not currently reflect the configuration of the sprinkler pipinj in the North Service Building. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

FCR 87-0111 This modification amplified the FSAR description of the PORY Block h10Vs to note that electrical separation was not required between the circuitry from hiOV 403 and h!OV 405.

FSAR Section 8.5 requires six 6) inches of separation or barriers between safety related circuits of different se 3aration (groups. The control board handswitches do not meet this criteria. The Safety 3 valuation concluded there was no unreviewed safety question or change in the Techmcal Specifications.

FCR 8%054 This change allowed the use of an alternate material (i.e.,15 SPH stainless steel in lieu of 300 series stainless steel) for the Charging Pump Fluid End Cylinders (i.e., blocks),

l_xakage from cracked blocks have been experienced in #12 and #23 charging pumps. A new block was installed on #12 pump and it experienced cracks in less than 1 year. A second new block was installed on #121 amp but it is the same design as the one that failed in less than 1 year. The original block from #12 pump has been repaired and was used to replace the #23 block.

The blocks had reached their service life and needed to be replaced on an "As Failed" basis.

The proposed change to 15-SPH material enhanced the fatigue endurance of these blocks.

. The Safety Evaluation concluded there was no unreviewed safety question or ch nge in the L Technical Specifications.

FCR 89-13.

This modification removed the internals from the safety related check valves 1-SRW.321,1-SRW 322 and 2-SRW 321 located on the service water return headers for the Emergency

- Diesel Generators (EDGs) 11,12, and 21 heat exchangers, respectively. The check valve bodies will remain in place as spool pieces.

-This modification was performed for the following reasons:

1. Nonconformance Report (NCR) 9255 was initiated to document that a commitment made to the NRC to perform periodic reverse Dow testing on check valves 1-SRW.

321,1-SRW-322 and 2 SRW-321 was not being implemented. The cause of this nonconformance condition was a failure to develop a controlled test procedure to -

L adhere to the commitments made to the NRC regarding the IE Bulletin 83-03. The purpose of reverse flow testing is to ensure that check valve disassembly failure has not occurred. Since the check valves do not perform a necessary function, (See J

m 10CFR50.59 Annual Report Pace 6 4 -

Reason 3) their presence in the system requires unnecessary testing and maintenance. - A separate response to NCR 9255 dated March 15,1990 addressed the failure to perform periodic testing and concluded that it would not create a safety hazard.- Removing the check valve mternals would eliminate the need for periodic testing and maintenance.

2. The Nuclear Regulatory Commission (NRC) issued IE Bulletin No. 83-03 which is primarily directed at the failure mode of disassembly or partial disassembly of check valve internals in the raw water cooling system of diesel generators, Several utilities have experienced failures of check valves whete Diesel Generators have been declared inoperable because of cooling loop blockage, Due to the frequency of check valve failures reported in this bulletin, it is BG&E's position (for safety concerns) to remove unnecessary check valves in the cooling systems of the EDGs.

Removal of the check valve internals would prevent the possibdity of Service Water blockage to the EDGs' heat exchangers caused by check valve internals failing.

The intent of NRC Bulletin 83-03 is to prevent blockage of the raw water cooling systems for EDGs by implementing testing programs to detect check valve disassembl failures. The removal of17e check valve internals would ensure that the intent of t e NRC bulletin is met. The only safety functions of the valves on the cooling systems for the CCNPP EDGs is to maintain the pressure boundary and to permit passage of the design flow. The removal of the check valve internals will not adversely affect the pressure boundary of the valve bodies and will ensure the passage of design flow.

3. The existing check valves 1-SRW-321,1 SRWa2 and 2 SRW-321 are not required for backflow prevention or separation criter.a and do not provide a necessary function other than to maintain the SRW pressure boundary and to permit passage of the design flow. A review of plant history indicated that these check valves were left over as part of the original 5RW design. In the original SRW design, the SRW system was not separated into two subsystems (two subsystems for each unit) and the check valves provided double valve isolation for single valve failure criteria, while the 1ressure boundary for an EDG's heat exchanger was breached during maintenance.

However, before the first fuelloading of Unit :. the SRW System was separated into two subsystems by replacin the cross over valves with blind flanges. The separation was completed around 197b' and is the existing configuration of the SRW sy this configuration, the check valves are not required for separation criterm. There are presently two isolation valves in series located upstream and two downstream of each EDG train which meets the single valve failure criteria.

Service water is sup ierated control valves 1-CV-1587,2 plied and CV-1587 to each dieselforgenerator 1-CV-1588 EDGs 11,21,through and 2,the air o respectively.

These valves automatically open upon receipt of a signal when the diesel reaches 260 rpm. The valves then modulate (automatically reposition) to maintain a 5 to 7 psig pressure drop across the three diesel generator heat exchangers. Thus, the control valves are closed when the respective.EDGs are shutdown and throttle open to maintain a positive differential pressure across the EDG's heat exchangers when the EDGs are operating. Therefore, backflow protection is provided by the operation of the control valves. In addition, the SRW is a steady state system where a positive differential pressure is always available across the EDG heat exchangers. A1/2" ass line is provided around the control valve and the heat exchangers for each II G; This bypass line continuously supplies SRW to the safety related air after cooler for the mr start system for the respective EDG. The potential for back flow through this line is limited by the relative small size of the piping and the frictional

10CFR50.59 Annual Report pace 7

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i losses through the system and after-cooler. In addition, these lines bypass the EDGs heat exchangers where if backflow did occur through these lines it would not pose a problem for the EDG heat exchangers. Therefore, the system does not require check valves to prevent backflow through the EDG heat exchangers as backflow prevention is provided by the control valves and the availabihty of a positive differential pressure.

Additional backflow protection is provided by the elevation difference between the  ;

EDGs and the SRW tie-in for the SRW Pump return line. The EDGs are located on the 45' elevation and the check valves are located immediately before the SRW Pump return line tie in on the 6'-9" elevation. Therefore, the EDGs are additionally protected from backflow by the inertia of an approximate 38' static pressure head.

4. From a maintenance standpoint, this modification will climinate the need for .

periodic testing and maintenance on the check valves and will facilitate the refilling of the EDG SRW piping system following maintenance to the system.

5. The removal of the check valve internals will acercase the aressure drop across the "

check valves and, therefore, will enhance the performance of the system. ,

related. However, removing the The internals'check fromvalves these valves affected did notby this activity compromise any are o safetyf the safety functions of the S system as the check valves do not provide a safety related function other than pressure boundary ai.d opening to allow flow passage. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

FCR 84-0068 NCR 7556 was initiated due to the FSAR description of the RWT l water volume and the actual useable volume being in disagreement.

Specifically, this activity will do the following:

1. Provide a better description of the quantity of water supplied by the RWT.

2. Correct the description of the level alarms on the RWT and clarify the purpose of the alarms,

n. Paragraph 6.3.Laage 6-4 Existing wording:

These headers are initially supplied with borated water from the Refueling Water Tank and after that tank is 10 percent full, borated water is recirculated from the sump of the containment.

Recommended wording:

After the headers are initially supplied with at least 360,000 gallons of borated water from the Refueling Water Tank, a Recirculation Actuation Signal (RAS) occurs.

The RAS shifts the suction of the headers from the RWT to the containment sump to recirculate the borated water, l.

10CFR50.59 Annual Report Page 8

b. Paragraph 6J.: page 6 6 Existing wording:

In the event the automatic transfer fails to occur u Ton Recirculation Actuation Signal (RAS), the redundant refueling water tank ow level alarms are provided to alert the controi room 0 3erator. The operator un simulate automatic RAS actuation by two manua: pushbuttons labeled RAS "A" and RAS "B" on the main control boards."

Recommended wording:

In the event the automatic transfer fails to occur Recirculation Actuation Signal (RAS), the operator can manually initiate recirculation using two buttons labeled Recirculation Manual Actuation Channel"A"("B") on the main control board.

c. Paragraph 6 3.7 4. page 6-12 Existing wording:

The Refueling Water Tank is provided with a high level alarm and redundant low level and temperature alarms.

Recommended wording:

The Refueling Water Taak is provided with both a wide range and a narrow range

-level indicator. TFe narrow range instrument provides both a high level alarm. The wide range instrument provides only a low level alarm. The high level alarm is to alert the operators of an impending overflow of water from the RWT to Miscellaneous Waste Processing System. The low level alarms are used to assist the operator in monitoring for sufficient water inventory in the RWT. Redundant temperature instruments provide both high and low temperature alarms.

d. Paragraph 6.3.3 page 6-14 Existing wording:

When the Refueling Water Tank is 10 percent full, a recirculation actuation signal (RAS), opens the isolation valves in the two lines from the containment sump.....

Recommended wording:

When the Refueling Water Tank level reaches the RAS set loint, a Recirculation Actuation signal (RAS) occurs which opens the isolation va ves.....

e. Paragraph 6.4.2 page 6-21 Existing wording:

The minimum capacity of the RWT is 400,000 gallons whereas the capacity of the reactor coolant system is 77,800 gallons.

10CFR50.59 Annual Repart yngt 9 l Hecommended wording:

The Technical Specification requires a minimum inventory of 400,(XXI gallons be maintained in the RWr. Additional RWT data is provided in Tahle 6-4. The Technical Specification limit and RAS setpoint level have been established to ensure the RWT l actuated,The povides at of capacity least 360,000 the reactor gallons coolant of is system usable 77,800water before the RAS is gallons.

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f. Paragraph 6.4.2.page 6 22 Existing wording:

When the low liquid level is reached in the Refueling Water Tank.....

Recommended wording:

When the Refueling Water Tank level reaches the RAS setpoint, a Recirculation Actuation Signal (RAS) occurs which opens the isolation valves and... .

g. Bm! graph 7.3.2,1.page 7-25 Existing wording:

In addition, each provides a high level and a low leve' alarm.

Recommended wording:

ranga indicator provides only a low level alarm. The narrow range The fullprovi indicator (wide) des both a high level and a low level alarm. The high level a alert the operators of an impending overflow of water from the RWT to the Miscellaneous Waste Processing System. The low level alarms are used to assist the operator in monitoring for sufficient water inventory in the RWT, The Safety Evaluation concluded there was no unreviewed safety question or a change to the technical specifications.

FCR 89-0026

~ This FCR initiated the following modifications to the Auxiliary Feedwater System turbine Main Steam supply lines. These lines were designed to ANSI 1131.7 Class 11 from the Main Steam valves. penetrations up to and including the isolation valves and to ANSI 1131.1 beyond the

1. Added an air-operated 2" bypass valve and associated position switches around each Main Steam admission valve (1/2 CV-4070 and 4071). The admission valve will begin opening a short time after the actuation signal is initiated. The time delay for each steam admissica valve will be accomplished via a safety-related (SR) adjustable time delay relay located in Main Control Room (MCR) Panel C04.

- 2. Added a flange pair and restriction orifice downstream of the 2" bypass valve.

3. Relocated the check valves from the vertical run of the steam supply line to a horizontal run.

~ JOCFR50.59 Annual Brport bpe 10 '

4. Added a locked o}.en manual gate valve u istream of each station's steam admission, new bypass, and existing manual bypass va ves.

5. Added a manual bypass valve around both valves in item 4 above. l

6. Added a manual, locked open gate valve downstream of each relocated check valves before the individual steam admission lines combine.

7. Replaced the handswitches for the steam admission valves so that both the admission valve and the bypass valve at each valve station me controlled from a single handswitch and provide position indication for cach valve individually. The existmg  ;

AFAS relay is used to actuate both valves.

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8. Replaced the handswitches for the Auxiliary Feed pump turbine trip with push i button switch;s.  !

9. Removed the air supply for Main Steam admission valves CV-4070 and CV-4071 ,

from the AFW accumulators B and A, respectively, and provide a new air source for i 4 these valves and their res )ective bypass valves CV 4070A and CV-4071 A. The new air supply will be normal from the NSR instrument Air header, backed up by the SR SWAC system- (whi h is normally isolated) as well as by dedicated SR accumulators. The accumulators will be charged by the existing AFW air amplifier system, which also includes normally isolated SWAC and nitrogen backup. The accumulators (one for each valve station) will be located in the Service Water (SRW) purnp room. Each one will be sized considering system leakage for two hours and then stroking its associated valves two times at the end of the two hours. A low pressure switch will be provided on each accumulator to initiate a low pressure alarm m the MCR.

! 10. Relocated the existing solenoid valves and associated air re :ulators for CV-4070 and

! CV-4071 from the MSIV Room to the East Penetration 3oom, Elev. 27' 0". The new solenoid valves, gauges, check valves, etc. for CV-4070A und CV-4071 A will also be located in the East Penetration Room. Sizing of the solenoid valves has taken into consideration the 5 second opening / closing criteria for these valves. All new and replacement com this modification.ponents are valves All solenoid SR, including tle electrical will be normally installation de energized o in support of '

accordance with the existing design for the Main Steam Admission Valve.peration, i In addition, FEC 89 26-12 requests a correction to the statement of small pipe high energy line break criteria as stated in FSAR Chapter 10A.

t The changes were required to allow the turbine governor to accelerate the Auxiliary Feedwater pump in a more controlled manner. The aafety Evaluation concluded there was no unreviewed safety question or change to the technical specifications.

- FCR 89-12].

FCR 89121 was being issued to correct the FSAR descri tion of Section 11.3," Radiation

.theFS Safety"AR to match those which are provided"Calvert in CCI-8l. The change Cliffs Radiation Safetycorrects the.

Manual". This 50.59 Evaluation was being written to evaluate the consequences of changino the administrative quarterly dose limits currently stated in the FSAR to match the actus administrative quarterly dose limits as shown in CCI-800C.

10CFR50.59 AnnualIkport Page 11 Quality Audits unit Finding Sheet, finding number 89-02-1 finds the "as-found condition of

-- the FSAR" as: "FSAR Section 11.3.3.2, Personnel Monitoring Progiam, states that "an alert system will be sued to emphasize those individu$ils who are approaching the administrative quarterly dose l'oit (1.25 rem)".

But CCl-800C, Cidvert Cliffs Radiation Safety Manual, Attachment 1) Section 1. C.,

Administrative Dose Limits, states that "the quarterly whole body dose li(mit for individ 19 years of age older, is administratively set at 2.0 rem. Individuals will be restricted at the

-Alert Point (900 mrein)..."

The QA finding lists the root cause of this error as " inattention to detail". The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

FCR 89 0179

.This modifiestion replaced existing piston operated dampers 1 PO 5406 and 1 PO 5407 in the discharge of ECCS pump room exhaust fans No. I1 and 12, respectively, with gravity dam _ pers. In addition, solenoid valves 1 SV 5406 and 1 SV 5407, pressure control valves l-PCV.5406 and 1 PCV-5407, instrument air lines and accumulators were removed. Also, limit switches 1.ZS 5406A & B and 1-ZS 5407A & B, which are installed on the dampers and indicate fully open or fuhy closed at control panel 1C34, and all associated circuits and raceways were removed or spared. The indicatmg function was added per FCR 841088, item No. 51 to meet the Reg. Guide 1.97 requirements.

In addition, this modification replaced the existing flexible connections between the existing aiston operated dampers and duct work with continuous molded Dexible connections aetween the new gravity dampers Imd the existing duct work.

The dampers are classified as safety-related, where as the Dexible connections are noa-safety-related.

This modification was made for the following reasons:

L To reduce the probability of a malfunction of equipment, the piston operated dampers require a source of compressed air, solenoid valves, and associated controls; whereas, the only external means necessary to operate the gravity dampers la the static differentie ipressure across the dampers which is provided by the operation of the exhaust fans.

- 2.- To reduce the load on the salt water air supply, which results in spare capacity for future safety-related use,

3. To maintain a consuer.t u tem configuration with the Unit 1 ECCS pump room  :

- exhaust system. Engineu up er the replacement of the piston operated dampers on the discharge of the Urp . ECCS pump room exhaust fans was issued under this-FCR.

In addition, this modific': ion replaced the existing flexible connections between the exiting liston operated dampers and duct work with continuous molded flexible connections

, aetween the new gravity dampers and the existing duct work. The existing flexible L connections develop leaks at the overlap folds over a period of time. The use of continuous Evaluation concluded molded there was no flexible unreviewed connection will or safety question eliminate thisTechnical change in the problem. The Safetk'pecifications.

W,- , , ,v-, -,w,-, .-~..w..-

,,w ,-----~,--%--.-w,-m-e-.m.- , ,, + ,w ym . _ , _ y ~~e r w .- - , - .- - -

10CillW1.59 Annuallkpatt PagtR FCR W 10 This safety evaluation was written to determine if an Umeviewed Safety Question exists due to the lack of the rated dampers in the ventilation ducte where the battery toom ventilation system (supply and exhaust) penetrates twu barri,trs. Soccifically, dampets are not installed in the barrier between the Unit 1 Cable Spreadinj lloom and Cable Chase til and the i barrier between Cable Chase !!) and Cable Chase lA. If no USO exists, then no dampeis will be installed.

It was discovered that a total of four fire rated dampers had not been installed where the 125VDC limtery Room ventilation system supply anu exhaust ducts peneirate two baniers.

A review of commitments regarding the rating of these barriers revealed that ilG&li had indicated that at least 11/2 hour rated dampers would be provided in ventilation i penetrations. Ilowever subsequent guidance provided b addressed deficiencies in barriers which maybeeventuall)y discoveted andthebarriers NRC which in Generic Lette 1 must undergo chan in this Generle Letter, (Section i 3.1.2 of Enclosure I')es as a result of plant modifications,, the NRC directs that an  ;

fire protection engineer to determine if a barrier which is not completely scaled ftom Hoor-  !

to ceiling will still provide adequate separation. ,

The Generic Letter provides guidance on Future Chances. This guidance recommends that an evaluation be made in conformance with 10CFR50.59 to determine whether an  :

1.'areviewed Safety Question exists primarily in the context of Appendix R compliance. The Safety Evaluation concluded there was no umeviewed safe:y question or change in the Technical Specifications.

FCR 90 0020 i 1

This change removed the function of the lodine Removal Unit (IRU) Dousing System. This change isolates the IRU Dousing System when either the containment spray system or the -

contamment iodine removal system is required to be operable by the Technical S ecifications (Modes 1,2,3, and 4). Manual valves SI 4949, SI 4950, SI 4951 SI 4958 St.

4359, SI 4960 will tovide the isolation. The Main Control Room switches for dousing valves SV 4952, SY-)4953, SV 4954, SV 4955, SV 4956, and SV 4957 will ma provide any function when the manual valves are closed. The control circuits for SV 4159 and SV 4160 will be classified as NSR. CV 4159 and CV 4160 will retain a safety related pressure l

boundary function. In Modes 5 and 6, the manual valves may be open to allow the dousing i syste,m to be functional during iodine removal unit maintenance to provide fire protection if j required. <

The IRU temperature indication and monitoring equipment is reclassified as NSR as the temperature lunction is not required nor is its functmn related to any automatic safety actions. Since an analysis shows that IRU overheatin ' is not credible, the temperature instrumentation is not needed as an indication that wouk trigger operator action.

Coincident with a LOCA or post LOCA, a single failure of the limit:n; component (the loss  !

of a diem generator or an IRU fan failure) will not result in the charcoal ignition or desorptiu temperature being reached. Such a failure during a LOCA will not cause an -;

increase m 'ffsite does consequences through the emoval of the dousing system function.

+

A detailed analysis was performed (NUCON Report No. 613G021/01, dated 1/19/90 and i Supplement 1, dated 7/25/90) to assess the maximum charcoal filter temperature possible L during a maximum Sypothetical accident (LOCA). System performance parameters (spray iodine removal, charcoal efficiency data) were obtamed from FSAR C1 apter 6 Source ,

i

10CFil50.59 AnnuttlRrport PagtM term and isotope distribution data were obtained from Chapter 14. Other inputs came from design drawings, manufacturen' data, and use of the computer code along with the input and assumptions used. The report concluded that under postulated accident conditions iodine desorption or charcoal filter fires are not credible, 11G&l3 did not conunit to monitoring litU charcoal bed tempeinture as ptut of the llegulatory Guide 1.97 submittal, nor has llG&B conunitted to having the dousing system perform a safety telated function. The Safety livaluation concluded there was no unreviewed safety question or change in the Technkal Specifications.

FCit !&llo The activity associated with FCil 90110 made permanent eleven CCI.ll7 Temporary Modifications of the Reactc./ Vessel Level Monitoring System IIVLMS, a ilented Junction Thermocou ile System . IljTCS). A secondary purpose of this(activity is to provide the e

for future c1anges to the RVLMS whkh are identical to the types of wiring modifications described in FLR 90110 and evaluated herein. Specifically, the following types of wiring modifications to the RVLMS's electronics are evaluated:

1. Revise the reference thermoeouple med for some sensors which have a failed reference junction thermocouple lead),(to restore operability of those sensorstopen or shorted U 2.- Jumper failed heated thermocouples (either due to a failed heater or a failed thermocouple- to adjust thermocouples immediatelv above it) so that the alarms n heca(if- possible, cleared tothe to restore thechannel operable to sensor operability'and a that the RVLMS indicators consemitively alert the operator to level changest and

3. Replace failed heaters with a dropping resistor to provide proper power to other -

heaters in its series string and allow t be remaining sensor string to operate properly.

The above changes are within the design intent of the sy> tem and document the current "As.

lluilt" condition of the plant. If a new IlJTCS probe is ever instal!cd, these changes would have to be re evaluated. All the above changes are performed by lifting leads and (noviding

-jumpers and/or replacement heater dropping resistors in the back of the iIJTCS clectronics cabinets.

iring modifications to the lijTS's This activity electronics makes permanent. Thethe existing closes oct elevenCClCClll7 117temporary w's on Units 1 and 2 com activity also documented the CCI.ll7 wiring modifications on the desi;n documentation for Units 1 and 2. In addition, the activity provided for future chnnges to t ie RVLMS which are identical to the ty )es of wiring modifications in FCR 90110. The changes to the RVLMS restore some disa iled sensors and clear alarms from inoperable sensors to allow operable sensors to alarm if required. This FCR allowed an RVLN S sensor with a disabled reference thermocouple to be operable for the purposes of meeting the o ierability requirements of the newly approved Technical Specifications No. 3/4.3.3.6 (Unit Amendment 147; Unit 2 Amendment 128). The Safety Evaluation concluded there was no unreviewed safety question or change in the Techmeal Specifications.

i-l

i; . i JLCFR50.59 Anmudikport l'ogda s t FCR 40117 i Revised the description of the subcooled margin monitor in UFSAR 7.5.9.1, Revised the first line of 7.5.9 to change the acronym for the subcooled margin monitor from Shihi to SChiht. Revised the second to last line of 7.5.9.2 for the same reason. .

The stated range and accuracy were not correct, and there is only one ;)ressure input per channel. This change also provides clarification of the lack of elecitica separation of the .

Signalinputs within each channel.

i Additionally, both the subcooled margin monitor and the shutdown margin monitor v,ere i referred to us SMhi. The subcooled margin monitor has been changed to SCMhi, The a Safety Evaluation concluded there was no unreviewed safety question or change in the  !

Technical Specifications.  ;

I ECLuh2iG This modification provided a new Ce*rosion Products Sampler (CPS) in the Turbine Plant Sample System (TPSS). The CPS is a passive device that allows a continuous sample stream to flow through a filter, where corrosion products me trapped for laboratory analysis.

FCR 91246, Sup a Radiological & Chemical Technology, Inc. (RCT) plement CP5, at each 0 provided engineering

'l PSS auxiliary panel 1 for installing /2 T21 A), connect main feedwater header, so that corrosion products sampli(n Corrosion product sampling only existed via a portable test rig.g is permanently TPSS, the CPS will permanently enable enhanced corrosion protection monitoring for the mma feedwater system, thereby providing added assurance that the feedwater chemistry >

and secondary ride corrosion rate remmns within acceptable limits. The Safety livaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

FCR 91208 This evaluation is being performed in response to NCR 8474, to ensure an unreviewed safety $1uestion does not exist when the following statement in UFSAR Section 9.8.2.3 is deleted, 'The relative humidity in the auxiliary duilding isn't likely to exceed 50 percent, ,

consequently, the nmximum iodine removal efficiency should be realized" and the following statement is added: " Periodic testing is conducted to ensure maximum filter efficiency of 909h or greater is maintained."

l No documentation has been identified to confirm the value of 5096 relative humidity in the auxiliary buidling. This value is not used in design bases in the accident scenarios of the UFSAll or in liVAC design base calculatiota for the auxiliary buidling. Further, the EO ,

Design Manual lists the maximum relative humidity for normal and I.OCA situations as 7096, For main steam linc breaks (MSLil) and high energy line breaks (llELil) the auxiliary building relative humidity is listed as 1009b. T ie charcoal filters are not required to ,

be in operation during IIELil and MSLllincidems. The Safety Evaluation concluded there '

is no unreviewed safety questions or change in the Technical Specifications.

J m _ __ . - , _ , _ . _ _ . - , _ _ . _ _ , ~ _ _ . _ . _ _ _ , _ _ _ , _ , _ _ . _ _ _ _ . _

10CFR50.59 AnnuallknMt Pag d <

FCit 91248 FSAR Section SA.3.2.2 states that ASMli Code Case N 411 will be used when performing new analy es or reconciliation on Seismic Class 1 pig!i ng sptems. This activity changes the requirement to allow Uplimmt use of ASMll Code Case N 411 when not using Regulatory Guide 1.61, in order to take advantage of increased piping system damping values.

l IlG&II's original 6 tent was to allow ASMli Code Case N 411 to be used as an nMinnal I method of pi? ing analysis. The Safety livaluation concluded there was no imreviewco saf ety I question or c 1ange in the Technical Specifications. l l

FEC 83 47-03  ;

This activity revised the description of the boronometer in the UFSAlt Chapter 9 to correctly describe the range and accuracy as modified by FCR 834X)47.

FCR 834X)47 installed new electronics and associated hardware to upgrade the baronometer to a temperature compensated, digital / analog display with an overall range of 0 to 5(XXI ppm boron concentration. The Safety livaluation concluded there was no unreviewed safety question or change in the Technical Specifications.

FEC 89-01-476 This minor modification added an isolation valve, pipe nipples, and pipe cap to valve 2CC. ,

118 sight glass isolatlan valve on the #21 camponent conhng chemical addition tank. 'he addition of this valve and piping made the design consistent with that of the Unit I chenucal addition tank.

This evaluation also includes the revision of FSAR Figure 9-6 to refleet the Unit 1 as built design. The modification identified above is the current as built Jesign for Unit 1. Figure 9-6 is being revised to reacct that current as built design. Since the design descaibed above currently exists for Unit 1, all discussions concerning the modification for Unit 2 would also apply to Unit 1.

There was no method which facilitated drainin;; the Unit 2 chemical addition tank to allow the addition of chemicals. The addition of the isolation valve and pipe allows venting of the tank by removing the pipe cap and opening the isolation valve, l'he Safety Evaluation concluded there was no unreviewed safety question or ch!mpe in the Technical Specifications.

FEC 00-01955 This activity removed pressure indicators 1/2 PI 294,1/2 Pl.295,1/2 Pl.290,1/2 PI 291, and 1/2.Pl.292 from the suction side of the 1 PSI and llPSI d the tubing where the iressure indicators tie into the s P&lD M.74 FSAR figure 6-1) and

?&lD M 462 (FSAR figure 6-10)ystem. In addition, pumps an c were also modified to tellect as ht It conditions identified during a system walkdown. These P&lD drawing discrepancies with as built conditions are the following:

1. 1.acations of 1.PP.302W and 1.PI 295 are reversed. -

2. Delete 1 PP-30lV and 1 PP.30lW.

9 10fFRSO.59 Annual RmoIL Pag d

3. Show 1 PP 30lY connected to Hanged spool piece.

4. Show capped tube connected to Ganged spool piece.

5. l.ocations of 2 PI 295 and 2 PS 302Y are reversed.

Pressure indicators on the suction side of the 1. PSI and flPSI pumps both units are used only during testing as required by existing STPs. Installed Pts were b(ecoming a ) me problem because of their tendency to drift out of calibration. Using high accuracy Pls when i needed as opposed to the permanently installed Pts will alleviate this utuation.

This activity involved the removal of pressure indicators that are not used during rmrmal operations. The system pressure boundary is : .aintained since existing isolation valves will remain in their normal operating (shut) condition. These isolation valves are identified as the safety related boundary. The modification lef t at least one Pressure Point (PP) on each pump that can be utilized for aump testing. This activity also modified P&lDs M 74 and M.

• 62 to reneet as built contitions identified during a system walkdown. The Safety Evaluation concluded there was no unreviewed safety question or change in the technical +

specifications.

FEC 90-01 ll5 DCR 891695 and DCR 901377) were initiated to revise P&lDs M 64 Sil FEC 90-01

1. Auxiliary Hui115 (lding Ventilation System. (DCR 90-0181 revises the Johnson Co Tubing / Actuator drawings to rencet the changes made on M 64 Sh.1)

1. Each of the FuelIlandling Area iIVAC units' dampers, O.PO 5414 and O-PO 5415,are shown on P&lD M 64 Sil 1 as havin.; a single pneumatic piston actuator.This activity proposes that each dam per ic shown as having two pneumatic piston actuators. This is the as built condition. There is no vendor documentation s pecifying the number of piston actuators required; however,  !

cach of the two dampers is supplied with two blade actuating rods. ,

2. Each of the Fuel Pool Exhaust Filter's dampers, O.PO 5417 and 0 PO.5418, are shown on P&lD M 64 Sil 1 as having a single pneumatic piston actuator.  ;

This activity proposes that each damper be shown as having four pneumatic piston actuators. This is the as built condition and is supported by the filter vendor's design drawing.

. The dampers described above are safety related. The proposed activities do not affect the

. function or operation of any safety related system, structure, or component. A 50.59 Evaluation is necessary since the changes create revisions to P&lD, M 64 Sil I which is FSAR Figure 9 21. The Safetv-Evaluation concluded there was no unreviewed safety question or a change in the techn'ical specifications. i FEC 90-01 10,,9 FEC 90-01 1054 authorized the addition of reinforcement bars to the Fuel Transfer Tube Blind Flange. The modification was intended to address minimum wall thickness concerns with the bhnd Dange.

i i

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4

, i if)p0.59 AmntilRenort -

Pqtjl I

l /> $ll;4 was done because FSAlt figure 914 is being changed as a result of FEC  :

% 0t.1054.  ;

i l

Prior to commenMng the U2 ILitT, the U2 Fuel Transfer Tube Illind Flange had to be i installed. During the reinstallation process, questions arose concerning the mmimum wall thickness of the blind flange. IlOLi'EC INTEllNATIONAL was contracted to calculate l the minimum wall thickness using finite element anal .

Scheduled to take several weeks and previous " hand"ysis.

calculations Thewas showed there finite element anal!

a high probability the existing blind flange was inadequately sized. Therefore, llG&B decided to I make a conservative modification to the blind flange immediately. The modification would t add cross members to the blind flange compensating for the 30tentialinadequate thickness  !

and allow the ILitT to begin without the delay of wmting for t le finite element analysis to be ,

completed (Modification details were obtained from IlOLTEC INTEltNATIONAL and  !

installed by way of a Provisional Modification). The Safety Evaluation concluded there was no unreviewed safety question or change in the technical specifications. ,

MASE 90 4  ;

Since the Service Water System provides essential functions, it is appropriate to establish allowed contamination limits, so that operation of the system may continue following a .

establishing a range possible of allowable contamination event.

contamination levels forThe proposed normally limits provide flexibility,d systems w non contaminate ,

continued operation is acceptable. '

IE llulletin 80-10 states the re l which become contaminated,gulatory requirements if continued operation of for theoperating system asnon radioactiveis systems contaminated ,

3 necessary, an evaluation is required to determine whether continued operation is neceptable. The evaluation considers the level of contamination, potential releases to the environment, the relationship of such releases to the radioactive effluent limits of 10CFil20 and the facility's Technical Speci0 cations, and the environmental radiation dose limits of 40CFR190. The evaluation sets forth the basis and criteria on which the determination was made. I This evaluation uses the methodology and parameters used in the ODCM to calculate off.

site doses due to both accident and normal releases from the system.

1 Chemistry Department rocedures already require periodic sampling of the Service Water System to provide earl identification of cross contamination. Lnteria established in this i evaluation assure that t ie regulatory limits of 10CFil20,10CFit50, and 40CFil190 are met.  ;

The allowable contamination levels are set such that either accident or normal releases from the Service Water System will contribute less than the Technical Specification limits for off.

site doses. Furthermore, the limits assure that releases to the environment will not contain ,

radioactive material in concentrations greater than the MPC. This low level of contamination still allows operational flexibility while ensuring that the incremental increase  ;

in off site doses remains low enough that the Technical Specification limits will not be exceeded. The Safety Evaluation concluded there was no unreviewed safety question or i change in the technical specifications.  !

I MASE 90-5 ,

Since the Demineralized Water System provides essential functions, it is appropriate to i establish allowed contamination limits, so that operation of the system may continue following a possible contamination event. The proposed limits provide flexibility,

M . N b b hUlllRI $ {POlt - _ _ _

UUMVM establishirig a rtinge of allowable cotitatitiriatiori levels for riorinally rioti-coritaininated systems % herein colitillued operillioll is acceptabic.

11! llulletin 80-10 states the regulatoiy requirements for o erating non radioactise systems whicli becorne containituited. 11 corititiued operatiori o the systerii as contaniinated is necessa ry, an evaluatiori is !cquired to detelliurie %hether colitiliued operatioll is acceptalile. The evaluation considers the level of contamination, potential releases to the environment, the aclationship of such releases to the radioactise ef fluent limits of 1(CFl!20 and the facility's Technical Specifications, and the emitomnental radiation dose limits of 40CFil190. The evaluation sets forth the basis and criteria on which the determination was l made.

This evaluation used the inethodology and pasarneters used in the ()DCN1 to calculate off-site doses due to both accident and normal releases from the system.

Chemistry Departmem procedures aheady require periodic sampling of the Demineralized Water System to provide early identification of cross contamination. Criteria established in this evaluation assure that the regulatory linuts of 10CFil20,10CFit$0, and 40CFilloo are met, The allowable contamination levels are set such that either accident or normal releases from the Demineralized Water System will contribute less than the Technical Specification limits for of f site doses. Furthermore, the limits assure that icleases to the environment will not contain radioactive materialin concentrations greater than the NiPC.

This low level of contamination still allows operational flexibility %hile ensuring that the incrementalincrease in off site doses remains low enough that the Technical Specification limits will not be exceeded. The Safety Fraluation concluded there was no umeviewed safety question or change in the technicalspecifications.

hjM 004 Since the Plant lleating System provides essential f unctiom, it is appropriate to establish allowed contamination liniits, so that operation of the system may continue following a possible contamination event. The proposed limits provide Desibility, establishing a range of allowable contamination levels for normally non contaminated systems %herein ' '

continued operation is acceptable.

IE Bulletin 80-10 states the regulatory requirements for operating non-radioactise systems which become contaminated. Il continued operation of the system as contaminated is necessary, an evaluation is required to determine whether continued operation is acceptable. The evaluation considers the level of contamination, potential seleases to the environment, the relationship of such releases to the radioact ve eftluent limits of 10CFit20 and the facilit/s Technical Specifications, and the environmental radiation dose limits of 40CFl(190. Tne evaluation sets forth the basis and criteria on which the determination was made.

j.,

This evaluation uses the methodology and parameters used in the ODE #hi to calculate of f-site doses due to both accident and normal releases f rom the systems.

Chemistry Department procedures already require periodic sampling of the Plant lleating System to provide early identification of cross-contamination. Criteria established in this evaluation assure that the regulatory limits of 10CFil20,10CFitSO and 40CFl(190 me met.

The allowable contamination levels are set such that either accident or normal releases from the Plant Heating System will contribute less than the Technical Specification limits for of f-site doses. Furthermore, the limits assure that releases to the environment will not contain radioactive materialin concentrations greater than the N1PC

l 10CFR50.59 Annuallyjwit page_J9 This low level of contamination still allows operational 11exibility while ensuring that the incrementalincrease in off site doses remains low enough that the Technical Specification limits will not be exceeded. The Safety Evaluation concluded there was no unreviewed safety question or change in the technical specifications.

M ASE 40-7  ;

I Since the Nitrogen System provides essential functions,it is appropriate to establish allowed  !

contamination limits, so that operation of the system may continue following a possible  !

, contamination eventi The e of  !

allowable contamination levels atoposed limits'tcontaminated for normally provide flexibility, systemsestablishing a ranf'nued wherein cont operation is acceptable.

IE Ilulletin 8010 states the regulatory requirements for operating non radioactive systems whleh become contaminated. If continued operation of the system as contaminated is necessary, )n evaluation is required to determine whether continued operation is acceptable. The evaluation considers the level of contamination, potential releases to the environment, the relationship of such releases to the radioactive efiluent limits of 10CFR20 ,

and the facility's Technical Specifications, and the environmental radiation dose limits of l 40CFR190. The evaluation sets forth the basis and criteria on which the determination was  ;

made. l This evaluation uses the methodology and pmameters used in the ODCM to calculate off. -

site doses due to both accident and normal releases from the system.  ;

chemistry Department procedures already require periodic sampling of the Nitrogen .

System to provide emly identification of cross contammation. Criteria established in this esaluation assure that the regulatory limits of 10CFR20,10CFR50, and 40CFR190 are met.

The allowable contamination levels are set such that either accident or normal releases from the Nitropen System will contribute less than the Technical Specification limits for off site doses. 1 urthermore, the limits assure that releases to the environment will not contain radioactive material in concentrations greater than the MPC. This low level- of '

contamination still allows operational flexibility while ensuring that the incremental increase in off site doses remains low enough that the Technical Specification limits will not be exceeded. The Safety Evaluation concluded there was no unreviewed safety question or change in the technical specifications.

i MM 91012 015 0 Minor Mod 91012-015 0 installed high accuracy transmitters to monitor the pressure drop across the tube side (saltwater) of the service water heat exchangers (SRWHX).

The condition of the SRWilXs ultimately affects the ability of the plant to reject heat Ic .

Chesapeake 13ay. As surface of the heat exchangers are progressively fouled, diffe xntial pressure increases and heat transfer is impeded. The gauges that were installed on the saltwater inlet and outlet of the SRWilXs,1/2 PI 5209, $210,5211, and 5212, did not have t the resolution or accuracy required for performing this function. The Safety Evaluation _

concluded there was no unreviewed safety question or change in the Technical  :

Specifications.

L

• - Pagt20

.1M'EMO.59 Annualltswit-Mht9MILMIMLOhnL2) hibi 91019.du 3 0 Olnitl)

This activity u pdates P&lDs hi 53 Sh. 2 (hiinor hind 91019 013 0), h1479 Sh.1 (hiinor hiod 91019-0: 2 0) and other impacted documents to sellect the "as built"(ondition of the Compressed Air System (Instrument Air and Plant Air). The P&lDs are requir' to be updated by incorporating the f ollowing:

On P&lD hi 479 Sh.1 (FSAll Figure 9 23A), show 3/4" diameter Plant Air supply line to Chemistry 12h at EL. 69-0" with isolation valve 2 PA 282 in the noimafly open position.

On Unit 1 P&lD hi 53 Sh. 2 (FSAll Figure 9 23-2), show 1" x 1/2" reducing bushing downstream of valve 1 IA 219 in lieu of hose connection. Further, change Dow ditection from 2 way flow to 1 way flow on Instrument Air piping header valves 11A-218 and 11A 220. The Safety livaluation concluded there was no unreviewed safety question or change in the Tec'hnical Specifications, hlhi 91043 005 0 This activity added permanent loop seals to ,ondenser air discharge header drain lines containing valves 1 Call 156 and 1 Call 157,or Unit 1 and 2 Car 156 for Unit 2 in the Condenser Air itemoval System. The permanent loop seals consist of piping, valves and caps, in order to prevent air from escaping or entering the condenser air discharge header through the train lines while the moisture from the air and other noncondensible gases are removed / drained during normal operation, permanent loop seals are added to the existing drain lines containing valves 1 Call 157 and 2 Call 156. A 50.59 Safety Evaluation is required because FSAll Figures 10 7 (Unit 1 P&lD hi 50) and 1012 (Unit 2 P&lD hi 451) are revised to show the drain lines with loop seals. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical $pecifications hihi 91-045@6-0 This activity relocated the non-safety related lustrument Air supply tubing run for the Steam Generator Fred Pump Turhine (SGFPT) No. 22 ilydraulic Assembly, Panel 2T30. The modification taps into an Instrument Air source, which is close (within one foot) to the Ilydraulie Assembly, and routes new supply tubing to the panel. The revised air supply includes a filter /reputator, a shutoff valve and a maintenance connection for a temporary air supply and is consistent with the original air supply configuration.

The revised tubing run has less pressure drop and provides an improved air supply to the SGFPT speed changer valves.

The air supply configuration is as follows:

The air supply is 20 psig, regulated by a new aitset 2 PCV 8040ll, located just outside 2T30; the supply air tubing w"ill be 3/8" O.D. and 1/4" O.D. by 6" long stainless steel with a ,

wall thickness of.065 and a totallinear length of 4 feet.

EER5R59 Aimuauknat _ Pagtn e

Minor Modification 910454%0 was written due to degraded performance of the SOFPT No. 22 speed chan er valves 2 CV 8040 and 8041 located hi the Turbine lluilding, SOFPT llydraube Assemb Panel 2T30, on lilev.12'.)". The change is to increase the sire of the air supply tubing to th 22 SGFP speed control valves.

The previous configuration for the air supply to the subject valves was as follows:

The supply Pneumatic air is Panel Assembly 20 psig,2T29;and is regulated by aliset 2 PCV.8040 located in e the supply air is routed from 2T29 to the liydraulle Assembly Panel 2T30 (where the valves are located);

e the supply air tubing is 1/4" O.D. stainless steel with a wall thickness of .065" and a total knear length of approtimately $$ feet.

This configuration imposed a large pressuie dro s on the alt supply and adversely af frets the response and operation of the speed changer va ves. A tem porary test configuration for the air supl?l p using 3/H" plastic tubmg approximately 20 feet in ieng was routed from 2T29 to 2T30. Ihe performance of the speed changer valves greatly improved during the test, and proved that the cause of the problem was, in fact, the restrictive and overly lengthy supply air tubing run. The Safety livaluation concluded there was no unreviewed safety question or a change in the technical specifications, hihi 91-059-005-0 Minor Modification 91059 005 0 (Unit No.1) and 91059-0051 (Unit No. 2) were initiated to evaluate / repair the Containment Pressure instrument tubing systems.

The pressure sensing line installation were analyzed using the MII 101 stress program and building differential movements which were reduced usmg a more current methodology which differs from the design basis analysis method provided in the FSAlt. As a result of the analyses, the activity removed supports from the instrument tubing systems.

When the original Cahert Cliffs seismic analysis was performed in 1970 for the main building structures, such as- the Containment and Auxiliary Buildings, a simplistle mathematical model was constructed for input to the original computer analysis.

The model for the building structures used excesshely conservative soil damping values.

The conservatism inherent in the original _ analysis can be quantified by evaluating the original results using more current define soil damping.

These evaluations vield building displacement values which are considerably less than those originally calculate'i! These reduced displacement values have bei n used as input to the tubing stress analysis.

This approach, which utilizes more reasonable damping value, has been previously used on Calvert Cliffs to formulate to response to NitC 111 flulletin 8011 with regard to the seismie qualifications of masonry block walls.

The 50.59 ws performed to document and support the approach based on the revised building differ ntial movement and is applicable to the Unit No. I and Unit No. 2 Containment Pressure instrument tubing systems.

I

• KFil50.59 Annunmepari Pagt22 A 50.59 Evaluation is necessmy since the tubing system analysis is Imsed on seismic )

movement values which were reduced using the more current methodology which differs '

from the design basis analysis method provided in the FSAlt. The Safety Evaluation i concluded there was no unreviewed safety question or a change in the technical specifications. ,

MM 9106MKil 0 i This activity is to omit reference to 250 #Jlli crane in l' Salt Section 9.7.2A.

MM# 9106MXil.0 is removina the 250 # Jill crane under the fuel pool restoration project. ,

The Safety Evaluation concluCed there was no unreviewed safety question or a change in the technical specifications. j PitOCI!DUltE llE 50 Procedure llE 50 has been initiated to Containment Sump Screen in Modes 4,5,when 6, and place a temporary Defoeled, whenevercover over the top of th mainte:mnce '

activities which could compromise the cleanliness of the sump screen box are to .iccur in  ;

boards, and a thermal containment.

barrier cloth (Mech. This temporary No. 55373) cover fastened cover securely will be to inthe thescalfo form of scalfolding,lding. The ,

be raised a minitnum of one foot above the sump screen so that flow through the top of the sump screen will not be impeded by the addition of this cover. Prior to entering Mode 3 the scaffolding is to be disassembled, and any debris found lying on the cover is to be removed from containment. i This 50.59 Safety Evaluation was written because of the impact this procedure would have on a structure described in the FSAlt. The Containment Sump Screen ilox construction ~

and the effective flow area through the screen are described in the FSAlt. Also , the sump screen is depicted in a drawinj contained in FSAlt. Since this cover can be considered to be mrt of the overall structure o: the sump screen the description in the FSAll is being altered.

Sinally, to pro)erly determine whether a 50.59 was required or not it _was necessary in thh:

case to do all t ie research required for a 50.59. In the fmal analysis it was still undetermined  ;

whether a 50.59 was required; however, because questions could arise concerning the safety significance of this procedure it was deemed prudent to write a 50.59 so that these concerns ,

could be identified and then formally addressed.

During maintenance activities in Modes 4,5,6, and when Defueled, debris may fall onto the sump screen. The debris may be of such a size (long and narrow) that it is able to pass through the sump cage wire mesh, and enter the contamment recirculation Suction lines.

This debris may either block these suction lines, or be transported to the ECCS pumps

- (llPSI, LPSI, and Containment Spray) where upon it may cause the malfunction of one or >

more of these pumps when these pumps are required to operate by taking suction from the containment sump. The main purpose of this activity is to prevent debris from collecting in ECCS suction piping the suction piping used during containment recirculation cover over the sump (enge. This will prevent debris from entering mment sumpthe conta) by recirculation lines from the direction in which it is likely to enter. Debris entering the Containment Sump Ca e by falling through the side is not considered a likely event. The Safety Evaluation conc;uded l there was no unreviewed safety question or a change in Technical Specifications.

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