ML14037A175

Independent Spent Fuel Storage Installation, Biennial Commitment Changes and 50.59/72.48 Report

ML14037A175
Person / Time
Site:	Columbia
Issue date:	01/21/2014
From:	Hettel W Energy Northwest
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
GO2-14-01 0
Download: ML14037A175 (8)

Text

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W~ilu 0h Nows Vias PrMeidL OpUilft P.O. Ba 00. Meg DM PE23 Pioh~ani4 WA 0935-0968 Ph. 509-377-311 F. 809477.467

,.1" ww0i"W0Wgyqnoehwgs00am January 21, 2014 GO2-14-01 0 * .?. r" ". '

10 CFR 50.59(d)(2) 10 CFR 7248(d)(2)

U.*. ,Nuclear Regulatory Commission ATTN:. Doc*un Control Desk W, D.C. 20555-00 Subject CQLUNWA GeNERATMA STATON DOCKET NO. 50497

,IWEPENEIT6PENT FUEL STORAGE INSTALLATION, DOCKET NO. 4 6IEr01L COMWiMENT CHANGES AND SO.5172A6 REPORT

Dear Sir or Madam:

Attached is the Columbia Generating Station 2012 - 2013 Commitment Changes and 50.59/72.48 Repofto This report is submitted pursuant to 10 CFR 50.59(d)(2), 10 CFR 72.48(dX2), and Guidelines for Managing NRC Commitment Changes (NEI 99-04).

There are no commitn'iets being made to the'NRC by this letter, however, two existing commitments have been changed.,

Hf you have anty questions or desire additional information pertaining to this report' please contact Me. L.L. Williams,. .LiPnsingS masor, at (50a)377-8148.

t r t Czoe -1,,e WG.Hettel '

Vice President, Operations ,

Attachment:

Biennial Commitment Changes and 50.59/72.48 Report cc NRCRegonIV Administrator NRC NRR Project Manager

-.aNRC AJ lac- Senior BPA/1399 Resident Inspector/988C "" i

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if BIENNIAL COMMITMENT CHANGES AND 50.$W72 REPORT Attachment Page 1of 7 10 CFR W59 Charge, TestsK n E#

This section contains a brief description of any changes, tests, and experiments, including the Summary of the evaluations for activities implemented during 2012 and 2013 that were assessed pursuant to 10 CFR 50.59 requirements.

Energy Northwest evaluated the changes summarized below and-determined prior NRC approval was not required.

5059404001 "810830 ONLINE NOWE CHEMISTRY ( )OLNCY This change will Implement the OnA.Une Noble Chemistry (OLNC) process at Columbia Generating Station (OGS), resulting In the ability to Inject the noble metal platinum In the reactor coolant system online instead of only during outages. OLNC was developed by GE Hitachi (GEH) as the next generation noble metal technology and represents an improvement over the classic Noble Metal. Chemical Application (NMCA) currently in use.at CGS.

This 50.59evaluation is focused on effects of OLNC on fuel performance and its consequences with respect to 10 CFR 50.59 requirements, The 50.59 evaluation on plant systems other than fuel is given in Evaluation 5059-10-0002. From extensive fuel inspections since 2005, it is concluded thatVOLNC does not degrade the fuel cladding or the other fuel bundle components below their design allowable specifications. The structural integrity of the fuel bundles will be maintained. There is no change to accident source terms. Therefore, OLNC wiln not result In more than a minimal increase in the consequences of an accident. No fuel failures have been observed from either classical NMCA or OLNC operations (Ref 4). OLNC fuel inspection results have been even better than those from classic NMCA. No new fuel failure modes from the OLNC are anticipated. It is also concluded that OLNC does not result in a design basis limit for a fission product barrier as described Inthe CGS FSAR being exceeded or altered.

Implementation of the OLNC does not require any changes to the method of evaluation described Inthe FSAR in establishing the fuel thermal limits because fuel integrity Is not affected by OLNC and all the methods of evaluation described Inthe FSAR remain valid. Since fuel itself Is not an accident initiator or an SSC (it is the source term), there is no effect on the frequency of an accident or the malfunction of an SSC or the consequence of a malfunction of an SSC.

5OW104O "1ECa08 OINUNE NOSLE Brlef*Desci....... ....

CH"MTRY (*L.C)" '*-

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This change will implement the On.Une Noble Chemistry (OLNC) process at Columbia Generating Station (CGS), resulting in the abil"t .to inject the noble metal Platinum Into the reactor coolant online instead of only duringoutages. OLNC wasdeveloped by GE Nuclear Energy (GEH) as the next generatiornoble metat technology and represents an improvement, over the classic Noble Metal Chemical Application: (NMCA) currently in use -at.CGS.

W, BENCNIO. OUMMII OHA SW92A6 REPORT MC1ANUAND Attachment Page 2 of 7 Summary CA Elut-This evaluation focuses on the effects of OLNC on vessel internals. Injecting noble metal compounds into the reactor vessel does not affect the safety of operations or the health and safety of the public. OLNC application will be conducted during a period when the reactor is at high power and high core flow operating conditions. During such a period, the reactor coolant system Is in a high-energy state as assumed in the applicable safety analyses in the FSAR. This evaluation has concluded that the frequency of occurrence or consequences of a LOCA (recirculation or main steam line break) will not increase. Because of the benefitfofOLNC, which will reduce the potential of IGSCC, the frequency of occurrence of a LOCA will be reduced after the noble metal injecton. The frequency of occurrence of a postulated CRDA or FHA is not affected.

OLNC does not affect the results of the accident radiological analyses. Therefore, the frequency of occurrence and the consequences of accidents previously evaluated will not Increase.- Dufiiithe Ome.tpiod when OLNC application is conducted, the conditions within-the reactor coolant pressure boundary are such that no new scenario can be postulated that could result in fission product release. Therefore, there is no possibility of an accident of a different type than any evaluated previously. Because OLNC does not create any adverse equipmentinteraction, no new malfunction of any equipment important to safety'can be postulated during OLNC application. After the noble metal application Is completed, the teactor vessel, the reactor internals and some of the associated. primary pressure boundary piping and equipment will potentially become better protected from IGSCC, and thus, have reduced probabilities of failures and/or malfunctions.. Therefore, it is concluded that OLNC does not affect any safety-related equipment, the safe operation of the plant, nor the health and safety of the public.

6059-i1-0003 "'4L04ARR10.425.Spptfuel pool tempemtm Ilmtft" M~if Desgdotion The spent fuel pool temperature limit for an anticipated operational occurrence, loss of a fuel pool cooling division, in FSAR Section 9.1.3 and FSAR Table 9.1-6 is being revised from, 155F to 175F.

Summary of Evaluat12n The purpose of this change is to restore,the allowed temperature of the fuel pool during the anticipated operational occurrence (loss of one train of fuelpool cooling) to the original licensed value of 175F,-consistent with NUREG-0892 and RS-001.

In evaluating the change in temperature'on Columbia Generating Station (CGS), it was determined that COS was designedto operate with a fuel pool temperature.of 175F in the abnormal -condition of an anticipated operational occurrence of the loss of a fuel pool cooling division. All equipment that would be subjected to the Increased heat and humidity from a fuel pool with 175F water temperature is qualified for this environment.

The fuel pool and itsooling sy~tem'ar6,4iafidfor this temperature. The fuel in the fuel pool is qualified for much higher temperatures than 175F. The sumps In the pump rooms have. sufficient capacity to handle fuel pool surface evaporation., The water makeup system was designed to handle a boiling fuel pool. CGS .was granted its,.'.

a, BIENNIAL-C¢01ITMOT CANGS AND 10.W?2AS REPORT Attachment Page3of 7 original license based on an allowable temperature of 175F for the fuel pool, and evaluated against this temperature in NUREG-0892.

As this change is not adverse to a design bases function or a safety analysis, will not cause an accident or malfunction previously evaluated, will not create a -ew accident or malfunction not previously evaluated, and is not a departure from a.method of evaluation described In the FSAR, NRC prior approval is not required.

0,59.124001 "PC 011109 DtmdttvUon of RHR-V4-0A, RHR-V-75A, RHR-V-405, RHR-V-?&W" andBrief.Summay.. .,... , * '

RHR valves RHR-V-60A and RHR-V-75A are being deactivated by remofn the valve internals and reinstalling the valve pressure retaining components. RHR valves RHR-V-60B and RHR-V-75B are being replaced with a pipe spool piece. The code break for RHR valves. RHR-V-60M.and RHR-V-758 Isbeing relocated from downstream of RHR-V-768 to downstream of RHR-V-732.. This design change will also deactivate the remote control and indication functions for these valves including the auto isolation features associate with the Nuclear Steam Supply Shutoff System circuits for high drYWell pressure and low RPV level. With the deactivation completed, the use of existing manual valveseRHR-V-731 , or RHR-V-732 will be used when a sample Is to be taken at sample points SP-22A, or-SP-22B.

Summary of Evaluation The proposed activity is to change from a remotely operated sample valve with an automatic closure function to a manually operated sample valve with no automatic closure function. Since this change is from the use of a valve with an automatic closure function to a valve with a manual closure function, an evaluation is required.

The sample valve is a;3/4' valve which will be *aened and maintained open under operator control in mode 4 or 5, when the severity and potential for a design basis accident Is lowered. Since it Is asmall valve,, closure,is not time consuming or difficult,,

Oncelthe sample is taken, the valve will be dosed.

The manual taking of a water sample from-the reactorctoolant system will not result in an Increase in the probability of any of the evaluated accidents or malfunctions to occur.

Since the valve Is maintained under operatorvontrol and can be quickly closed, its use will not result in-a more than minimal Increase in theconsequences oftan accident or equipment malfunction. The taking of-a water sample does not result in a design basis limit for aifisuion product barrier to be exceeded or altered. Taking the sample under operator control will not result in a design basis limit for a fission product barrier to be exceeded or altered. Finy, the are no malfunctions oraccidents not already evaluated that could be caused by taking a water sample via a manual valve.

50.4-134001 ",'EC 19704 CYCLE 22 CONE DESH ..

Brief ummary The Cycle 22-reload design is analyzed using GE Hitachi (GEH)/Global! Nuclear Fuel

(GNF) Methodology. The proposed change is an update to the approved GEWGNF

f BIENNIAL COMMIMENT CHANGES AND IOwG2A REPORT Attachment Page 4 of 7 methodology or changes due to error corrections to computer codes associated with the approved methodology.

Summary gf EvluaiWon The changes identified in the approved GEH/GNF methodology as applied to the Cycle 22 reload analysis do not'represent a departure from a method of evaluation. The results from these changesare all either conservative or essentially the same. The methodology Complies with the associated NRC SER's.

5059-3,4003 "EC11062 INSTALL BLEED AND FEED SYSTEM FOR SW SPRAY PONDS TEMPORARY MODIFICATION FOR REDUCING TSS INSW PONDS" A temporary pumping system Is being Installed near the Service Water spray pond B as a means to blowdown the system to remove suspended solids from the spray pond water while at the same time making up to the ponds with water from the TMU system (temporary feed-and-bleed).

Summary of Evaluation -"

The installation and use of the temporary feed-and-bleed system to remove suspended solids from the UHS spray ponds have been evaluated for Impacts on SSCs important to safetyi,namely the Standby Service Water system and the Ultimate Heat Sink. The potential impacts evaluated relate to operator actions, seismic effects, spray pond water inventory requirements, tornado effects, ash fall and the effects of localized flooding.

None of these potential impacts result in a change in the ability of the Standby Service Water system or the Ultimate Heat Sink to perform all of their design functions. The proposed activity does not require prior NRC approval.

10 CFR 72.48 Chnge, Tesls, and Experiments No.changes, tests or experiments were conducted during 2012 and 2013 pursuant to 10 CFR 72.48 requirements. . -

However, the below change was incorporated in January 2008, and was not reported In subsequent biennial submissions.. This Was discovered after the 2011 biennial submittal. Below is a brief description of that change, including the summary of the evaluation for the activity Implemented pursuant to 10 CFR%7la48 requirements. Energy Northwest provides this summary to:restore compliance with. 10CFR 72.48(d)(2)'

reporting requirements.

7241474002 "Reduced Spent Fuel CoolIng Tim INE-0407-1 1100 CPR 72.212 Report, Rv. 4VPPM9.8.1S, Rev. 2"1,.'...

CoC Amendment 1 (Reference 4) Cask System activities:,

Reduce the minimum allowable cooling time for spent fuel stored Inthe ISFSI Reference 4 cask systems from 13 years to 5 years.

SIENNIAL COCMIHTET CtANGES AND 50.57IMAS REPORT Attachment Page 5 of 7 Reduce the number of Reference 4 cask systems analyzed for effluent dose contribution from 90 to 15.

CoC Amendment 2 (Reference 5) Cask System activities:

The minimum allowable cooling time for spent fuel stored in the ISFSI Reference 5 cask systems Is reduced,from 13 years:to 5 years for fuel exposure less than or equal to 45,000 MWD/MTU and within the restraints of the Reference 2 analysis. Cooling times less tan 5 years to no less than 4.747.years may be loaded provided fuel exposure is not greater than 42,390 MWD/MTU and the fuel Is loaded in the interor of the cask in Region 1.

The site boundary dose calculation in Reference I for the Columbia Generating Station (CGS) was revised in Reference 2. The site boundary dose calculation predicts an Increase in total dose consequence under accident conditions for the ISFSI when the cooling time is reduced from 13 years to 5 years. The dose consequences for normal and off-normal conditions do nrot increase. Further evaluation indicated that if the cooling time is further reduced to 4.747 years, the dose consequences for all conditions are bounded by the 5 year analysis with the constraints, described above. The selected fuel enrichment input was evaluated in Reference 2 to be bounding.

No change was made to the physical characteristics of the ISFSI cask systems.

An allowed spent fuel cooling time of 3 years has been reviewed and approved for the HOLTEC cask systems by NRC per Reference 5. Therefore, the Impact of the Amendment 2 Implementation on the HOLTEC FSAR screened out and this evaluation will only consider the Impact of the-change in spent fuel cooling time onrtheEnergy Northwest IOCFR 72.212 report.

The cooling time for spent fuel loaded in the ISFSI cask systems was reduced. Also, the number of Holtec Amendment I cask systems analyzed for effluent dose was reduced from 90 to 15. No additional loaded Amendment 1 casks will be stored at the Columbia Generating Station (COS). The site boundary confinement and radiation shielding analyses for the ISFSI at the Columbia Generating Station had to be revised to evaluate the impact of reduced spent fuel cooling time and the lower number of Amendment I cask systems. The reduced cooling time provides improved nianagement of future cask loadings. The reduced number of Amendment I cask systems reduces the contribution of effluent dose because. the Amendment 2 cask systems are leak tight and do not have an effluent dose contribution. The total predicted dose for normal and off-normal events, specifically plugged vents, either stayed the same or decreased. The total doae predicted for accident conditions, confinement boundary leakage, was less than a minimal Increase and was significantly less than 10%/6 of the margin to the limits.

The evaluation result is utilized to concloud that.prior NRC approval is not required to Implement the change.

P 1IENMAL COMWMITENT CHANGES AND SOAW A8 REPORT Attachment Page 6 of 7 Regulatry Commltment ChWa e (NEI 9944 P9 ees)

This section reports changes to regulatory commitments consistent with the information pertaining to Regulatory Commitment Changes (RCC) and is included pursuant to the NEI 99-04 criteria for reporting..

In responft to Geneic Letter 9-13 (RWC11070741)

The change proposed is tO perdot the air side inspections of the SW cooled cooling coils on a regularly scheduled basis determined from cooler Inspection trending history, instead of.an every two year commitment.

The intent of the original commitment was to visually verify the air side of the.room coolers was free cit dirt and debrs that could hinder heat exchanger performance and to clean as necessary. At Columbia, the majority of these air-to-water cooling coil heat exchangers support'utandby air handling units which only have air flowing through them during testactivities. "The RRA and WMA units are loceted in clean areas where dirt, dust, and other bontaminants are minimized. The DMA units have a'pre-filter and a medium efficiency fitter on the make-up duct to limit dkt and dust intrusion. None of the cooling coils have shown evidence of cooling water leakage.

Based on a review conducted of maintenance on air-to-water heat exchangers in the Service Water. system at Columbia, a regular inspection frequency Is adequate to prevent air flow blockages. Essential PM instructions have been updated to Include cleanliness expectations and to ensure all coil surfaces are clear of debris; To maintain the Intent of the original commitment, inspections should still be required on a frequency to verify heat exchanger cleanliness. However, the wording will be changed tO allow for eidended inspection frequencies based on multiple years of inspections that find no air side fouling that would challenge the ability of the equipment to meet its design function and do not require cleaning.

The original commitment stated:

Note: The original commitment from G02-90-017 was changed per RCC-1 10797 in 2004.

This is -the current commitment: ...*in lieu of precise thermal 4performance testing, Columbia will perform biennial preventive maintenance inspections on the air side of all'the'air-to-water cooling coils in the SW system to ensure the coil fins are clean and clear-and Columbia will perform biennial preventive maintenance Inspections on the air side of all the air-to-water service water cooling coils to ensure the coil fins are clean and clear.*

Commitment description has been revised to say:

"...in lieu of precise thermal performance testing, Columbia will perform regularly scheduled preventive maintenance inspections on the air side of all the air-to-water cooling coils Inthe SW system to ensure the coil fins are clean and clear.

Y BIENN1AL COMMIMENT CHANGES AND 5012* REPORT Attachment Page 7 of 7 Frequency of preventative maintenance inspections will be based on trending history of cooler cleanliness.* and TColumbia will perform regularly scheduled preventive maintenance inspections on the air side of all the air-to-water service water cooling coils to ensure the coil fins are clean and clear. Frequency of preventative maintenance inspections will be based on trending history of cooler cleanliness."

From O249463 (RCC-1 1268340)

This change addresses those instances where the need for the verbal procedure change is Identified close to the end of a shift. In this case, documenting the verbal procedure change using a procedure change notice (PCN) cannot be completed prior to the end of the shift. The original intent of the commitment was to ensure that documentation occurred relatively soon after the verbal change was made to minimize errors. The decision to use the standard of "prior to the end of shift" was arbitrary. One of the requirements to use a verbal procedure change is that completion of the work is essential to plant operation. If a verbal procedure change is required close to the end of a shift, it Is very likely that the work will be continued on the next shift. Another requirement to use a verbal procedure change is that it be documented in the Control Room Log upon approval. This ensures that there is a record of what was changed. 01-9, Operations Standards and Expectations, step 17.2.5 requires that the oncoming main control room staff review the Control Room Log entries. The use of wlthin 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s*

instead of "prior to the end of shift" continues to provide adequate controls to minimize errors or misunderstandings which may arise from verbal procedure changes.

The original commitment stated:

From G02-89-053: The use of a verbal deviation for a period greater than one or two shifts was not envisioned when PPM 1.2.3 was written, and as a result, this Incident highlighted a procedural weakness. To ensure that the possibility of errors or misunderstandings are minimized, PPM 1.2.3 will be revised to include the requirement for documenting verbal deviations prior to the end of shift of the personnel performing the work activity.

Commitment description has been revised to say:

The procedure change process is now controlled by SWP-PRO-02. To ensure that the possibility of errors or misunderstandings are minimized, SWP-PRO-02 will be revised to include the requirement for documenting verbal deviations within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of approval of the verbal deviation by the Shift Manager.