ML20147F206

Forwards Documentation Presented by Util at 880209 Meeting W/Nrc Re 860425 Application for Amend to License DPR-28, Consisting of Proposed Change 133,on Spent Fuel Pool Expansion.Expeditious Review Requested

ML20147F206
Person / Time
Site:	Vermont Yankee
Issue date:	03/02/1988
From:	Murphy W VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
FVY-88-17, NUDOCS 8803070270
Download: ML20147F206 (20)

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l- VERMONT YANKEE NUCLEAR POWER CORPORATION

- FVY 88-17 RD 5, Box 169, Ferry Road, Brattleboro, VT oS3c1 ,,,L y ,o, y ENGINEERING OFFICE 1671 WORCESTER ROAD

• FRAMINGHAM, M ASSACHUSETTS 01701 TELEPHOpiE 617-872-4100 March 2, 1988 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk

References:

a) License No. DPR-28 (Docket No. 50-271) b) Letter, VYNPC to USNRC, FVY 86-34, "Proposed Technical Specification Change for New and Spent Fuel Storage",

dated 4/25/86 c) Letter, VYNPC to USNRC, FVY 87-87, "Vermont Yankee Proposed Change No.133 - Spent Fal Pool Expansion",

dated 9/1/87 d) Letter, USNRC to VYNPC, NVY 88-05, "Forthcoming Meeting with Vermont Yankee Nuclear Power Station", dated 1/21/88

Subject:

Vermont Yankee Proposed Change No. 133 - Spent Fuel Pool Expansion Pursuant to the NRC staff's letter of January 21, 1988 [ Reference d)], a meeting was held on February 9, 1986 during which Vermont Yankee responded to the remaining NRC staff technical information requirements associated with the subject spent fuel pool expansion amendment request [ Reference b)]. In accor-dance with the NRC staff's request, Attachments 1 and 2 to this letter pro-vide the documentation and information presented by Vermont Yankee and requested by the NRC staff at the February 9, 1988 meeting.

In order to expedite the NRC staff's review of the subject license amend-ment request and definitively resolve all remaining staff review issues, and in an attempt to resolve the issues pending before the Atomic Safety and Licensing Board, Vermont Yankee has committed to design, install, test and make opera-tional, a redundant seismically designed Spent Fuel Pool Cooling System prior to the time Vermont Yankee exceeds the existing 2,000 spent fuel assembly storage limit in the Vermont Yankee spent fuel pool. This system will be operational no later than the end of Cycle 16 (Projected to be 1993). This commitment is reflected in Attachment 1 and 2. Attachment 1 specifies the design and perfor-mance criteria for the enhanced system. The design, installation and testing of the enhanced system will be in accordance with 10 CFR 50.59 and the NRC's normal inspection program.

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M: U.S', Nuclear Regulatory Comission March 2, 1988 Page 2 Attachment 2 to this letter documents the,information presented by Vermont Yankee at the February 9,1988 meeting which directly addressed each of the NRC staff's remaining open technical issues as described in the January 21, 1988 status report [ Reference d)] of.the staff's review of Vermont Yankee's spent fuel pool expansion amendment request. As documented in Attachment 2, each of the remaining open technical issues is addressed for~ both the existing Vermont Yankee Spent Fuel Pool System and the proposed enhanced Spent Fuel Pool Cooling .

System.

On the basis of the information submitted in support of the subject amend-

' ment request since April 1986 and the commitments and information presented herein, Vermont Yankee requests that you expeditiously complete your review of the spent fuel pool expansion application allowing Vermont Yankee to rerack the spent fuel pool to 2,870 assemblies.

Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION h[Y Warren P. Murphy Vice President and Manager.of Operations

/dm i' cc: Office of Nuclear Reactor Regulation Mr. Steven A. Varga, Director Division of Reactor Projects I/II U.S.N.R.C.

Region 1 Mr. William T. Russell, Regional Administrator U.S.N.R.C.

Resident Inspector Vermont Yankee Nuclear Power Corporation ASLB Service List

- ATTACHMENT 1 Design and Performance Criteria for the Enhanced Vermont Yankee Spent Fuel Pool Cooling System Vermcst Yankee has committed to providing spent fuel pool cooling capacity via an Enhanced Fuel Pool Cooling System.

The Enhanced Fuel Pool Cooling System will be designed and installed in I accordance with Vermont Yankee's Operational Quality Assurance Program.

The functional and performance criteria for the system are as follows:

1. Cooling from spent fuel pool to ultimate heat sink will be available from Seismic Category 1 equipment, independent of the RHR System.

2. System will be Safety Class 3 and single active failure proof.

o System will be designed to ensure that heat removai capacity assuming the naximum normal heat load and a single active failure, will be suf-ficient to preclude any restriction on plant operation. The system will also address the following in accordance with FSAR criteria and Technical Specifications:

Detection and isolation of leaks Flooding Missiles

- Inservice testing capability Fire Protection o The spent fuel pool cooling pumps and all other essential electrical equipment will be environmentally qualified per the Vermont Yankee EQ Program, seismically qualified per the FSAR criteria and powered from safety class electrical emergency power sources, o System layout and installation will take into account ALARA con-siderations in accordance with the Vermont Yankee ALARA Program, o Fuel pool temperature monitoring will be provided for all plant operating modes.

The structural and mechanical design of the piping will be in accordance with ASME/ ANSI B31.1-1977, which is consistent with the Seismic Reanalysis Program and Recirculation System replacement designs. Seismic input will be the appropriate Reactor Building spectra, based on USNRC Regulatory Guide 1.60 and ASME Code Case N-411 criteria, as was approved by the NRC for the Seismic Reanalysis Program and the Recirculation System replacecient.

l l Material selection and processing will use NUREG 0313, Rev. 2, as guidance.

However, the maximum operating temperature for the system is only 150*F, which is below the temperature at which IGSCC is a concern. It is Vermont Yankee's policy to use IGSCC resistant material unless significant cost or schedule penalties would result.

Attaachment 1 Page 2 The installation will~ be performed under ASME Section XI. repair program using the Engineering Design Change Request (EDCR) process,.as was_used in the.

Seismic Reanalysis ~ Program and Recirculation System replacement project.

Post-installation pressure' testing will be in accordance with ASME/ ANSI B31.1 for isolable portions and ASME Section XI for portions of new piping un-sisolable from er.isting piping or components.

Start-up testing will be performed to ensure that the system meets spe-cified performance criteria.

Vermont' Yankee will continue to comply with the Administrative Guidelines as described in our September 1,1987. submittal [ Reference c)] until such time that the Enhanced Fuel Pool Cooling System is operable.

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ATTACHMENT-2

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. Vermont Yankee Response to NRC Staff Technical Issues Identified'in the January 21, 1988 Status Report x

, INTRODUCTION This attachment documents and expands upon the information presented by Vermont Yankee at the February 9, 1988 public meeting and directly addresses each of the NRC staff's remaining open technical issues as described in the January 21, 1988 status report of the staff's review of Vermont Yankee's spent.

fuel pool expansion amendment request. Additionally, Vermont Yankee wishes'to clarify two points with respect to the information contained in your letter of January 21, 1988 (NVY 88-05). Specifically, two items discussed in the attach-ment under Section A, Background, should be corrected.

First, to date, Vermont Yankee has installed racks of the current design sufficient to store 1,690 fuel. bundles, not 1,680 'as stated. This discrepancy.

is due to a typographical error contained in Vermont Yankee's original amendment request submittal (FVY 86-34, dated Apr11 25,1986) .

Second, the proposed enhanced high density storage racks would increase the storage capacity of the spent fuel pool to 2,870 fuel bundles and are projected n

to provide storage capacity while maintaining full core reserve discharge capa-bility until 2001, not 1999 as stated. The date of 2001 was documented in a letter dated November 24,.1986 (FVY 86-107) in response to Question No. 10 and is a realistic projection based on Vermont Yankee's extended fuel cycle manage-

. ment plan (i.e., eighteen-month fuel cycles).

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RESPONSE TO NRC QUESTIONS VY SPENT FUEL POOL EXPANSION OPEN ISSUES:

so 1 HEAT REMOVAL CAPABILITY ITEMS:

1A. 1971 ANS DRAFT STANDARD USED 1B. 9.1 MBTU/HR USED AS HEAT LOAD 1C. FSAR LISTS 2.23 MBTU/HR 1D. SINGLE FAILURE

RESPONSE

1A. NOT USED FOR SFP DECAY HEAT; USED FOR REACTOR VESSEL DECAY HEAT 1B. CONSERVATIVE VALUE FOR A SPECIFIC SCENARIO 1C. DESIGN VALUE NOT ACTUAL PERFORMANCE CAPABILITIES 1D. VY IS SINGLE ACTIVE FAILURE PROOF

(VY HEAT EXCHANGERS CAN BE CROSS CONNECTED) l l CONCLUSION

ITEMS A,B,C AND D FULLY ADDRESSED BY EXISTING SFP SYSTEM HEAT REMOVAL CAPABILITY. PROPOSED UPGRADE ALSO FULLY ADDRESSES ITEMS B,C AND D.

. DISCUSSION 1A. Vermont Yankee has performed heat load calculations for fuel stored in the spent fuel pool in accordance with the guidance of Standard Review Plan, Section-9.1.3. Reference to the 1971 ANS draft standard and 9.1 MBtu/hr are specific _only to the scenario described in Attachment 1 of the September 1, 1987 submittal. This scenario describes how torus cooling and spent fuel . pool cooling can be accomplished by the RHR System only.

The initial conditions established an operating reactor and recently discharged spent fuel (ten days). These conditions are essentially impossible to achieve since an actual refueling at Vermont Yankee could not be done in less than ten days, thus, these conditions establish a very con-servative analysis. The 1971 ANS draft standard was used in determining the heat load from the reactor vessel just after scram, not the SFP heat load. The heat load in the SFP was determined by using the SRP methodo-logy.

18. The 9.1 MBtu/hr is a conservative value picked because it is the heat transfer capability of one pump and two heat exchangers. It is not the maximum normal heat generation rate. Using SRP guidance, the analysis per-formed by Vermont Yankee shows the heat' load at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> (six days) is approximately 10.3 MBtu/hr, which is in agreement with the analysis done by the HRC staff.

IC. Table 10.5.1 of the Vermont Yankee FSAR lists the original design heat transfer rate of the heat exchanger. The original design heat transfer rate was used to purchase the equipment but does not limit the actual heat transfer rate of 2.23 MBtu/hr in the FSAR does not limit the heat exchangers performance to just 2.23 MBtu/hr. Based on the conservation of energy, as the inlet parameters change so does the heat transfer _ rate. In the. Vermont Yankee SFP cooling analysis, the original heat exchanger data

' sheet inlet parameters were analyzed as a bench mark. The analysis yielded the same outlet parameters as listed on the heat exchanger data sheet, showing that the original design is just another point within the heat exchanger performance capabilities.

10. Single failure for Vermont Yankee is defined as "single active failure."

The definition is contained in SRP 9.1.3 and the response to Interrogatory 26; "NRC Staff Response to NECNP's First Set of Interrogatories and Document Request to the NRC Staff", dated August 5, 1987. Based on this, Vermont Yankee is single active failure proof with one pump in standby and one pump operating with two heat exchangers operating in parallel. The Vermont Yankee SPFCS piping is arranged in such a way to provide easy pump discharge cross connnection allowing the two heat exchangers to be operated in parallel (refer to FSAR Figure 10.5-1). Considering only one pump to be in operation, and the cross connection valve open, the piping to each heat exhanger is routed in such a manner to provide a relatively equal flow resistance. This provides a fairly equal division of flow from the

running pump to each heat exchanger. Figure No.1 sumarizes the Vermont Yankee pool temperature capability for all combinations of existing SFPCS equipment. As can be seen from the one pump and two heat exchanger curve (middle curve), the pool temperature can be held to less than 150*F af ter 11 days of fuel decay. This figure'is based on SRP heat load analysis methods that yield results comparable to heat. loads calculated by NRC staff and attached to the "NRC Staff Response To NECNP's First Set Of Interrogatories And Document Request To The NRC Staff", dated August 5, 1987. Figure No. 2 makes a comparison of NRC staff and Vermont Yankee calculated heat loads at several days of decay. The heat loads at these points compare very well with each other, so it can be concluded that Vermont Yankee calculated heat loads are not in disagreement with the NRC calculated heat loads.

CONCLUSION The design of the existing Spent Fuel Pool Cooling System heat generation calculation methods, heat removal requirements, and single failure requirements comply with Standard Review Plan 9.1.3.

The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SFPCS that meets the applicable requirements of SRP 9.1.3 would, therefore, be qualified for a seismic event, be single active failure proof, and powered by a safety class electrical emergency power source. As such, reliance on the RHR System to provide seismic spent fuci pool cooling would not be necessary. The enhanced system would acceptably close Open Issue No.1 also, since it would-meet the applicable SRP 9.1.3 requirements for single active failure and heat removal capabilities.

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HEAT LOAD COMPAR_ISON NRC VY

__ DAYS HEAT LDAD DAYS HEAT LDAD 6.25 10.17 N8TU/HR 6 10.35 MBTU/HR 6.92 9.91 7 9.93 7.92 9.58 8 9.59 8.92 9.31 9 9.32-9.92 9.09 10 9.1 FIGURE #2

RESPONSE TO NRC QUESTIONS VY SPENT FUEL POOL EXPANSION OPEN ISSUES:

No.e SPENT FUEL POOL TEMPERATURE LIMIT ITEM:

NO FUEL POOL TEMPERATURE MONITOR WITH SFPCS NOT OPERATING

RESPONSE

POOL TEMPERATURE MONITORING IS PROVIDED FOR ALL PLANT CONDITIONS AS DOCUMENTED IN VY LETTER 9-1-87 CONCLUSION:

EXISTING VY POOL TEMPERATURE MONITORING SYSTEM ACCEPTABLE AND ADEQUATE.

PROPOSED SYSTEM WILL ADDITIONALLY PROVIDE SFP TEMPERATURE MONITORING FOR ALL PLANT CONDITIONS.

DISCUSSION As detailed in Vermont Yankee's letter of September 1,1987, spent fuel pool temperature.is continuously monitored when the system is in operation. A Control Room alarm will sound when temperature exceeds an administrative limit of 125*F.

In recognition of the fact that the temperature monitors would not provide accurate temperature indication of the Fuel Pool if the Fuel Pool Cooling System was inoperable, Vermont Yankee has committed to directly monitor fuel pool tem--

perature every four hours if one or both fuel pool cooling trains were ino-perable (see Vermont Yankee letter, dated September 1,1987, Attachment 2) until the enhanced Fuel Pool Cooling System is operable. Even at the maximum heat-up rate of 3*F/hr ample time would exist for operator action to secure the demi-neralizer before the inlet temperature exceeds the NRC imposed limit of 140*F.

In the refueling mode, when the Spent Fuel Pool Cooling System components could be out of service for maintenance, the spent fuel pool and refueling cavity temperature is monitored by the Residual Heat Removal System temperature indicators.

CONCLUSION Based on the above, Vermont Yankee cencludes that appropriate. temperature monitoring exists for all operating modes, satisfying the requirercents of Standard Review Plan 9.1.3.

The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SPFCS that meets the applicable requirements of SRP 9.1.3 would, therefore, pro-vide. temperature monitoring under all plant conditions thus the enhanced system would acceptably close Open Issue No. 2 also, since pool temperature monitoring would be provided under all plant conditions.

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RESPONSE TO NRC QUESTIONS VY SPENT FUEL POOL EXPANSION OPEN ISSUES:

mo__ s.

_ P.00L_C.00 LING FOLLOWING SEISMIC EVENTS ITEM:

3A. FIRE WATEF. SYSTEM CONNECTION TO SERVICE WATER NOT GEISMIC CAT.I

38. ALT. C O OL I /JG CELL PIPING SEISMIC CAT.I NOT DEMONSTRATED 3C. RHRSW TO RHR CROSS-CONNECT SEISMIC CAT.I NOT DEMONSTRATED

RESPONSE

3A. SWS PIPING IS SEISMIC CAT.I PER FSAR App.A THROUGH A NORMALLY CLOSED, MANUAL, FIRE WATER. SYSTEM ISO. VALVE 3B. PIPING IS SEISMIC CAT.I PER FSAR App.A 3C. PIPING IS SEISMIC CAT.I PER FSAR App.A CONCLUSION:

EXISTING VY SW PIPING AND CROSS-CONNECTION ARE SEISMIC CATEGORY I. PROPOSED FPC SYSTEM UPGRADE WILL BE SEISMIC CATEGORY I.

DISCUSSION 3A. The Fire Water System is not a seismically qualified system and is isolated from the Service Water System by a normally closed manual valve. The fire water piping.and valve making the connection to the Service Water. System is seismically qualified Category I in accordance with the Vermont Yankee FSAR, Appendix A to prevent degradation of the Service Water System in a seismic event.

38. The cooling tower deep. basin alternate cooling cell piping connecting to the Service Water System is seismically qualified as noted in the Vermont Yankee FUR, Appendix A.

3C. The Vermont Yankee service water pumps are powered from an on-site emergency electrical source and meet Seismic Category I requirements. The Vermont Yankee service water path to the fuel pool meets Seismic Category I requirements from the service water pump suction in the river through the RHR service water connection into the RHR System and through the FPC con-nection into the spent fuel pool.

CONCLUSION Based on the above, Vermont Yankee concludes that the existing service water piping and cross connections are Seismic Category I as described in Appendix A to the Vermont Yankee FSAR, Standard Review Plan, Sectiorr9.1.3 is satisfied.

The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SFPCS that_ meets the applicable requirements of SRP 9.1.3 would, therefore, be qualified for a seismic event.

RESPONSE TO NRC QUESTIONS VY SPENT FUEL POOL EXPANSION OPEN ISSUES:

N o_. 4 RADIOLOGICAL CONSEQUENCES OF BOILING ITEM:

4A. PROVIDE ASSUMPTIONS FOR OFF-SITE DOSE 4B. PROVIDE ASSUMPTIONS FOR ON-SITE DOSE I 4C. PROVIDE ON-SITE DOSE i

RESPONSE

l 4A. ASSUMPTIONS CAN BE PROVIDED 4B. ASSUMPTIONS CAN BE PROVIDED 4C. ON-SITE DOSE CAN BE PROVIDED CONCLUSION:

10CFR2O REQUIREMENTS MET DURING POOL BOILING FOR OFF-SITE AND ON-SITE DOSES.

PROPOSED NEW SYSTEM WILL PRECLUDE POOL BOILING I

FOR ALL PLANT CONDITIONS.

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DISCUSSION In order to assess the on-site and off-site radiological impact of a postu-lated boiling spent fuel . pool, a scenario was developed to maximize the release of fission products through boiling. The scenario assumes that the plant is shut down for' refueling with a normal 136 bundle fuel load. The 136 spent fuel bundles discharged completely fill the spent fuel pool to its capacity of 2,870.

Just before the start of the outage, maximum Technical Specification activity levels are assumed to be present in the Reactor Coolant System while normal activity levels (as determined from Plant Chemistry data) are assumed to be pre-sent in the spent fuel pool water. The two volumes and their activities are then mixed at the start of the outage when the refueling gates are removed.

The length of the outage is assumed to be 21 days. This is consistent with the shortest refueling outage in Vermont Yankee's history. At 21 days, the refueling gates are reinstalled, the fuel pool volume is segregated from the Reactor Coolant System volume, and the fuel pool _ is on spent fuel pool cooling.

All spent fuel pool cooling is then assumed to be lost and the pool is allowed to heat up. The heat load in the pool is then determined based upon Standard Review Plan methodology and uncertainties assuming the spent fuel operated at 1,665 MWt with a 100% capacity factor. The rate of heat up and subsequent boiling were then determined assuming a spent fuel pool bulk temperature of 150'F when all cooling is lost. The resulting boil-off rate is then r?lculated as a function of cime. Only the maximum boil-off rate is used in the radiologi-cal calculations.

The following assumptions were used to evaluate the on-site and off-site radiological conditions resulting from a spent fuel pool boiling incident.

ASSUMPTIONS FOR SOURCE TERM

1. Constant maximum boil-off rate if 16.6 gpm.

2. The volatile elements in the spent fuel pool (iodine, tritium) are released during boiling.

3. Tritium concentration in water equals 2 x 10-2 micro Ci/ml.

4. Initial concentration of I-131 dose equivalent (DE) at minimum detectable level (NL) (i.e., 4 x 10-7 micro Ci/ml) in spent fuel pool.

5. At t=0, reactor coolant at long-term Technical Specification limit of 1.1 micro Ci/ml 1-131 DE.

6. Partition Factor (PF) of 100 for iodine during boiling. Based on SRP 15.6.3, Radiological Consequences of Steam Generator Tube Failure (PWR),

establishes a partition factor of 100 betwc.. the steam generator water and steam phases.

7. No credit taken for fuel pool cleanup'.of iodine via tht spent fuel' pool demineralizers prior to initiation of pool boiling.

8. Recovery operations restore spent' fuel pool cooling within thirty days.

ASSUMPTIONS FOR OFF-SI1E CONDITIONS

1. Blow out panels are not present; ground level unfiltered release assumed.

2. Maximum off-site accident X/Q value for release from Reactor Building equals 6.83 x 10-4 sec/m3

3. For tritium, adult dose conversion factor and breathing rate which combine for most conservative dose rate (from Regulatory Guide 1.109).

Adult Inhalation Dose Factor = 1.58 x 10-7 mrem /pCi inhaled 0 8,000 m3 year.

4. For I-131, infant thyroid dose conversion factor is most conservative 1.62 R/hr (EPA-520/1-75-001) uCi/m3 (Infant Breathing Rate of 2.5 x 10-5 m3/sec)

RESULTS OFF-SITE IODINE Maximum I-131 DE off-site concentration = 1.5 x 10-10 micro Ci/cc.

MaxirAJm dose rate thyroid = 2.5 x 10-1 mrem /hr.

30 days dose at maximum rate = 1.8 x 102 mrem or 0.06% of Part 100 Limit (300 rem) and 6% of Part 20 Limit (3 rem /yr implied).

RESULTS OFF-SITE TRITIUM Maximum tritium concentration = 1.4 x 10-8 micro Ci/cc.

Maximum adult whole body dose rate = 2.1 x 10-3 mrem /hr.

30 day dose at maximum rate = 1.5 mrem or 0.01% of Part 100 Limit (25 rem) and 0.3% of Part 20 Limit (0.5 rem /yr).

ASSUMPTIONS FOR ON-SITE CONDITIONS

1. For on-site evaluation, assume the activity is released into a closed -

volume (blow-out panels remain intact) equivalent to the top floor of the Reactor Building.

2. Assume 90'F and 100% relative humidity in the SFP area as a result of boiling (for tritium concentration).

3. Assume the concentration of tritium in the water vapor in. the SFP area is the'same as the concentration in the SFP water (2 x 10-2 micro Ci/ml).

4. . Allow the iodine to be released into a closed volume equivalent to the Reactor Building top floor and compute the time it takes to reach 10,000 x-MPC. .This is based on an assumed protection factor of 10,000 for supplied air to a worker in the building.

RESULTS FOR ON-SITE For I-131 DE with very conservative bounding assumptions and credit for supplied air, recovery operations could take place for a 30-day period without exceeding the. limits of 10 CFR,.Part 20. The iodine concentration never reaches 10,000 MPC in the Reactor Building.

H-3 concentration will remain below the limits of Part 20. 7.2 x 10 ~ micro ci/cc calculated tritium concentration as compared to MPC for restricted area = 5 x 10-6 micro Ci/ml.

CONCLUSION This calculation has shown that boiling of the SFP at Vermont YarJee could occur without exceeding the off-site dose limits of 10 CFR, Part 20 (0.5 rem whole body and 3 rem. thyroid). The airborne tritium concentrations in the SFP area should not exceed the limits of Part 20. Using conservative assumptions and taking credit for supplied air, the airborne I-131 concentrations chould not exceed the limits of 10 CFR, Part 20.

Therefore, considerir g a complete loss of spent fuel pool cooling by both the SFPCS and the RHR System the radiological releases associated with postu-lated spent fuel pool boiling are below the limits established for an operating plant by 10 CFR, Part 20.

The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SFPCS that meets the applicable requirements of SRP 9.1.3 would, therefore, be qualified for a seismic event, be single active failure proof, and powered by a safety class electrical emergency power source. As such, spent fuel pool cooling would be available under all plant conditions and spent fuel pool boiling would not occur.

The enhanced system would acceptably close Open Issue No. 4 also, since it would provide spent fuel pool heat removal under all plant conditions and prevent pool boiling.

RESPONSE TO NRC QUESTIONS VY SPENT FUEL POOL EXPANSION OPEN ISSUES:

No.5 SUPPLEMENTAL COOLING ITEM:

5A. PROVIDE ADDITIONAL INFORMATION ON PARALLEL HEAT EXCHANGER OPERATION 5B. SWITCHING RHR BETWEEN TORUS COOLING AND SFP COOLING IS UNACCEPTABLE

. RESPONSE:

5A. ADDITIONAL INFORMATION CAN BE PROVIDED SB. SWITCHING RHR BACK AND FORTH FROM TORUS COOLING TO SFP COOLING IS WITHIN THE CAPABILITIES OF THE PLANT i

CONCLUSION

EXISTING VY SFP SUPPLEMENTAL COOLING SYSTEM IS l ADEQUATE AND ACCEPTABLE. PROPOSED NEW 1

SYSTEM ALSO FULLY ADDRESSES THESE ISSUES.

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DISCUSSION SA'. Additi_onal information concerning SFPCS operation using one pump and two.

heat exchangers was previously supplied within the Vermont Yankee response to Open Issue No. 1.

5B. For a seismic event during power operation, Vermont Yankee's method, pre-sented within the September 1, 1987 submittal, of using.one train of RHR to cool both the spent fuel in the . fuel pool and the residual heat in the reactor is not considered appropriate by NRC staff since too many operator actions and RHR pump starts are involved.

This scenario describes how torus cooling and spent fuel pool cooling can be accomplished by the RHR System only. The initial conditions assumed the reactor was operating and recently discharged (ten days) spent fuel. These conditions are essentially impossible to achieve since an actual refueling at Vermont Yankee could not be done in less than ten days; thus, these con-ditions establish a very conservative analysis.

The RHR cycle involves six hours of torus cooling and one hour of Augmented Spent Fuel Pool Cooling (AFPC), with 20 minutes allowed for valve realign-ment between modes. This conservative scenario is within system capability and is well within the RHR pump's starting limitations listed in plant Operating Procedure 0P 2124, Rev. 19, "Limit RHR pump starts to 3 in 5 minu-tes followed by a 20 minute run or a 45 minute shutdown for cooling."

If realistic spent fuel pool heat loads were used (i.e., less conservative than those required by SRP 9.1.3), the spent fuel pool heat up would be slower, which would allow a longer duration on torus cooling, thus limiting the cycle frequency and reduce operator actions.

CONCLUSION It is Vermont Yankee's conclusion that using RHR to ensure cooling of the spent fuel pool considering a seismic _ event is within the capabilities of the plant, even if conservative scenarios and heat loads are used.

The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SFPCS that meets the applicable requirements of SRP 9.1.3 would, therefore, be qualified for a seismic event, be single active failure proof, and powered by a safety class electrical emergency power source. As such, reliance on the RHR System to provide seismic spent fuel pool cooling would not be necessary.

The enhanced system would acceptably close Open Issue No. 5 also, since it would meet the applicable SRP 9.1.3 requirements, operate under all plant conditions and, therefore, eliminate switching of one RHR train between the fuel pool and the reactor for heat removal.