Text

NRC Staff Response to New England Coalition on Nuclear Pollution First Set of Interrogatories & Document Request to NRC Staff.* Certificate of Svc & Affadivit of Jn Ridgely Encl.Related Correspondence
	ML20236P098
Person / Time
Site:	Vermont Yankee
Issue date:	08/05/1987
From:	Hodgdon A, Robert Weisman; NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:	; NEW ENGLAND COALITION ON NUCLEAR POLLUTION
References
	CON-#387-4201 OLA, NUDOCS 8708120218
	Download: ML20236P098 (186)
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UNITED STATES' OF AMERICA NUCLEAR REGULATORY COMMISSION U+ . .

- 50Cni f? ;n , , . y Jd th

' BEFORE .THE- ATOMIC SAFETY AND LICENSING . BOARD l'n t'he Matter off )

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VERMONT ' YANKEE NUCLEAR ) ' Docket No. 50-271-OLA POWER CORPORATION .) (Spent Fuel Pool Amenement)

)

(Vermont Yankee Nuclear Power - )

Station) -

NRC STAFF RESPONSE TO NECNP'S FIRST SET OF INTERROGATORIES AND DOCUMENT REQUEST TO THE. NRC STAFF On June 16, 1987,' the New England Coalition on Nuclear Pollution

. (NECNP)~ submitted its First Set of Interrogatories and Document Request to the' Nuclear Regulatory Commission (NRC) Staff.

The Staff notes that it need answer. interrogatories only if-the presiding officer, in this case

'the Licensing- Board, orders the Staff to' do so based on findings that answers to the interrogatories are necessary to a proper decision in the

. proceeding: and that answers to the . interrogatories are not reasonably obtainable from any other source. 10 C.F.R. 9 2.720(h)(2)(ll) (1987). i The Staff notes further that the Licensing Board has issued no such order in this case, nor made the required findings. The Staff is not a l party that answers interrogatories pursuant to 10 C.F.R. 9 2.740b. The l 1

Commission's regulation concerning production of NRC records and I

" documents,10 C.F.R. 9 2.744, requires that a request to the Executive j

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Director of Operations for the production of an NRC record or document !

i not available pursuant to 9 2.790. by a party to an initial licensing proceeding state, among other things , why the requested record or 1 8700120219 DR 870005 6 ADOCK 0500 1

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document is . relevant to the " proceeding. NECNP has made; no such statement., Notwithstanding the regulations in 10 C.F.R. ll 2.740, 2.744 and 2.720 (b)(2 ) (li)', the' Staff hereby voluntarily answers portions of NECNP's First Set of Interrogatories and Document Request.

On' July 21, '1987, the Atomic Safety and Licensing Appeal Board

. ( Appeal Board) . issued a decision II ( ALAB-869) in which 'It affirmed the

. Licensing Board's admission of a safety contention concerning the Residual Heat Removal (RHR) system to this spent fuel pool license amendment proceeding .but reversed the Licensing Board's admission of two

-environmental contentions. Because NE,CNP's questions in its first set of

. interrogatories related to .the environmental contentions are no longer

. relevant 'to this. proceeding, the Staff has not responded to them. See 10 C.F.R 9 2.740(b)(1). Thus, In answering NECN P's first set of interrogatories, the Staff answers .only those interrogatories relevant to

' the safety contention concerning the RHR. Id.

4 INTERROGATORY 24~ -

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Provide any -Internal ~ agency- guidance used by the staff in its review of spent fuel' storage reracking. requests in general and/or. the VY proposal in particular, q j

RESPONSE

Copies of the requested documents are provided in Enclosure 1. )

, INTERROGATORY 25:

Identify all rules, criteria and guidance, whether or not formally promulgated, which the staff will apply to its review of the safety issues involved in.the instant application.

-1/ Vermont Yankee Nuclear Power Cor . (Vermont Yankee Nuclear i Power Station), ALAB-869, 25 N. . . (1987).

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RESPONSE

The 'following is a listing of the criteria which the Plant Systems Dranch applies to review of spent fuel pool expansions:

(1) 10 C.F.R. Part 50, Appendix A (General Design Criteria 2, 4, 44, 45, 46, 61, 62 and 63)

(2) Regulatory Guides 1.13 and 1.29 (3) NUREG-0800, Standard Review Plan (Sections 9.1.2, 9.1.3, 9.1.4, 9.1.5, and Branch Technical Position ASB 9-2)

(4) N UREG-0612, " Control of Heavy Loads at Nuclear Power Plants" (5) Memorandum for ASB Members from O. D. Parr, dated November 17, 1983 (copy enclosed)

INTERROGATORY 26:

Does the single failure criterion apply to the VY spent fuel pool cooling systems? Provide the reasons for your answer and any documents in support thereof.

RESPONSE

Yes . Spent fuel pool cooling system reviews are performed utilizing the Standard Review Flan (NUREG-0800). Section 9.1.3 states , in part, that the review will include consideration of provisions to preclude loss of function resulting from single active failures. Further, the acceptance criteria for the spent fuel pool cooit.ng system (Paragraph II.1.d) state, in part, that the design is acceptable if it is in accordance with General Design Criterion 44, including suitable redundancy of components so that safety functions can be performed assuming a single active failure of a component coincident with the loss of offsite power. Copies of the requested documents are pro' 8ded in Enclosure 1.

, . -n-INTERROGATORY 27:

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Is it.the staff's position that VY must demonstrate that the spent i fuel pool water .will be maintained at all times at 140 F or below? 150 F or below? Explain the reasons for your answer and identity and provide the rule criteria and/or guidance, whether or not formally promulgated, which support your. response.

RESPONSE

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It is the staff's position that the licensee must maintain the spent fuel pool water temperature equal to or less than 140 F under normal storage conditions in conformance with Standard Review Plan Section 9.1.3, Paragraph lli.1.d, or provide adequate justification for a different water temperature limit.

INTERROGATORY 28 On the assumption of an 18-month refueling schedule and a full core discharge, for what period of time will the RHR system be required to oc keep _ the pool water below 140 F7 Provide all documents which support your answer.

RESPONSE

For a full spent fuel poo; (2870 fuel bundles) during normal refueling, the RHR system would be required for approximately 149 days from shutdown to keep the pool water temperature btlow 140oF as determined by calculations. Since a full core offload is indicative of a potential problem, an extended o,utage is anticipated and therefore the RHR is available indefinitely (since there is nothing left in the reactor to l

cool) . Since a single failure is not assumed for the abnormal heat load case (full core offload) and the 140 F temperature limit does not apply per SRP Section 9.1.3, no calculations have been performed related to the time that the RHR would be required to keep the pool water temperature

, below 140oF. Copies of the requested documents are provided in Enclosures 1 and 2.

INTERROGATORY 29:

On the assumption of an 18-month refueling schedule and a discharge of 136 assemblies for what period of time will the RHR system be required to keep the pool water below 140oF?

RESPONSE

No analysis was performed for the heat load based on a dischargc of 136 fuel bundles per reload. Refer to the response to interrogatory No. 28.

INTERROGATORY 30:

Provide all documents, analyses, and/or evaluations in the staff's possession which cvaluate the probability and/or consequences of failures in the RHR system during the period of time that the system is required I

to augment the cooling of the spent fuel pool.

RESPONSE

The staff has not performed an analysis or evaluation related to the probability and/or consequences of failures in the RHR system during the period of time that the system is required to augment the cooling of the spent fuel pool. Thus, no documents are available.

INTERROGATORY 31:

Provide all documents analyses and/or evaluations in the staff's possession which evaluate the probability and/or consequences of accidents at VY in particular and boiling water reactors in general when the plant is not operating.

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RESPONSE

The. 0ffice of Analysis and Evaluation' of Operating Data issued a i

report AEOD/C503 in December 1985 which discusses the effects of loss of

' decay heat removal systems for PWRs. This report identifies a number of loss-of-all-decay-heat-removal incidents at PWRs and has theoretical i application 'to ' BWRs as well. A copy of this report is provided in Enclosure 1.

INTERROGATORY 32:

Is it the staff's position that the spent fuel pool cooling system at VY meets the single failure criterion with the spent fuel storage facility filled with normal refueling and maintainir.g the pool water temperature at less than 140 F? (See Question 17 and response thereto, VY " Response to Request for Additional Information - Proposed Change No.133, Spent Fuel Pool Expansion." Enclosure 1, (November 24, 1986, FVY 86-107).

Explain 'your answer and provide any documents in support thereof.

RESPONSE

No. As determined by calculation, the spent fuel pool cooling j system does- not have sufficient capacity to cool the normal spent fuel heat load and maintain pool water temperature below 140oF in the event of ]

l a single failure. Copies of the requested documents are provided in l 1

Enclosures 1 and 2. l INTERROGATORY 33:

Is it the staff's position that the spent fuel pool cooling system at VY is required to meet the single failure criterion with the spent fuel storage facility fTIfed with normal refueling and maintaining the pool water i temperature at less than 140oF?

RESPONSE

lt is the staff's position that the systems utilized to cool the spent fuel . pool, which may include systems in addition to the spent fuel pool i

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~7-cooling system itself, must be single failure proof (reference General Design Cr,iterion 44, " Cooling Water").. The 140oF pool water temperature i'

limit is' not a requirement. Refer to response to interrogatory No. 27.

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, INTERROGATORY 34:

Answer the questions posed in No. 32 and 33 above, but on the ,

assumption of a full core discharge. l RES PONSE -

a. The VY sr>ent fuel pool cooling system is not single failure proof for the full core officad case. Copies of the requested documents are provided in Enclosures 1 and 2.

b. It is the staff's position that the spent fuel pool cooling system need l

not be single failure proof for the full core offload case. Further, the 1400F pool water temperature limit guideline is not applicable for a full core discharge. (Reference Standard Review Plan 9.1.3, Paragraph lil.1.d.) )

. INTERROGATORY 35:

Has the staff approved other applications for reracking where it is necessary to use the RHR system to maintain the pool cooling water temperature below 140oF? If so, identify all such cases.

' RESPONSE The review of the applications for reracking has taken place over a i

. number of years and by a number of reviewers. However, to our

. knowledge, the staff has not approved any other application for reracking where it will be necessary to use the RHR for spent fuel pool cooling as a

routine operation. However, other plants may need the RHR system to maintain p,ool temperature limits following a full core offload.

INTERROGATORY 36:

VY states in response to staff question 17 (FVY 86-107, November 24, 1986, cited supra): "

...VY does not consider the fuel pool cooling system to be single active failure proof (depending on RBCCW temp) until after approximately 42 days decay of a normal spent fuel discharge." ;

Does the staff consider the system single active failure proof? Explain '

your reasons and provide any documents in support thereof.

RESPONSE

No. Approximately 68 days after shutdown, one train of spent fuel

_ pool cooling will be adequate to maintain a bulk pool temperature of 150 F and thus will meet the single failure criterion. During the time period immediately following reactor shutdown until 68 days, calculations indicate .

that at times both trains of the spent fuel' pool cooling system will not be adequate to maintain the peo! water temperature below 150 F. One train of the RHR ' system will be required to maintain the pool water temperature below 1500F. Copies of the requested documents are provided in Enclosures 1 and 2.

INTERROGATORY 37:

Has the staff previously approved reracking applications where the resultant pool cooling system was not single active failure proof? If so, Identify all such cases.

RESPONSE

No.

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, INTERROGATORY 38:

During the period of time when the RHR system is required to cool the spent ~ fuel pool to below 140oF, are the reactor cooling systems, including but not limited to decay h< sat removal system single failure proof? Explain the reasons for your answer and provide any documents in support thereof.

RESPONSE

No. While General Design Criterien 44 requires plant cooling I systems to be single failure proof, it does not specify for which plant

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operational conditions the cooling systems are to be single failure proof, it has been the staff's practice not to require the plant to meet the single l l

failure criterion wher[ the reactor is 'in cold shutdown or refueling l l

conditions as shown by not requiring redundancy in the Standard Technical Specifications. This is based on the assumption that accidents i

at power potentially will result in significant health and safety j l

consequences and that accidents in either cold shutdown or refueling i I

I conditions potentially will not result in significant health and safety l

consequences. (No documents specifically address this issue).

INTERROGATORY 39:

Assume that VY is shutdown for refueling, the spent fuel has been discharged and it is within the period of time that one train of the RHR system is required to augment spent fuel pool cooling:

a. How will VY assure adequate cooling of both the core and the spent fuel pool in the event of failure of one RHR train?

b. Provide all evaluations, analyses or other documents related to the above question,

c. Has the staff (or contractors) reviewed the VY operating or emergency procedures which apply to this situation? If so, provide the procedures and/or any documents reflecting the sta ff's (or contractors') evaluation or analyses. l

-d. What is the probability of failure of one RHR train? Explain your answer and provide all documents in support thereof.

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, RESPONSE 1

a. The Staff does not require the RHP to be single-failure-proof when the reactor is in cold shutdown or refseling condition. Refer to the response to interrogatory No. 38. Based on independent calculations, a loop of the RHR system is needed to provide redundancy for the spent fuel pool cooling system until 68 days into the refueling outage. The requirements of General Design Criterion 44 are satisfied by having both loops of the spent fuel pool cooling system and one loop of RHR in the spent fuel pool cooling mode operational for the first 68 days. The licensee can assure adequate cooling of the core and the spent fuel pool by taking suction from the reactor vessel vilth the RHR system, returning the cooled water to the spent fuel pool, and allowing the water to return to the reactor vessel via the open gate between the two. While both trains of the RHR system are available to cool the core, only the "A" RHR train can normally operate in the spent fuel pool cooling mode.

Should this RHR train be unavailable during the first 68 days of the refueling outage, either 1) the spent fuel pool temperature could exceed the 150 F limit with only the spent fuel pool cooling system in operation and the "B" RHR loop providing cooling to the reactor, or

2) the "B" RHR loop can be cross-connected via an obscure piping arrangement to cool both the spent fuel pool and the core similarly to the "A" RHR loop flow path. The licensee recently identified this approach to the Staff but has not specified whether operating procedures exist to provide instructions to the operator to define

- when and how the "B" RHR loop will be used to cool the spent fuel pool and the core simultaneously.

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b. Copies' of the: requested documents are provided in ' Enclosure 3.

Other documents related to this request may be found in the NRC's public document room located at 1717 H Street, N.W., Washington, D.C.

-C. No.

d. The staff has not determined the probability of failure of one RHR i

train.

Respectfully submitted, Ann P. Hodgdon M 4P./h$&m Counsel for tfC Staff

$,Y W Robert M. Weisman Counsel for tac Staff Dated at Bethesda, Marylarmi this 5th day of August,1987

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UNITED STATES 0F' AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC. SAFETY AND LICENSING APPEAL BOARD in the Matter of )

)

VERMONT YANKEE NUCLEAR ) Docket No. 50-271-OLA POWER CORPORATION ) (Spent Fuel Pool Amendment)

) )

(Vermont Yankee Nuclear Power -)

Station) )

AFFIDAVIT OF JOHN N. RIDGELY 1, John N. Ridgely, state as follows:

1. I am employed as a Mechanical Engineer in the Plant Systems Branch of the Division of Engineering and Systems Technology in the Office of Nuclear . Reactor Regulation of. the Nuclear Regulatory Commission. I have provided the responses to interrogatories 24 through

39. propounded 'by the New England Coalition on Nuclear Pollution and am duly authorized to do so.

I hereby certify that the answers given are true and correct to the oest of my knowledge and belief.

ohn N. Ridgey Subscribed and. sgorn to before me this S& day of /.fua u s -f- , 1987 I

YW TAW?t )

Notary' Pttbil(~

My commission expires: '/// /90

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. l PROFESSIONAL QUALIFICATIONS JOHN N. RIDGELY PLANT SYSTEMS BRANCH DIVISION OF ENGINEERING AND SYSTEMS TECHNOLOGY I am employed as a Mechanical Engineer in the Plant Systems Branch, Division of Engineering and Systems Technology, Office of Nuclear Reactor Regulation, United States Nuclear Regulatory Commission, Washington, D.C. My duties consist of reviewing and evaluating the associated safety considerations on nuclear power and fuel handling systens and associated engineering fields on power reactors. I am responsible for providing technical input to various documents including Safety Evaluation Reports.

I attended the Virginia Polytechnic Institute in Blacksburg, Virginia and received a B.S. degree in Nuclear Science in 1972. In July of 1972, I joined the Philadelphia Electric Company's Mechanical Engineering Division as a Mechanical Engineer.

At the Philadelphia Electric Company, I worked with both fossil and nuclear power plants. I designed systems, prepared specifications, performed computer analysis, and managed contracts. During this time I developed and patented a process for removing tritium from High Temperature Gas Cooled Reactors. I wrote purchase order specifications for high density spent fuel storage racks for the Peach Bottom Atomic Power Station, reviewed the' bids, awarded the contract, and performed a field audit at the manufacturer's i facility. I also performed the preliminary work for the high density spent fuel storage racks at the Limerick Generating Station. ] l I

From August 1977 through November 1980, I was employed by the Potomac I Electric Power Company in Washington, D.C., as a Mechanical Engineer. During {

this time, I worked exclusively with fossil power plants. My duties in this !

position were similar to those at Philadelphia Electric Company. In addition, I have designed water treatment subsystems and assisted in other system designs including water intate and discharge treatment systems.

From December 1980 to the present, I have been employed by the United States Nuclear Regulatory Commission. I have been in the Auxiliary Systems i Branch of the Division of Systems Integration until November 24, 1985 when the )

Office of Nuclear Reactor Regulation was reorganized and I was assigned to the Plants Systems Branch, Division of BWR Licensing. On April 12, 1987 the Nuclear Regulatory Commission was reorganized and I was assigned to my present )

position. I have revised portions of the Standard Review Plan and have been l the Task Manager for the resolution of two Generic Issues. My duties include l safety reviews and evaluations of the design and operation of systems at '

nuclear power plant facilities. As required, I prepare safety evaluations and l make presentations to the Advisory Committee on Reactor Safeguards and to the Committee fer the Review of Generic Requirements. I am presently managing a contract for subcompartment environmental analysis and reviewing applications for proposed modifications to technical specifications and spent fuel pool expansions. Todate, I have reviewed the design of the spent fuel storage f acilities for 12 reactor sites, have per, formed the analytical review for six additional facilities, and am currently reviewing three additional facilities. ,

This represents 75 spent fuel storage facilities. I I

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, , ,, 7 UNITED STATES OF AMERICA .

, NUCLEAR REGULATORY COMMIS$lON ' 87'D 10" A9 :37

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BEFORE THE ATOMIC SAFETY ' AND LICENSING BOARD:

vu m w,, y n' my in the Matter'of )

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VERMONT, YANKEE NUCLEAR ) Docket No. 50-271-OLA ,

POWER CORPORATION ) (Spent Fuel ' Pool Amendment) , <

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(Vermont Yankee Nuclear Power ) . .t Station) !

CERTIFICATE OF SERVICE t

I hereby- certify that copiss of "NRC STAFF RESPONSE TO NECNP'S FIRST SET OF INTERROGATORIES AND DOCUMENT -REQUEST TO THE NRC' STAFF" in the above-captioned proceeding have been served on the ,

following by deposit in the United States mal!, first class, or as indicated j by an asterisk through deposit in the Nuclear Regulatory Commission's i internal mail system, this 5th day of August,1987:

Charles Bechhoefer, Esq. 'Mr. Glenn 'O. Bright !

Administrative Judge Administrative Judge - 1 Atomic Safety and Licensing' Board Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commi'ssion U.S.

Nuclear Regulatory Commission '

Washington, D.C. 20555* Washington, D.C. 20555*

Dr. Jsmes.H. Carpenter George Dana Bisbee Administrative Judge Senior Assistant Attorney General Atomic Safety and Licensing Beard Environmental Protection Bureau

.U.S. Nuclear Regulatory Commission 25 Capitol Street Washington, D.C. 20555* Concord; NH 03301-6397

' Atomic Safety and Licensing Board Ellyn R. Weiss, Esq. !

U.S. Nuclear Regulatory Commission Harmon s Weiss .

, Washington, D.C. 20555* 2001 S Street, N.W.

Washington, D.C. 20009 David J. Mullett,. Esq. Carol S. Sneider, Esq.

Special Assistant Attorney General Assistant Attorney General Vermont Depart. of Pub?ic Service Office of the Atdorney General l 120 State Street One Ashburton Place,19th Floor Montpeller, VT 05602 Boston, MA 02108 I. )

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Thomas G. Dignan, Jr., Esq. Jay Gutierrez Ropes and Gray Regional Counsel 225 Franklin Street USNRC, Region i Boston, MA 02110 631 Park Avenue King of Prussia, PA 19406*

Atomic Safety and Licensing Appeal Docketing and Service Section Board Panel Office of the Secretary U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555* Washington, D.C. 20555*

& bat A,v>Se Robert M. Weisman Counsel for NRC Staff

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j, I ENCLOSURE 1.

STAFF GUIDANCE DOCUMENTS l'

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Port 50, App. A Wear Regulatory Commission 10 CFR Ch.1 (1 137 Edition) !

Crsterion 44 ~Cboling water. A system to nrnon 40-Tesfine of containment heat transfer heat from structures, systems, and ffff& h clad m a fo snterfers u:th contanjg 88 In pre W nted and a cal 8Fsfem. The containment heat re-psi system shall be designed to permit componenta important to safety, to an ulti-ter reaction is limited to twg[

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,opnate periodic pressure and function- mate heat sink shall be provided. The Buftable redundancy in componertf an hsting to assure (1) the structural and system safety function shall be to transfer features, and suitable interconnections. le husht integrity of its components. (2) the combined heat load of these structures, detection, isc*ation, and Contamment ts , operability and performance of the systems, and components under normal op-b!11tles shall be provided to assure that .

.Se components of the system, and (3) ersting and accident Bultable redundancy conditions.

in components and onsite elertrie power system operation ta ? operability of the system as a whole, suming offstte poser is not avallath : a , ,, cf under conditions as close to the design features, and suitable interconnections, leak for offsite e;ectric power system operatik . gractical the performance of the full detection, and isolation capabilities shall be

. (assuming onsite power la not available) trI 1r'stional sequence that brings the system provided to assure that for onsite electric system safety f unetton can be accomplar ;coperation, including operation of appil- power system operation (assurning offsite assuming a stragle failure, ge portions of the protection system, the power is not available) and for offsite elec-Crite t n J6-.-Inspection of egg, wfer between normal and emergency tric power system operation (assuming onsite power is not sys11& bid the system core cooling system. The emergency e xser sources, and the operation of the as- safety function can be accomplished, assum-cocling system shall be deslaned to ,x:sted cooling water system, appropriate periodic inspection of itskrson df-Co Itainment atmosphere ing aCriterion single failure,

-inspection ctf coohng yafer

tanup, Systems to control flasjon prod. system. The cooling water system shall be ta.nt reactor components, pressure vessel, such asinjection water no in th'z4: s. hydrogen, oxygen, and other sub-rpray rings mees shich may be released into the reac- designed to permit appropriate periodic in-FJes, and piping to assure the integrity ar4 containment shall be provided as neces- spection of important ecmpanents such as capability of the system.

as to reduce, conalstent sith the function- heat exchangers and piping, to assure the Criterion 37-Testing of emergency .y of other associated systems, the concen- integrity and46-Testing capability of c/

thecooling system.wafer cooling system The emergency core coe won and quality of flasjon products re-Criterton system shall be des 4ned to permit appt ,ued to the environment following postu- system. The cooling water system shall be ate periodic pressure and functional tes, red accidents, and to control the concen- designed to permit appropriate periodic to assure Mon of hydrogen or oxygen and other pressure and functional testing to assure (1) territy of its'f)components, the structural (2)and the leaktight operat,U i ; ,;tcances in the containment atmosphere the structural and leaktight integrity of its and performance of the acttte componci$1, Moring postulated accidents to assure components,(2) the operability and the per-of the system, and (3) the operability of th v containment integrity is raaintained. formance of the active components of the system as a whole and, under condittore av Each system shall have suitable redundan- system, and (3) the operab1Hty of the close to design as practical, the performann lin components and features, and suitsble system as a whole and, under conditions as of the full operational sequence that br =

connections, leak detection, isolation, close to design as practical, the performance the system into operation, includin d containment capabilities to assure that of the full operational sequence that brings ation of app!! cable portions of the pro e for or, site elecitic poner system >peration the system into operation for accidents, reactor shut-tion systrm the tratufer betaeen norm down and for loss of coolant in-usumit:r offsite power is not available) and cluding operation of applicable portionr of i e

a.nd attonemerge:Ay of the associated power sources,toonnsandsathe e -Q f offs te electric power system operation the protection system and the transfer h-system. unmir:g onsite pov.er is not available)Its tween normal and emergency power sourerb.

afety function can be accomplished, assum, Crften .r. J6-Containment Accl remors, """ Y, Rese or Containment ,

A system te remove heat from the reactoh n I"N"#mcNn n of containmed q

i containmaini shall be provided. The systor. ;'mo@ hire cleanup systems. The contkin- Criterton 50-Containment desten bans.

safety funetton shall be to reduce rapidh SPD! atmDsphere CleanVp systems shall be The reactor containmc~nt structure, includ-consistent a1th the functioning of other i- ;es gned to permit appropriate periodic in- ing access openings, penetrations, and the sociated systems, the containtnent pressurr r;etion of important compcnents, such as containment heat removal system shall be and temperature following any Joss of.coci '. iter frames, ducts, and piping to nasure the designed so that the containment structure ant ace, dent and mMntatti them at accer* Itegnty and capability of the systems. and its internal compartment can accom.

ably low leveis, Criterten 4-Testing of containment at- modate, without exceeding the design leak-Bultable redundane) in cotnponents and age rate and with sufficient margin, the cal-im vrp sphere tre cleanup cleanupsystems. Theshall systerns containment be de culated pressure and temperature cond!-

featurenisolation, detecticn, acd sultatie interconnections'can and containtnent med to perroit appropriate periodic pres tions resulting from any loss of coolant acci-bilities shall be provided to assure that for sure and functionai testing to assure (1) the 6ent. This margin shall reflect consider-onsite electrl ,

r.ructural and lenktight intc grity cl ita com- ation of (t) the effects of potential energy suming offsite a!er o a i$a le ames N h M h MW M W mn (2) W @eram and wh determination of the peak conditions, such for offsite electric poser system o rat - ace o the active components et the sys-lassuming onsite power is not avalla eit l 8 as f ans, s, dams, pumps, as nem M mam pnuah W as m system safety function can be accom bhed ud tahes and (3) the operability ~ f the sys- guired ' i 60.44 energy frorn rnetal-aater m as a whole and. under condittoras asresult and other chemical reactions from degradation that but not total mat failure masumint e,.iteron5Jp-Inspectio single fsMare' n of containment close to design as practical, the performance of emergency core cooling functioning, (2) remooal systenL The containmem hes: of the f ull operational sequence that brings the limited experience and experirnental femoval system shall be designed to perm 1 me systems into oper*, tion, including oper- data available for defining accident phe-appropriate periodic inspecti>n of impM- suon Df applicable portions of the protec- nouen 2 and containment responses, and (3) tant components, such as the torus, surnpt Lon system, the transfer between normti the conservatism of the calculati9nal model ray nozzles, and piping to assure the in ud emergency power sources, and the oper-suon of assoetated systems. and input parameters.

grity and capabit!ty of the system.

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UNITED 5TATES y m g NUCLEAR REGULATORY COMMISSION

E WASHINGTON, D. C. 20555 o ?.

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NOV 171983 MEMORANDUM FOR: ASB Members FROM:

Dian'D. Parr, Chief, Auxiliary Systems Branch, Division of Systems Integration DECAY HEAT CALCU'.ATIONS - BRANCH TECHNICAL POSITION ASB 9-2,

SUBJECT:

REVISION 2 The purpose of this memorandum is to provide a unifomity of understanding withia ASB concerning the use of BTP 9-2.

The BTP ASB 9-2.was written to provide guidance in detemining the heat

^

genereted by spent fuel after an accident which results in the offloading

.of'the core.. By using the uncertainty factor specified in SRP 9.1.3, the i BTP ASB 9-2 has been used for determining the heat generated by spent fuel in tLa spent fuel pool for verifying the sizing of the spent fuel pool cooling system. i The equation in the BTP ASB 9-2 uses the time that the fuel was irradiated as one of the parameters,- The BTP ASB 9-2 suggests using 16,000 hours0 days 0 hours 0 weeks 0 months for the

.rradiation time. The 16,000 hour0 days 0 hours 0 weeks 0 months irradiation time was based on a PWR with annual refuelings of-one-third core and was to be the average irradiation r

time of the entire core for emergency core offloading heat generation.

The irradiation time to be used for spent fuel pool cooling system verifica-tion calculations should be the actual anticipated irradiation time for each batch, based on the equilibrium cycle. For example, a reactor with annual refuelings of one-fourth core would have an irradiation time of 4.0 years (35,040 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months ) while a reactor with an 18 month fuel cycle and a batch of one-third of a core would have an irradiation time of 4.5 years (39,*20 hours).

This notwithstanding, the utility's irradiation time correction factors (such as for capacit" factors) should also be applied.

I would like to remind everyone that the calculation of decay heat loads for the spent fuel pool cooling system does not have to be manually calculated.

We have a computer program operational on the IMB-PC and Apple 11 computers which will automatically perform this calculation for us.

for the service, contact J. Ridgely, X29566.

Olan d , ar- Lok Chief _Q 4<

Auxiliary Systems Branch Division of Systems Integration cc: R. Mattson .

L Rubenstein

NUREG-0800 (Formerly NUREG 75/087) p.

U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN

%(i

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OFFICE OF NUCLEAR REACTOR REGULATION

'3 i. SPENT FUEL POOL COOLING AND CLEANUP SYSTEM REVIEW RESPONSIBILITIES ma n, Audlia>j Systems Branch (ASB)

Suenor 3 - Chemice Engineering Braraa (CMEB)

AREAS OF REVIEW Ali nuclear reactor plants include a spent fuel pool for the wet storage of spent fuel assenblies. The methods used to provide cooling 'or the removal of decay heat from tne stored assemblies vary from plant to plant depending upon the individual design. The safety function to be performed by the system in all cases remains the sar..c; that is, the spent fuel assemblies must be cooled and must remain covered with water during all starage conoitions. Other f unctions performed by the system, nut related to safety, includt water cleanup for the spent fuel pool, refueling canal, refueling water storage tank and other equipment storage pools; means for filling and draining the refuiling canal and other storage pools; and surface skim-ming to provide clear water in the storage pool.

4 Ihe ASE revie of the spent ft.el pool cooling and cleanup system covers the system from ir.let to and exit from tre storage pool and pits, the seismic Category I water suurce ar.) piping used for fuel pool makeup, the cleanup system filter-cemineralizers and th9 regenerative process to the point of discharge to the radwaste system.

7 The capability of the spent fuel pool cooling and cleanup system to provide adequate cooling to the spent fuel during all operating conditions is reviewed on nne of two bases. The first basis requires the cooling portion of the sys-ten, to be designed to seismic Category I, Quality Group C requirements. The seco'd basis allows a non-seismic Category I, Quality Group C, spent fuel pool coolire system provided that the following systems are designed to seismic Category I requirements and are protected against tornadoes: the fuel pool make-up water system and its source; and, the fuel pool building and its ven-tilatien and filtration system. The makeup, ventilation and filtration sys-tems must also withstand a single active failure. In addition, the transient tempetoturc (Ta ) used in evaluating combined load on structures Rev. 1 - July 1981 USNRC STANDARD REVIEW PLAN Star. dero ec.s - plant au prey a ed for the guidance of the office of Nuclear Reactor Regulation staff responsible for the review of r

apphcatiou e c onceur ' anc operate r'uclear power plants These documents are made airailable to the pubhc as part of the Commies,oa pobry to en'orrn the nuclear indust v and the general pubhc of regulatory procedures and pohcies. Standard review plans arr not sutatitutes f or regulatory guides or the Corrmission's regulations and comphance with them is not required The

( enndaer. s eu plan sections s'e hered to the Standard Forrr at and Content of Safetv Ana'ysis Reports for Nuclear Power Plants

f. 1 sh staeo~s of the standa'd f ormat have a correspond.rg review pion.

n.,bhsher sta ndard review plank will be revised periodicaHv as appropriate to accurnmodate comments ano tu reflect new informa tion and emperience i Comments and s aggestions for improvement will be considered and should be sent to the U.S. Nuclear Regulatory Commission.

Office of Nucles: Reactor Regulation, Washingto9. O C. 20565

4 shall be the boiling temperature of water when the cooling system is not oesigned to seismic Category I requirements.

2. The ASB reviews the capability of the spent fuel pool cooling, makeup, .

and cleanup systems to provide adequate cooling to the spent fuel during all operating and accident conditions. The review includes the following considerations:

a. The quantity of fuel to be cooled, including the corresponding require-ments for continuous cooling during normal, abnormal, and accident conditions. . l 1

b. The ability of the system to maintain pool wster levels.

c. The ability to provide alternate cooling capability and the associated time required for operation.

d. Provisions to provide adequate makeup to the pool.

e. Provisions to preclude loss of function resulting from single active failures or failures of nonsafety-related components or systems.

f. The means provided for the detection and isolation of system components )

that could develop leaks or failures. ]

1

g. The instrumentation provided for initiating appropriate safety actions.

h. The ability of the system to maintain uniform pool wacer temperature l conditions.

3. ASB also performs the following reviews under the SRP sections indicated:

a. Review for flood protection is performed under SRP Section 3.4.1.

b. Review of the protection against internally generated missiles is performed under SRP Section 3.5.1.1.

c. Review of the structures, systems and components to be protected against externally generated missiles is performed under SRP Section 3.5.2.

d. Review of high- and moderate-energy pipe breaks is performed under SRP Section 3.6.1.

4 A secondary review is performed by CMEB and the results used by the ASB to complete the overall evaluation. CMEB provides an SER input to ASB on a routine basis that includes an evaluation of the capability and capacity of the spent fuel pool cleanup system to remove corrosion products, radio-active materials and impurities from the pool water. Also upon request the CMEB will provide ASB with an evaluation of the spent fuel pool and the spent fuel pool cooling system materials--fluid compatibility and potential for metal corrosion degradation. ASB will request such input if the materials used in the design differs significantly from previously approved designs.

9.1.3-2 Rev. 1 - July 1981

i

.Coordir.ated reviews.are performed by other branches and the results used

' Dy MEL in the overall evaluation of the SFPCCS. The coordinated reviews are as follows: 'The~ Structural Engineering Branch (SEB) determines the 71 - {- acceptability of the design analyses, procedures, and criteria used to establish the ability of seismic Category I structures housing the system and supporting systems to withstand the effects of natural phenomena such as the safe shutdown earthquake (SSE), the probable maximum flood (PMF),

?and tornado missiles as part of its' primary review responsibility for SRP

.f:

Sections' 3.3.1, 3.3.2, 3.5.3, 3.7.1 through 3.7.4, 3.8.4, and 3.8.5. 'The

' Mechanical Engineering Branch (MEB) determines that the comppnents piping l, and structures are designed in accordance with applicable codes =and standards as' part of its primary review respor.sibility. for SRP Sections 3.9.1 through N 3.9.3. The MEB, also,. determines the acceptability of the seismic and quality group classifications for system components as part of its primary review responsibil> y for SRP Sections 3.2.1 and 3.2.2. The MEB also review: the adequacy of the' inservice testing program of pumps and valves as part' of its primary review responsibility for SRP Section 3.9.6. The Materials Engineering Branch (MTEB) verifies that inservice inspection requirements:are met for system components as part of its primary review responsibility'for.SRP Section 6.6, and,'upon request, verifies the compati-bility of the materials of construction with services conditions. The

, review for Fire Protection, Technical Specifications, and Quality Assurance are coordinated and performed by the Chemical Engineering Brarch, Licensing Guidance Branch, and Quality Assurance Branch as part of their primary review responsibility for SRP Sections 9.5.1, 16.0 and 17.0, respectively. The EQB reviews the~ seismic qualifications of Category I instrumentation and electrical equipment and the environmental qualification of mechanical and electrical equipment as part of its primary review responsibility for SRP Sections 3.10 and 3.11, respectively. The Instrumentation and Control

( Systems Branch (ICSB) and the Pcwer Systems dranch (PSB) will verify the adequacy of the design, installation, inspection and testing all electrical systems (sensing, control ano power) required for proper operation of the SFPCCS as-part of their primary review responsibility for SRP Section 7.1 and Appendix 7-A for ICSB and SRP Section 8.3.1 for PSB. The Effluent Treatment Systems Branth (ETSB) will verify that the limits for radio-activity concentrations are not exceeded as part of its primary review responsibility for SRP Sections 11.1 and 11.2.

For those areas of review identified above as being the responsibility

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of other branches, the acceptance criteria and methods of review are contained in the SRP sections corresponding to those branches.

JL ACCEPTANCE CRITERIA Acceptability of the design of the spent fuel pool cooling and cleanup system, as. described in the applicant's safety analysis report (SAR), including related sections of Chapters 2 and 3 of the SAR is based on specific general design

, criteria and regulatory guides, and on independent calculations and staff judgments with respect to system functions and component selection.

J. The design of the spent fuel pool cooling and cleanup system and its makeup system is accepteble if the integrated design is in accordance with the following criteria:

~( L General Design Criterion 2, as related to structures housing the system and the system itself being capable of withstanding'the 9.1.3-3 Rev.1 - July 1981

't,g h

L effects of natural' phenomena such as earthquakes, tornadoes, and Murricanes. . Acceptance for meeting this criterion is based on con-formance.to. positions C.1,zC.2, C.6 and C.8 of Regulatory Guide 1.13 and position C.1 of Regulatory Guide 1.29 for. safety-related portions ,,

I land position C.2.of- Regulatory Guide 1.29 for nonsafety-related portions of:the system 4 This criterion does not apply to the. cleanup portion of the system and need not apply to the cooling system if the fuel pool makeup water system and its source, and the. fuel pool building and its ventilation and filtration system meet this' criterion, and-the ventilation and' filtration system meets the guidelines of Regula-tory Guide 1.52. .The' cooling and makeup system'should also be designed

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.to' Quality Group C requirements in accordance with Regulatory Guide 1.26.

However, when the cooling system is not designated Category I it need

.not meet the-requirements of ASME Section XI for inservice inspection of nuclear plant components.

- b. General Design Criterion 4, with respect to structures housing the

. systems and the system being capable of withstanding the effects of

' external missiles.L Acceptance is based on meeting position C.2 of-Regulatory Guide 1.13. This criterion does not apply to.the cleanup system:and need not apply lto the cooling water' system if the makeup system.and its source, and the building and its ventilation and

' filtration system are tornado protected and the ventilation and filtration system meets the guidelines of Regulatory Guide 1.52.

c. General Design Criterion 5, as related to shared systems and components important to safety being capable of performing required safety func-tions,

d. General Design Criterion'44, to include:

(1). The capability to transfer heat loads from safety-related structures, systems, and components to a heat sink under both normal operating and accident conditions.

'(2) Suitable redundancy of' components so that safety functions can be performed assuming a single active failure of a component-coincident with the loss of all offsite power.

(3) The. capability ta isolate components, systems, or piping, if required, so that the system safety function will not be compromised. l (4) In meeting this criteri$n acceptance is based on the recommenda-tions of Branch Technical Position ASB 9-2 for calculating the heat loads and the assumptions set forth in item 1.h of subsection III of this SRP section._ The temperature limitations of the pool water identified in item 1.d of subsection III of this SRP section is also used as a basis for meeting this criterion.

e General Design Criterion 45, as related to the design provisions to pe mit periodic inspection of safety-related components and equipment.

f. General Design Criterion 46, as related to the design provisions to permit operational functional testing of safety-related systems or components to assure structural integrity and system leak tightness, 9.1.3-4 Rev. 1 - July 1981

operability, and adequate performance of active system cornponents, and the capability of the integrated system to perform required functions during normal,: shutdown, and accident situations.

,,(

.g. General. Design Criterion 61, as related to the system design for fuel storage land handling'of radioactive materials, including the following

. elements:

~(1) The capability for periodic testing of components important to safety.

(2). Provision, for containment. ,

(3) Provisions for. decay heat removal.

(4) The capability to prever.t reduction in fuel. storage coolant inventory under accident conditions in accordance with the guidelines of position C.6 of Rogulatory Guide 1.13.

(5) The capability'and capacity to remove corrosion products, radioactive materials and' impurities from'the pool water and reducing occupational exposures to radiation.

h. General Design Criterion 63, as it relates to monitoring systems provided to. detect conditions that could result in the loss of decay heat' removal..to detect excessive radiation levels, and to initiate appropriate safety actions.

-( i. 10 CFR Part 20, paragraph 20.1(c) as it relates to radiation doses V being kept as low as is reasonably achievable (ALARA). In meeting this regulation Regulatory Guide 8.8, positions C.2.f(2) and C.2.f(3) will be used'.as a basis for acceptance.

III.. REVIEW PROCEDURES The procedures set forth below are used during the construction permit (CP) application review to determine that the design criteria and bases and the pre-liminary design as set forth in the preliminary safety analysis' report meet the acceptance. criteria given in subsection II of this SRP section. For the review of operating license (OL) applications, the review procedures and acceptance criteria and bases have been appropriately implemented in the final

' design as set forth in the final safety analysis report. The review procedures for OL. applications include a determination that the content and intent of the technical specifications prepared by the applicant are in agreement with the requirements for system testing, minimum performance, and surveillance developed as a result of the staff's review.

Upon request from the primary reviewer, the coordinating revies branches will provide. input for the areas of review stated in subsection I of this SRP section.

.The secondary review branch, CMEB, will provide an input on a routine basis for those areas of review indicated in this SRP section. The primary reviewer (ASB) obtains and uses such input as required to assure that this review procedure is complete.

lhe' review procedures given below are for a typicel system. Any variance of

( ths review, to take account of a proposed unique design, will be such as to 9.1.3-5 Rev. 1 - July 1981 ;

assure that the system meets the criteria of subsection II of this SRP section.

.In the evio ,,the spent fue1 pool cooling and cleanup system and its makeup system are evaluated with respect to their capability to perform the necessary ,

safety functions during all conditions, including normal operation, refueling, abnorm 4 storage conditions; and accident conditions.

1 The malety-function of the system for refueling and normal operations is identified by reviewing the information provided in the SAR pertaining tc the design bases and criteria and the safety evaluation section. The SAR section on the system functional. performance requirements is also reviewed to determine that it describes the minimum system heat transfer and system flow requirements for normal plant operation, component operational degrada-tion requirements (i.e., pump leakage, etc.) and describes the procedures, that will be followed to detect and correct these conditions should degrada-tion become excessive. The reviewer, using failure modes and effects analyses, determines that the~ system is capable-of sustaining the loss of any active component and evaluates, on the basis of previously approved systems or-independent calculations, that the minimum system requirements (cooling load and flow) are met for these failure conditions. The. system piping and instrumentation diagrams (P& ids), layout drawings, and component descriptions are then reviewed for the following points:

a. Essential' portions of the system are correctly identified and are isolable from the nonessential portions of the system. The P& ids are reviewed to verify that they clearly indicate the physical divi-sion between each portion and indicate required classification' changes.

System drawings are also reviewed to see that they show the means for accomplishing isolation and the system description is reviewed to identify minimum performance requirements for the isolation valves.

For the typical system, the drawings and description are reviewed to verify that adequate isolation valves-separate non-essential portions and components from the essential portions.

b. Heat exchangers, pumps, valves and piping for the cooling portion of

,the system are constructed to Quality Group C and designed to seismic Category I', requirements in accordance with the guidance provided in Regulatory Guides 1.26 and 1.29. As an acceptable alternative, the cooling loop may be constructed to Quality Group C and nonseismic Category I requirements provided the spent fuel pool water makeup system, and the building ventilation and filtration system are designed to seismic Category I requirements, are protected from the effects of tornadoes and meet the single failure requirements. The ventilation and filtration system must also meet the guidelines of Regulatory Guide 1.52. The review for seismic design is performed by SEB and the review for seismic and quality group classification is performed by MEB is indicated in subsection I of this SRP section.

'c. The stated quantity of fuel to be cooled by the spent fuel cooling system is consistent with the quantity of fuel stored, as stated in Section 9.1.2 of the SAR.

d. For the maximum normal heat load with normel cooling systems in opera-tien, and assuming a single active failure, the temperature of the pool should be kept at or below 140 F and the liquid level in the pool should be maintained. For the abnormal maximum heat load (full 9.1.3-6 Rev. 1 - July 1981

L h

core unicad) the temperature of the pool water should be kept below Doiling and the liquid level maintained with normal systems in opera-

.('. ticr.' A single active failure need not be considered for the abnormal E case- The. associated parameters for the decay heat load of the fuel E assemblies,-the temperature of the pool water, and the heatup time or rate of pool temperature rise for'the stated storage condition; are reviewed on the basis of independent analyses or comparative analyses of-pool conditions that have been previously found acceptable.

e. The spent fuel pool and cooling systems have been designed so that in the event of-failure of inlets, outlets, piping, or drains, the pool level will not be inadvertently drained below a point approxi-mately 10 feet above the top of .the active fuel. . Pipes or external 1ines extending into the poo1 that are equipped with siphon breakers, j' check valves, or other devices to prevent drainage are acceptable as l7 a means of implementing this requirement.

1 L f. .A seismic Category I makeup system and an appropriate backup method b to add' coolant to the spent fuel pool are provided. The backup system need not be a' permanently installed system, nor Category I, but must

.. ? 'take water from a. Category I source. Engineering judgment and compari-son with plants of similar design are used to determine that the makeup i capacities and the time required to make associated hookups are con-sistent with heatup times or expected leakage from structural damage,

g. .Desic)n provisions have been made that permit appropriate inservice inspection and functional testing of system components important to l jz safety. It will be acceptable if the SAR provides a statement that

( the spent fuel. pool cooling, makeup, and cleanup system is-included in the inservice inspection program per SRP Section 6.6 and the inservice testing program.of SRP Section'3.6.6. These SRP sections are reviewed by the MTEB and MEB respectively.

n. .The calculation for the maximum amount of thermal energy to be removed by the spent fuel' cooling system will be made in accordance with Branch Technical Position ASB 9-2, " Residual Decay Energy for Light-Water Reactors for Long-Term Cooling" (located in SRP Section 9.2.5) under the following assumed conditions.

i. The uncertainty f actor K is set equal to 0.1 for long-term cooling (greater than 107seconds).

ii The normal maximum spent fuel heat load is set at one refueling load at equilibrium conditions after 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months decay and one ref ueling lead to equilibrium conditions af ter one year decay (Maximum pool temperature 140*F) iii- The spent fuel pool cooling system should have the capacity to remove the decay heat,from one full core at equilibrium conditions after 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months decay and one refueling load at equilibrium condi-tions after 36 days decay, without spent fuel pool bulk water boiling. ' Cooling system single failure need not be considered concurr ent fur' t his condi tion.

o for pools with greater than 1-1/2 core capacity, one additional refueling batch at equilibrium conditions after 400 days decay should be included in the cooling requirements.

9.1.3-7 Rev. 1 - July 1981

},, 4 L

k The-rev, ewer verifies that the system has been designed so that system

. fur ctions will be maintained, as required, in the event of adverse natural phenomena _such as earthquakes, tornadoes, hurricanes, and floods. The ,

reviewer evaluates the~ system, using engineering judgment and the results o' failure modes and effects analyses to determine the following:

s. The failurb of portions'of the system, or-of other systems not designed to seismic Category'I standards and located close to essential portions of the system, or.of non-seismic Category I structures that house, support, or~are close to essential portions ofLthe pool and cooling D system,.will not preclude essential functions. Reference to SAR Chapter 2,' describing site features and the general arrangement and layout drawings, will be necessary as well as.to the SAR tabulation

.of seismic design classifications for structures and systems. State-ments in the SAR to the effect that the.above conditions are met are acceptable. '(CP) !

b. .Tne essential portions of the spent fuel pool cooling system are protected from the effects of floods, hurricanes, tornadoes, and internally or externally generated missiles. Flood protection and missile protection criteria are discussed and evaluated in detail-under'the SRP sections for Chapter 3 of the SAR.

The reviewer utilizes the procedures identified in these plans to

?-, assure that the analysts presented cre valia. A statement to the effect that the system is located in a seismic Category I structure that is tornado missile and flood protected, or that components of the system will be located in individual cubicles or rooms that will withstand the effects of both flooding and-missiles is acceptable.

The location and~ design of the system, structures, and pump rooms (cubicles)'are reviewed to determine that the degree of protection provided is adequate.

N. The system design information and drawings.are analyzed to assure that the following features will be_ incorporated. A statement that these features will be included.in the design by some appropriate means is a basis for acceptance. (CP)

a. A leakage detection system is provided to detect component or system f leakage. An adequate means for implementing this requirement is to i provide sumps or drains with adequate capacity and appropriate alarms 1 in the immediate area of the system.

b. Components and headers of the system are designed to provide individual isolation capabilities to assure system function, control system leakage, and allow system maintenance.

c. Design provisions are made to assure the capability to detect leakage of radioactivity or chemical contamination from one system to another and to preclude long-term corrosion, organic fouling, or the spreading of radioactivity. Radioactivity monitors and conductivity monitors located in the system discharge lines are acceptable means for imple-menting this requirement. ,

4. The SAF descriptive information, P& ids, layout drawings, and system analyses are reviewed to assure that essential portions of the system will function 9.1.3-8 Rev. 1 - July 1981

p fjliowingdesignbasisaccidents,assumingaconcurrentsingleactive component failure. The reviewer evaluates failure mode and effects analyses-presented in the SAR to assure function of required components, trace the

-h.- availability of-these components on~ system drawings, and check that minimum

-system flow, makeup,'and heat transfer requirements are met for each degraded situation over the requireo time. spans. For each case the design

]

- will be acceptable.if minimum systea requirements are met. ,

5. 'The spent fuel pool cleanup system'and various auxiliary systems are designated as'nons9fety related systems and are designed accordingly (nonseismic Category I). These systems are evaluated to assure that.

.their' failure cannot affect the functional performance of any safety-related system or component. The relationship and proximity between the nonsafety-related system and safety-related systems or components-arE determined by reviewing the integrated' structure and compone_nt layout' diagrams. Independent analyses, engineering judgement, and comparisons with previously approved systems are used to verify that where a nonsafety-related system interconnects or interfaces with the cooling system, its failure by any event or malfunction will not' preclude adequate functional ;

performance of the cooling system.

7. The cleanup system is also reviewed to assure that it has been designed with the capability to maintain acceptable pool water conditions. The P&lDS and associated information provided in the SAR is reviewed to verify the following:

a. A'means has been provided for mixing to produce a uniform temperature throughout the pool.

b. The cleanup system is reviewed by CMEB to verify they have the capacity and capability to remove corrosion products, radioactive materials, and impurities so that water clarity and quality will enable safe operating conditions in the pool. This includes instrumentation and sampling to monitor the water purity and need for demineralized resin replacement'. including the chemical and radiochemical limits such as conductivity,. gross gamma and iodine activity, demineralized dif-ferential pressure, pH and crud level which are used to initiate c< ective action,

c. The capability for processing the refueling canal coolant during refueling operations has been provided.

d. Provisions to preclude the inadvertent transfer of spent filter and demineralized media to any place other than the radwaste facility have been provided.

IV. EVALUATION FINDINGS L

The reviewer verifies that sufficient information.has been provided and that his review supports conclusions of the following type, to be included in the j staff's safety evaluation report: ;

The spent fuel pool cooling and cleanup systeni includes all components and piping of the system from inlet to and exit from the storage pool and j

( pits, the seismic Category I water source and piping used for fuel pool makeup, the cleanup system filter-demineralizers and the regenerative l

9.1.3-9 Rev. 1 - July 1981 1' i

l proc es to the point of discharge to the radwaste system. The scope of review M the spent fuel pool cooling and cleanup system included layout dran ny , process flow diagrams, piping and instrumentation diagrams, ana t

y descriptive information for the system and the supporting systems that ws i are etsential to safe operation. The cooling portion of the system and !

the (mcrgency primary makeup system are designed to seismic Category 1, Qualit, Group C requirements since they are necessary to remove decay heat from the spent fuel and to prevent fuel damage that could lead to unacceptable releases of radioactivity. The cooling portion of the system need not be designed to seismic Category I requirements if the makeup system and the building ventilation and filtration system are seismic Category I, and if the ventilation and filtration system meet the guidelines of k gH =tnry Guido 1.52.

The stah cencludes that the design of the spent fuel pool cooling and cleanup system aM its- makeup system meets the requirements of General Design Criteria 2, 4, S., 44. 45, 46, 61 and 63. This conclusion is based on the following:

1. The applicant has met the requirements of General Design Criterion 2 with !

respect to safety-relatec' portions of the system being protected against natural phenomena. Acceptance is based on meeting the guidelines of Regulatory Guide 1.13, position C.1 which recommends a seismic Category I design fcr necessary portions of the spent fuel storage facility, position C.2 regarding protection against winds and wind generated missiles, posi-tio; C.6 as it relates to the system being capable of withstanding earth-quakes without loss of coolant that would uncover the fuel, and position C.8 ahich recommends a seismic Category 1 makeup system with appropriate redundancy or a backup from a Category I water source. Acceptance is also baseri on meeting the seismic design requirements of Regulatory Guide 1.29, position C.1 for safety-related portions of the system necessary for adenuate cooling to prevent excessive radioactivity releases (position C.I.p

'f Pegulatory Guide 1.29) and position C.2 as it relates to the failure of m safety-related portions of the system. If the fuel pool building ventilation and filtration systems are designed to seismic Category I requirements and in accordance with the guidelines of Regulatory Guide 1.52 the n d ing portion of the system need not be seismic Category 1. !

2 The 1esign meets the requirements of General Design Criterion 4 with regards to protection against the effects of externally generated missiles since )

it is in accordance with position C.2 of Regulatory Guide 1.13 since no 4 loss of watertight integrity or fuel damage occur in the event of tornado miss41re ,

1 The h ign meets the requirements of General Design Criterion 5 regarding thc sharing of safety-related structures, systems, and components since no single failure will prevent the system from performing its safety-related function which is cooling the spent fuel.

4 The design meets the requirements of General Design Criterion 44 regarding decay heat removal redundancy and power supplies, since the system has the capability to remove decay heat from the spent fuel under both normal operating and accident conditions. The system has redundancy so that decay heat can be removed assuring a single active failure coincident with a loss of all offsite pov.er, and is designed with isolation capability of I syster components and piping, if required, such that th? ability of the l system te remose decay heat will not be compromised i 9.1.3-10 Rev.1 - July 1981 l

l l

_L

The system meets:the inspection and testing requirements of General Design

.F Criteria 45 and .46 since the. system is designed and constructed with L . suitable clearances and: location to allow periodic inspection of major h-- " components, and is designed to permit functional operational testing to

assure; structural integrity and system leak tightness, operability, and adequate performance.of active: system components.

- 6. The. system is designed in accordance with the requirements of General

-Design Criterion 61 as it relates to the system design for fuel storage since'the system has the following design capabilities: the system has the capability for periodic testing of components important to safety.

The system.is designed to provide suitable shielding by maintaining a minimum' water. level above the fuel. There is redundancy.and testability of the decay heat removal portions of the system, and the system is designed to prevent reduction in fuel storage coolant inventory under accident conditions in accordance with position C.6 of Regulatory Guide 1.13. The spent fuel pool cleanup portion of the-system (1)'provides the l capability and capacity of removing radioactive materials', corrosion.

products, and impurities from the pool water and thus meets the require-ments of Criterion 61 as it relates to appropriate filtering systems for fuel cooling'and storage, (2) reduces occupational exposure to radiation by removing radioactive. materials from the pool water and thus meets the requirements of 10 CFR Part 20, $20.1(c) as it relates to maintaining radiation exposures as low as reasonably achievable (ALARA) and, (3) retains radioactive materials-and crud in the pool _ water in the dsmineralizer and filters and thus meets positions C.2.f(2) and (3) of Regulatory Guide 8.8.

7. .The system design meets the requirements of General Design Criterion 63 since it has provisions to detect the loss of heat removal function through

(' the use of. loss of flow and temperature alarms, and to detect conditions that would result in excessive radiation through the use of coolant low

' level alarms and radiation monitoring alarms. And the system has the

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capability to initiate appropriate safety actions since it has an automatic makeup system and the cooling system and ventilation and filtration system can be-operated from the control room in the event of high radiation or low level alarms.

V. IMPLEMENTATION The following is intended to provide guidance to applicants and licensees

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regarding the NRC staff's plans for using this SRP section.

Except'in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.

Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced Regulatory Guides.

'G REFERENCES i 10 CFR Part 20, s20.1(c), " General Provisions for Standards for Protection Against Radiation."

(

1 9.1.3-11 Rev.1 - July 1981 l

2 10 rFR Part 50, Appendix A, General Design Criterion 2, " Design Beses for Protection Against Natural Phenomena."

r 10 CrR Part 50, Appendix A, General Design Criterion 4, " Environmental -

7.nc Missile Design Bases."

4 10 CFR Part 50, Appendix A, General Design Criterion 5, " Sharing of Structures, Systems and Components."

5. 10 CFR Part 50, Appendix A, General Design Criterion 44, " Cooling Water."

6. 10 CFR Part 50, Appendix A, General Design Criterion 45, " Inspection of Cocling Water System." 'j

). 10 CFR Part 50, Appendix A, General Design Criterion 46, " Testing of Cooling Water System. "

6. 10 CFR Part 50, Appendix A, General Design Criterion 61, " Fuel Storage and Handling and Radioactivity Control."

9. 10 CFR Part 50, Appendix A, General Design Criterion 63, " Monitoring Fuel and Waste Storage."

30. Regulatory Guide 1.13, " Fuel Storage Facility Design Basis."

11. Regulatory Guide 1.26 " Quality Group Classifications and Standards for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power Plants."

12. Regulatory Guide 1.29, " Seismic Design Classification."

11 Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plantr."

14. Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable."

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l 9.1.3-12 Rev. 1 - July 1981 ,

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AE0D/C503 CASE STUDY.. REPORT *

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DECAY HEAT REMOVAL PROBLEMS AT U.S. PRESSURIZED WATER REACTORS Decenber 1985 Prepared by: Dr. Harold Ornstein Office for Analysis and Evalua. tion of Operational Data U.S. Fuclear Reculatory Commission

This report documents results of a study completed to date by the Office for Analysis and Evaluation of Operational Data with regard to a particular operational situation, The findings and recommendations do not necessarily represent the position or requirements of the responsible program office or the Nuclear Regulatory Commission.

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.. -TABLE ,0F : CONTENTS *. ' ' .. . .. l j

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' EXECUTIVE SUMMAFY=. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

. 1.0!NT,PODUCTION . . . . . . . . . . . . , . . . . . . ... . . . . . . . . 3

.. .. ,, ,. i . .

2 c0 : DEC AY. HE AT RkM0YAL. SYSTEM *. . . . . . ' . . . . . ' . . . . . '. . . . . ' . ' . .5 l

, 2.1 functional Description and System Design . . . . . .......5

~

2.2 Consequences.of the loss of.the' Decay Heat. Removal Function . . . 12.

2.3 -Actions to' Recover the Decay Heat Removal Function . . . . . . 15 h 3.0 OPERATIONAL EXPERIENCE . . . . . . . . . . . . . . . . . . . . . . . . 17 .

3.1 Losses of Decay. Heat Removal Systems . . . . . . . . . . . . . 17-4.0 ANALYSIS-ARD EVALUATION OF THE UNDERLYING OR ROOT LAUSES OF REPOPTED 25' LDECAY HEAT REMOVAL SYSTEM LOSSES . . . . . . . . . . . . . . . . . .

4.1 Human Factors . . . . . . . . . . . . . . . . . . . . . . . . . 25 .;

i

.4.2 Equipment. Failures ......................30

(. 4.3 Technical Specification Deficiencies . . . . . . . . . . . . . 31 l

i 5.0 FINDINGS AND CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . .'34 5.1- Human Factnrs Considerations .................35

.5.2 Desion Considerations - Flow Path fron.the Reactor Coolant

_ System to the Decay Heat Removal System . . . . . ......36 5.3 Technical Specification Deficiencies . . . . . . . ......38 6.0 RECOMMENDATIONS ..........................40 L

7.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 APPEND]CES ...............................46 Appendix A - Loss of Decay Heat Removal Systems et U.S.

Pressurized Water Reactors During 1982 and 1983 . . . . 46

'. Appendix B - Selected Loss of Decay Heat Removal Systen Events at U.S. Pressurized Water Reactors During 1984 . . . . . . . 51 Appendix C - Decay Heat Removal System Losses at Davis-desse . . . . 53 1

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.' Fiq. ure l'. . Sche'niatic Diagram of..DHR .Systsms at.tCS. 'PWRs'

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,Line DHR System Configuration. ... . ... . ; . . . . . 10 ..

Figure 3. Schematic. Diagram of the Davis-Besse Plant's DHP Suction Bypass Line Configuration . . . . . . . .- . . 11 Figure 4. 'DHR Recovery Time Margin . . . . . . . . . . . . . . . 14 I . . y I

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LIST OF' TABLES i 1

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1 Table 'i. Plan.t Operational Modes . . . . . . . . . . . . . . . .6 l Table,2. Some Backup Methods for Decay Heat Removal Upon Loss j of'the'DHR System . . . . . ............. 16' )

1 Table 3. Frequency of DHP Losses . . . . . . . . . . . . . . . 18 Table 4 Duration of Reported DHR Loss Events ........ 20 f

. Table S. Categories of 130 Reported Total DHP System Failures When Required to Operate (Loss of Function) at U.S.

PWPs 1976-1983 . . . ............,... 23 q Table 6. Underlying or Root Causes of Reported DHP System Failures . . . . . . ................ 26

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4 e , - : s, 3; (1 .- ~~,. EXECOTIVE SUtiMARY .

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The report analyzes U 5. pressurized water reactor TPWR)'e'xperience

, involv.ing. loss;of-operating decay heat removal (DHR)(systems. Between 1976

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" and '1983','130' loss-of-DHR events were report'ed to have occurred during ,

approximately 50.0 reactor. years.of operation. The DHR systems are-not-

.safetygradesystensfonmost'operatin@. plants..Onlyth'pseplants1.that,have,

'the DHR systems , designed ahd constructed'to NRC's-Branch Technical Position RSB 5-1 have'truly safety grade systems; Total. loss of the_DHR systems j under certain conditions could lead to core uncovery, and resultant fuel <

damage. .The results of scoping analyses of total loss-of-DHR scenarios l presented in this study indicate that for certain postulated events, unless-timely corrective actions are'taken, core uncovery could result on the order of one to three hours. To date, no serious damage has resulted from the

. loss-of-DHR system events that have occurred at U.S. PWRs. However,.many of the events which have occurr'ed thus far may serve as important precursors to more serious events.

Analysis of operating data indicates that the underlying or root causes of most of the loss-of-DHR system events are-human factors deficiencies involving procedural inadequacies and personnel error. Most of the errors were committed during maintenance, testing and repair operations.

The leading category of loss-of-DHR events (37 of 130) was the iriadvertent' automatic closure of-the suction / isolation valves, most of which resulted from

(. human errors, This repor.t presents summaries.of loss-of-DHR events which occurred during the years 1982 and 1983, and the'most significant loss-of-DHR events in 1984. Since the 1984 data base was not complete at the time of this analysis, the available 1984 data was used to confirm the observed trends in frequency and severity of loss-of-DHR events for prior years. Reference is made to an industry report, Nuclear Safety Analysis Center report NSAC-52, for summaries of loss-of-DHR events which occurred during the years 1976 through 1981. Reference is also made to the findings of,a recent probabilistic risk assessment that was done by the Nuclear Safety Analysis Center to quantify the risks associated with DHR (NSAC-84).

The analysis of recent operating data indicates that the situation involving loss-of-DHR systems is not improving. In terms of the frequency or duration of loss, no clear trend towards improvement is evident. However, it may be too early to see the results of the implementation of the recommendations contained in NSAC-52 or industry actions.

This report makes several recommendations based upon the potential safety si<Jnificance of loss-of-OHR events. Implementation of those recommendations should significantly improve DHR system reliability and availability. The recommendations-include; ' improving human factors by upgrading coordination, planning, and. administrative control of surveillance, maintenance, and testing operations which are performed during shutdown; providing operator aids to. assist in determining time available for DHR recovery an1 to assist b operators in trendino parameters during loss-of-DHR events: upgrading the

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' t'raiking 'aniqualNic'a'tionIeku'fre'm'eni.s, fit opei akibns. aidi dillitesahceb ' ' '

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' staff; requiring the'u.es of reliable, well-analyzed methods"for' measuring !

- reactor vessel level during shdtdown modes; modifyi'ng plant. des'ign to remove ) i

. autodios~Ure 'in'terlocks and/or power tolthe DHR.suetio6/ isolation Ivahes . . . . . '.

l during periods'~which do not' require valve'motionj'an'd clarifying. plant. .

l .. , , , l techni cal s_pecificat.f onyt,o' e.lliningte ambigditj,'es ' ass'ociated,with' operating' mode ~ definitions. '

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Thereportacknow'1edgesNRC's"ongoingeffortsto'.addiessshutdobndecay. heat removal requirements (Unresolved Safety Issue A-45). ,

tions are applicable to A-45, and should'be consider.The ed in the resoluf. AEOD*recommenda-ion of !

this generic issue.

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'l.0; INTRODUCTION' , 9

, The pur: pose.of.this s'tudy is'to evalyate..operationallexperience.and c to..

.. analyze,the. safetyjtnpl.icationsf assoc.i.ated:with?totalfloss of. decay heat:

removal (DHR) systems;-[also referred to'as the residual heat removal (RHR) systems, and shutd

- ' reactors -(PWRs).* own ; cooling (SDC-) ,sys.tems)?

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U.S. pressurized','wate,r c

The safety . function of the.DF{R.hstem.'is' tii./tiansferlfis$1on. product decay?

heat'from the' reactor core at a rate which will assure that the fuel P sign limits andfthe reactor coolant pressure boundary design limits are not exceeded. An extended loss of the DHR function could lead to core uncovery, and associated fuel damage. (As noted in section 2.1, in addition to the DHR system there are systems which can be used to remove decay heat--e.g.

steam generators and the auxiliary feedwater system.)

We note that NRC's' General Design Criterion (GDC) 34 requires the DHR safety function to be accomplished assuming a single failure. However, we also note that there have been numerous single failures which have caused losses of this system. During the years 1976 through 1983, the licensees have reported about'130 loss-of-DHR events. Most of those events were of a short duration; however, several have extended beyond an hour with some lasting more than two hours.

To date, no serious damage has resulted from the total loss-of-DHR systems at U.S. PWRs, and there has not been any danger to the public. Nonetheless, the large number of loss-of-DHR events which have occurred thus far

( (occurrence frequency of 0.25/ reactor year), may serve as important precursors which warrant corrective actions before a far more serious event occurs.

Numerous studies have been performed and numerous reports have been written on DHR systems. The most significant enes are:

o In 1975, WASH 1800 (Ref. 1) noted that the loss of the DHR function In this report, a total loss of a DHR system is defined as failure of both DHR (RHR, or SDC) trains to perform their function when required.

Such losses include momentary as well as long duration events.

Momentary events represent challenges to plant safety from'which timely recovery has-taken place. We view the many short and long duration loss-of-DHR events which have occurred thus far without any serious consequences as precursors. Analysis and evaluation of those events are necessary because timely recovery from similarly initiated events cannot always be assured. Inoperability of both DHR (RHR, or SDC)

. trains during times that they are not required to perform their function are not included as total losses. Similarly, DHR (RHR or SDC) systems which are administrative 1y declared inoperable, but could still perform their function (e.g., inoperability due to missed surveillance or faulty snubbers), are not included as total losses, i I

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c l subsequent to a ~iransient can be'al potent ~ial')y..significanticontributor

': to the' total" risk associated w'i.th nucleat' power plants.

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' o 'T6?1980,L.'th'e NRC declared "Shutd'own' Decayiieat 'Remov'ai RequN'ements"

' ~ iIn U'nresolved Safety Issu'e (USI)':" SubYequ'entli, th'e' Office of Nuclear g

! Reactor Regulation (NRR.) has implemented task. action plan A.-45;to -

fesol'ye this .itsue.'lThe overall purpose of.A-45-is .ot ev.aluate the s, .

adequacy of current. licensing design. requirements i'n' order to ensure.

I o that: nuclear. power piknts do. hot pose,an, Unacceptable risk due to

failure'to remove shutdown decay. heat. -

o In 1982, Oak Ridge National Lab' oratory (ORNL) evaluated events involving DHR systems in U.S. PWRs and.U.S. boiling water reactors (BWRs) for the period June 1979 to June 1981 (Ref. 2).

ORNL found 38 loss-of-DHR system events which met their criteria for safety significance (which is equival.ent to our definition of a total loss of a DHR system). ,,

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o In 1983, the Nuclear ~ Safety' Analysis Center (NSAC) published a.

report (NSAC-52,.Ref. 3) which reviewed DHR losses at U.S. PWRs during the years 1976 through 1981. It made numerous recommendations which, if implemented, could have improved DHR system reliability and overall safety.

Interest in this case study was first initiated because of the large number of loss-of-DHR events which occurred at the Davis-Besse plant (see Appendix C). Subsequen't analysis of the data, and additional licensee event reports (LERs) detailing DHR losses at other PWRs showed that the problems

- at the Davis-Besse plant were not.uniq'ue to Dav.is-Besse or.other Babcock and

.) ;

Wilcox (B&W) plants. As a result,-the scope of the study also includes events which occurred at PWRs having reactors designed by Combustion Engineering (CE) and Westinghouse (W).

This report' highlights'some facets of DHR losses and DHR operations which are not addressed in previous reports, and it presents six recommendations which, if implemented, have the potential for significantly improving reactor safety.

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F 2.0 ' DECAY HEAT REMOVAL': SYSTEM ' -

[ 2.1 Functional Description and'59 stem Des _ign i -

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The DHR system is designed to' remove fission product decay heat from the reactor ~ core. The safety function of.the DHR system is to remove heat from.

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the' p'rimaty sys't'em at' 'a rateith'atwill1." enable operstor's'. to blririg ttie' p1 ant '

from hot shutdown conditions to cold shutdown'or refueling conditions (see T.able 1), and tof'm.aintain"the. plant in'soch shutd'own" conditions'for extended.

perio'ds 'of ' time. For the transition phase associated with cooling the plant from operating pressures and temperatures after a reactor trip, for example, to hot shutdown, the steam generators and the auxiliary feedwater system are used to remove heat from the primary system. Upon reaching the reduced pressures and temperatures associated with the hot shutdown condition, the DHR system is activated.

During accident conditions, most DHR systems can be aligned to perform emergency core cooling functions [ low pressure. coolant injection. (LPCI),

recirculation,' and in some designs containment spray]. In W and B&W plants, the DHR system can also act as a booster system to provide the net positive suction head (NPSH) required by the safety injection (SI) or high pressure injection (HPI) pumps for operation in the recirculation mode (" piggyback" operation). In B&W plants, the DHR system also provides auxiliary spray to the pressurizer to assist in depressurization after the reactor coolant pumps are secured.

The DHR system is typically composed of two redundant 100% capacity trains.

I It is usually located outside containment. A schematic diagram of a repre-sentative DHR system appears in Figure 1. Most DHR systems have a single suction or "drpp" line which.is tapped off one of the reactor coolant system (RCS) hot legs. Because of the single suction or " drop" line design, most DHR systems are susceptible to loss.of the ability to perform the decay heat removal function due to a single failure of a suction line valve. It should i be noted that for most DHR systems, much of the systems piping and compo-nents are fully safety grade because they are used for the LPCI function.

In fact for most DHR systems, single failure vulnerability exists in the single suction or " drop" line which is not used with the'LPCI function.

From the DHR pump discharge, the primary coolant flows through a heat ex-changer where heat is transferred to the component cooling water system. ,

Af ter the' primary coolant leaves the DHR heat exchangers, it returns to the {

reactor vessel. There are not many significant differences among DHR (

systems at U.S.'PWRs. One significant difference is the location at which I the DHR flow returns to the RCS. In B&W plants, the DHR flow returns to the l reactor vessel through piping which is shared with the core flood tanks' l dischrrge. In other PWR designs the DHR flow returns to the reactor vessel j throt.gh the cold legs. Other differences are: number of DHR system trains ;

(2 vs. 3 trains), number of " drop" lines (newer plants having 2 vs. 1), W '

plants having letdown coming off the DHR system, and CE system 80 plants having single failure proof suction / isolation valve closure logic.

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Most DHR systems operate at temperatures of 350*F or less, and at pressures )

less than 600 psig. Because the DHR system has a low pressure design and is located outside containment, significant efforts (administrative controls, system interlocks, etc.) are made to ensure isolation of the system when the d

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Reactivity

% of Rated Coolant Operational Mode Condition, K,ff Thermal Power ** Temperature i

1. POWER OPERATION > 0.99 > 5%

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2. STARTUP. > 0.99 - < 5% >

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3. HOT STANDBY 1 -

<.0.99 0 - >l

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4. HOT SHUTDOWN < (,,9 0 .(TDHR) F> T,yg> 200*F

5. COLD SHUTDOWN < 0.99 0 < 200*F

6. REFUELING *** -< 0.95 0 -< 140'F .

T DHR= temperature at which the DHR system is initiated (generally 280*F - 350*F) ;

As defined in B&W, CE, and W standard technical specifications (e.g.,

Ref. 4). Note - many plants do not use standard technical speci-

-fications.

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RCS$ pressure Lexceeds t'hb DHR'sys'temlesign pressure. Overp res,suri z at'iork a'nd the potential rupture of the low pressure system .is commonly referred to as an " interfacing LOCA - Event V." WASH-1400 (Ref:.1) s.howed that f or. the PWR . .

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' studied)($urry, a"W ple.nt).,. ' Event,v :.cou'id.'.represcentL a.higti ri.sk', core damage. l accident' sequence. , l

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.DHR' system' requirements contained in tha'gener i desi.gn criteria (GDC) have. j

. changed over the years. The 1967 GDC:did not address single f.ailure aspects

. ? :of.the DHR system. The 1971 GDC,~criteriori. number:34, requires the.DHR system to meet the single failure criterion. Newer' plants which are.

designed to the 1971 GDC do not f ully meet GDC 34's single failure criterion. "

Although some of the newer plants do have double drop lines and redundant valves, the control circuitry in most of those plants is such that a single failure can cause a lots of the DHR system. . Recognizing this, NRC has declared " Shutdown Decay Heat Removal Require.ments" an Unresolved Safety Issue (A-45). To resolve A-45,.NRR is evaluating the adequacy of current licensing' design req'ireme'nts u in order to ensure that plants do not pose an unacceptable risk due to failure to. remove shutdown decay heat. The specific objectives of tne task incluoe:

o Assess the safety adequacy of DHR systems in existing power plants for achieving both hot shutdown and cold shutdown conditions.

o Evaluate the feasibility of alternative measures for improving DHR system reliability, including diverse alternatives dedicated to the DHR function.

o Assess the value and impact (or cost-benefit) of the most promising ,

alternative measures.

o Develop a plan for implementing any proposed new licensing requirements for DHR systems.

Many plants are designed such that single failures in the DHR suction /

isolation valve interlocks, and single instrument bus failures or single valve failures can result in the loss of the decay heat removal function of the DHR systems. (The LPCI f unction of the DHR system is not vulnerable to sutb single failure.) The DHR suction / isolation valve interlocks are des ( i to prevent an interfacing LOCA (Event V) at the expense of interi yting DHR system operation. The functions of the interlocks are to:

o Prevent opening of the suct' ion / isolation valves when the reactor coolant system (RCS) pressure exceeds the DHR system pressure.

o Assure that the suction / isolation valves are closed for plant startup and repressurization.

Irf essence, the suction / isolation valve logic is single failure proof with regbrd to closing the valves (to prevent an interfacing LOCA), but it ne-cessitates an interruption of the DHR system function. Justification for thi* prioritization (interf acing LOCA first, and decay heat removal second) is based upon the design decision that there is less recovery time, and greater risk, associated with the interfacing LOCA than with a loss of DHR.

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. '5.- a din an'ef fort'to ' reduce the'"s'thgie2 fail'ure vuln'erability of .the.DHR systems -

some recent' designs have two " drop lines" from the RCS, and two' suction j lines,'-each having motor-operated isolation valves in series (Fig. 2).

. ,d. , However, from,theastandpoint.. of. the. interfacing LOCA, the double drcp line,-

do'uble su-tion line configuration' presents ~an additional . failure path and, theref ore, ~results i,rf a . higher. ris'k than the single suction line' configuration.

Most.of .the plants *f which. have'-two drop lines and two', suction' lines have OHR suction / isolation valve closure logic which would close valves in both lines as a result of a single failure (e.g.. control logic failure that' closes the ,

valves, or a single erroneous' closure signal). Consequently, a single '

active failure could cause the loss of the decay heat removal function of the DHR system for such plants. For example, as noted in Reference 5, the Catawba plants, which have two drop lines, can lose both trains of DHR due to a single instrument bus failure. It should be noted that the control logic for,most double drop line designs is . fail-safe--where fail-safe implies preventing an interfacing LOCA, not sustaining DHR flow. As a result, the double drop lire designs with the aforementioned closure logic do not represent any improvement against loss-of-DHR events associated with automatic closure of the suction / isolation valves. They do present a significant improvement against " stuck shut valve" problems.

As noted in Reference 6, Westinghouse has evaluated a recent licensee proposal to remove the autoclosure interlocks on the Kewaunee plant's DHR suction vahes, The analysis concluded that for Kewaunee, such a modification would be a safety improvement. FRR has approved that modification.

In an attempt to provide redundant DHR flow. paths, while minimizing the possibility of an interfacing LOCA outside containment, the Davis-Besse plant has,a configuration whi.ch lies between the single and double drop line configurations. The Davis-Besse plant has one drop line with a smaller diameter bypass line as shown in figure 3. The valves in the bypass line are manually operated (normally closed). This bypass configuration provides an additional flow path to enable DHR cooling in the event there is a problem with the suction / isolation valves; yet, it does not provide the

Some plants with a double drop line/ double suction line configuration are:

Palo Verde 1, 2, 3 Vogtle 1, 2 San Onofre 2, 3 Shearon Harris 1 WPPS 3 Comanche Peak 1, 1

. Kewaunee Beaver Valley 2 Catawba South Texas 1, 2 Callaway Byron 1, 2 Summer Breidwood 1, 2 Farley 1, 2

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SCHEMATIC DIAGRAM OF DOUBLE DROP LINE, DOUBLE SUCTION LINE DHR SYSTEM CONFIGURATION

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" DROP LINES" FROM TWO RCS I HOT LEGS ' i m

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SCHEMATIC DIAGRAM OF THE DAVIS-BESSE PLANT'S i

DHR SUCTION BYP' ASS LINE CO,NFIGURATION

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riskl.for. a' LDCAJ outside containment'thit'. is1.i'nherent> in "

' a doub1E drop line configuration. '

. : 2.2. Tonse' qui.nces of th' e' Loss' of. the Decay. Heat Removal Fun'ction

. The time margin available for restoring 'the DHR system, or establii,hing alternate methods of1 heat removal (prior to bulk boiling, core. uncovery, fuel' damage,.etc.) depends'upon1the RCS temperature, the decay heat rate.

(which is' dependent upori time interval elapsed from rea'ctor trip to' DHR' '

system failure and core power operating history), and the amount of 'RCS inventory. During some shutdown operations, the RCS may be. partially drained (e.g., to perform steam generator inspections or rep. airs).

Decreased primary system inventory can significantly reduce the time )

available to recover the DHR function prior to bulk boiling and core uncovery.

It should also be noted that reduced primary' system inventory can result in-rapid heatup rates and decreased time margins available prior to primary system boilof f even though the DHR loss may happen many days af ter shutdown.

For example, Reference 7 indicated Sequoyah 2 had a 92*F heatup of the primary system water in 77 minutes with reduced RCS inventory, even though the reactor was shutdown 18 days before the DHR loss occurred.

The results of an AE00 scoping calculation showed that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a reactor trip, in the absence of successful operator action, a partially drained RCS at a B&W plant could boil off enough coolant to uncover the top of the core approximately one hour after losing the DHR system. The calculation was based upon the assumptions that the RCS was drained down to the top of the. hot-leg nozzle, the coolant in the reactor vessel was at a.

bulk temperature of.140 F, and the RCS was "open" to 'the .atmosphpe.

This analysis compares favorably with operating experience and licensees' analyses. For er. ample, References 8, 9 and 10 reported that on August 29, j 1984, 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months after a reactor trip at ANO-2 (a CE plant), the DHR system failed while*the plant was in a partially drained condition. The reactor vessel water heated up from 140*F to 205*F in about 30 to 40 minutes.* In Reference 11, D.C. Cook (a W plant) reported the results of a corresponding }

analysis indicated that the onset of core uncovery would take place about I hour after the loss of the DHR system. Recently, a foreign PWR experienced a loss-of-DHR during draindown. Subsequently, the foreign country's regulatory body performed a calculation and concluded that,under slightly different conditions, in the absence of successful operator action, core uncovery could begin in about 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 20 minutes, with fuel failure beginning about two and one-half hours after the loss of the DHR system.

The entire event lasted 50 minutes. However, during the first 20 minutes, DHR was provit by make-up water which was gravity fed from the RWST, and by oscill.' ery DHR system flow (flow was provided by the DHR pump which was cavitating).

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- The time available'for' recovery'wod1d be mludh chortir T' O if' the. 08R los's'is init'ia'ted by a LOCA'(such as two events at Sequoyah which . involved inadver-tent opening of the, containment spray system). . A LOCA could cause a rapid

, , vessel draindown .resul, tin.g.in .the . loss-of-DHR pump suctiona NSAC. calcul a-:;

tions indicate that under such circumstances, core uncovery could begin in.

, 25. minutes (Ref. 3). - -

Another AEOD scoping calculationfwas performed.for.a loss-of-DHR, event at a

.. B&W plant shortly ~after activating the DHR sy' stem. It was' based.upon,a. c licensee' calculation (Ref. 12) which assumed a loss'of the DHR system about 3

three hours after reactor trip with a full RCS (no draindown). The'results indicated that in the absence of successful operator action, the RCS would

. heat-up to saturation conditions,'and pressurize to the low temperature overpressurization (LTOP) setpoint within one-half hour.* Upon reaching the LTOP setpoint, the RCS coolant would boil off at the LTOP setpoint pressure and escape through the PORV. The results of our calculations indicate that

' core.uncovery would begin.within about, two and one-half! hours af ter the loss-of-DHR. occurred.

The time available for restoring the decay heat removal function prior to core uncovery can be as short as about an hour and can extend up to many hours or days.- The time available depends upon the plant's operating history and status at the time of the DHR system loss. Figure 4 shows a typical time margin plot (time available for recovery of the DHR f. unction vs. time after rod insertion that the DHR function was lost).

The results of AE0D's scoping calculations indicate that if losses of the

('

DHR system occur during early stages of shutdown (e.g., 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor trip) with the RCS partially drained, or shortly after activating the DHR system before the. primary system is drained, corrective actions must; be taken'promptly to either restore the DHR system or to imp'lement alternate met, hods for removing reactor decay heat. These calculations highlight the fact that a loss of the DHR system can lead to a safety significant event l unless timely recovery actions are taken.

Historically, the NRC and the U.S. nuclear community have considered hot standby to be a safe end state. As a result, until recently no probabilistic risk assessments of U.S. reactors have quantified the risks associated with operations in shutdown modes 4, 5, and 6.

Because of recent interest in DHR, NSAC is funding probabilistic assessments of the risks associated with modes 4, 5 and 5 at two plants (one'PWR and one BWR).

L

Similarly, as noted in Reference 13, closure of the suction isolation

. valves at W plants have led to rapid LTOPs. They occurred because the l plants were operating solid, and closure of the suction / isolation L

valves isolated letdown without stopping make-up. In one case, the situation was further aggravated by the fact that the RCPs were operating, and the LTOP relief valves were inoperable.

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. , .DHR RECOVERY TIME' MARGIN..

' TIME TO CORE UNCOVERY 6 -

BEGINNING OF' LIFE - ' '

OF REACTOR FUEL 3 DAYS AT 100% POWER

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O 10 20 30

- TIME AFTER ROD INSERTION AT WHICH DHR LOSS OCCURS (days)

STARTING CONDITIONS: RCS DRAINED TO HOT LEG CL, OPEN TO ATMOSPHERE.

AVER AGE TEMPERATURE 140*F l ... . . .

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Because human f actors are ' major contHbutors to' DHR : loss'ev'ents- (as' dish

cussed in'section.4) . and because_ estimates 'of human performance have relatively large error bounds,.we believe that quantification of the

risk

{ .from DHR system 1.ossesVduring. modes;4 , 5, and 6 is:s'ubject to'gre'at'er error than,most other reactor risks. . In view o' f the extensive effort that is L

necessary to'obtain a quantitative, assessment.of f)HR risk '.and because of. n

. the"large uncertainty associated with such assessments, the' undertaking of a

,probabilistic risk assessment;on'DHR systems is outside.the scqpe of this

' case study,

, J , . - .

Shortly af ter a draf t of this case study was issued 'for. p'eer review comment in July 1985, NSAC published their Probabilistic Risk Assessment of the Zion Nuclear Plants, NSAC-84 (Ref. 14). NSAC-84's evaluation of risk from DHR operation supports AEOD's position; e.g., in Reference 15, which forwarded EPRI's peer review comments on a draft of this case study, EPRI stated that one of the important lessons learned from the Zion DHR PRA project was that,

" Quantifying risk during shutdown is extremely difficult,and that risk' numbers in this regime are more un'ertain c than ' traditional' PRA results.. "

Furthermore, EPRI noted that:

"An equally important lesson was that shutdown risk is highly plant specific, primarily because different plants use different procedures i and system lineups during shutdown...,"

and,

.(

. "We believe.that especially for DHR, accurate bottom.line-numbers based on generic PRA analysis techniques are neither achievable nor even desirable."

2.3 Actions to Recover the Decay Heat Removal Function As noted in section 2.2, the time aveilable for recovery from a loss of the DHR system prior to uncovering the core can be as short as about an hour.

Except for LOCA initiated events (during which break isolation would have top priority), restoration of the DHR system function appears to be the preferred recov'ery method. Iri addition to restoring the DHR ~ system, many alternate methods are available for removing decay heat when the DHR system is lost. Table 2 presents some backup methods which could be and have been used to remove decay heat upon loss of the DHR system. It should be noted that not all of the methods listed in Table 2 are available at all plants.

It is also important to note that use of miscellaneous makeup water sources (e.g., plant fire protection system) requires that precautions be taken to prevent boron dilution. Other than restoring the failed DHR system, there is. no single backup method which is applicable for all loss-of-DHR events.

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Some Backup

Methods for. Decay Heat Removal.Upon Loss

..of the DHR System *

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RCS' Intact-Vessel Head On RCS Incapable of Being' Pressurized' (RCS Capable of Eeing Pressurized) (e.g., Vessel Head Off or Detensioned, Manway Cover Off, LOCA_ Path.Open and

'Unisolable,etc.) j

~

Use'of steam' generators and RWST/BWST (gravity f1'ooding if main condenser (requires. condensate' 'available) or auxiliary feedwater pump for secondary makeup)

Use'of steam generators and HPI pumps to inject from RWST/BWST atmospheric-steam dump valves (requires condensate or auxiliary feedwater pump for secondary makeup)

Nonnal charging and letdown l '

Normal charging and letdown Spent fuel' co'oling systemf(cross-ties-' Spentfuel'cooli'ngsistem if available)' (cross-ties if available)

Chemical and volume control Chemical and volume control system to inject cold water. system'to inject cold water ,

from the RWST from the RWST feed and bleed - HP1 and PORV or Pool boiling with makeup pro-1 pressurizer safety valves ** vided from miscellaneous water

. sources (e.g.,-fire hoses)***

Not a'11 methods are available at all plants. l l

- - Precautions must be taken to control HPI flow to prevent low temperature overpressurization and pressurized thermal shock. Some plants rack out power to HPI pumps while at low temperature and

, pressure (Ref. 12).

*** Use of miscellaneous water sources (e.g., fire hoses would require that precautions are taken to prevent boron dilution).

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i.0 OPERAT10f4AL' EXPERIENCE 1 ' '" - - -'

L3.1 Lesses of' Decay Heat Remov51 Systems ,

There have been many events-in which both trains of the DHR system were

. - .' unable.to perform.their required decaysheat removal functions. From 1976' through'1983, there have been ' reports of at least 130 events'in' which ^

operating DHR ' systems .f ailed.* .This represents..a .'f requency.of about:. 0.25 V per* reactor year, based.ontabout 5'JO years of commercial'US.' PWR operation:

.There were about 90 events' reported.for ,the period 1976 a 1981, and there' were about.40 more events-reported which occurred during 1982 and 1983.

Ou'r analysis and evaluation of DHR system failures were based upon two (groups of data: operating experience from 1976 through 1981 was obtained from-LERs and Nuclear Safety Analysis Center report,NSAC-52.(Ref. 3); and operating experience for the years 1982 and 1983 was,obtained fromlLERs and NRC reports. The. reader .is directed to Reference'3. for" summaries of DHR-system losses that occurred from 1976 to 1981. Appendix A of this report..

presents summaries of the DHR system losses which occurred during 1982 and 1983. Because the 1984. data base was not yet complete, the 1984 operating.

experience was not included in our statistical presentation of the catego-ries and causes of-DHR system failures. However, the 1984 events were evaluated for significance in comparison to previous events.. The most significant DHR system. failures of 1984 are summarized in. Appendix B.

We evaluated.the loss-of-DHR operating data to determine if there were any significant trends, For the 130 reported events which occurred between 1976 and 1983, Table 3 shows that 11 plants accounted for 95 events (approxi-

-h mately 8.6 events per plant). Essentially one-fifth (21%) of the 56 opera-ting PWRs covered by this study accounted for three quarters (73%).of.the loss-of-DHR events. . .

Table 3 shows that the Davis-Besse plant has reported the most losses of DHR. However, Table 3 also shows that there has been a marked improvement at that plant since 1981. During 1980, the Davis-Besse plant experienced nine loss-of-DHR events, six of which involved' inadvertent closure of the suction / isolation valves. The repeated DHR' losses at Davis-Besse during the spring 1980 outage were . reported to Congress in an Abnormal Occurrence Report (Ref. 16). In Reference 16, the NRC stated that the licensee had a

" serious deficiency.in management or procedural controls in many areas."

Subsequently, Davis-Besse management took action to impro've administrative' controls, operating and emergency procedures, and personnel training associated with plant shutdowns. In addition, the plant's technical specifications were modified to allow removal of power from the DHR suction /

isolation valves during plant shutdown (in order to preclude their inadvertent' closure). It appears that since these improvements have been made, there have been no reported losses of the DHR system at the L

We have found 130 loss-of-DHR system events which were reported by l licensees; however, there have been many other loss-of-DHR events which have not been reported to NRC. I t

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-1976 1977~ 1978 19'79 1980 1981 1982 1983' Total i.

LDavis-Besse~. ,

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Bea'ver, Valley ' 1 1 1 4 2 1 1 10

' 2 ..10

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Calve'rt. Cliffs '2' ^1' t -

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. Salem - 2 .

2 ., , ., 8, 10

.'2 9; I" ' 'Ctysta1 ' Riy. e f "3 '~ 2 '2* . .

Calvert Cliffs - 1 2 5 li 1 9'

Trojan- 1 5 ,

1 7 North Anna .1 1 2 2 2 7

' North Anna - 2 4 3 7 Salem - 1 ' l- / 3; '1 -

5 Farley - 1 2 '2 - 1 5' McGuire - 1 2 1 3 Millstone - 2 3 'l 1 3 ANO - 2 2 2 Ginna 2 2 Maine Yankee 2 2 Palisades 1 1 2 .

. Rancho Seco' 1 1 2

'2' St. Lucie - I' 1 it Sequoyah .>1- 1 1 2 Turkey. Point - 3 2 2

, Turkey Point.- 4 2 2

. indian Point - 3 1 1 Fort Calhoun 1 1 San Onofre - 1. 1 1 Oconee - 1 1 1 1

Oconee - 2 1 Zion - 1 1 1 Surry'- 1 1 1 Sequoyah - 2 1 1 Farle'y - 2 1 1 McGuire - 2 1 1 l 1

Sumner - 1 1 i

130 Annual Frecuency f

. of DHR' Losses .06- .1 . 5. .. 3 .6 .5 . 35 .5 J 3,, (4 of events). . , , . ,

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' Davis-Besse' plant?'I(See'dppen. dix ~ C forddditionai detail's"6n' DHRtlessEs '-af

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Davis-Besse s losdof-DHR events were'the. stimuius for. I.E' Bulletin.80-12 l

(Ref. 17). .That. bulletin required licensees of PWR facilities to review utheir plants',cepability,.to' prevent.DHR;, loss events;.to review plant hard-ware-and ' analyze ' procedures for adequacy of safeguardirig against loss of-redundancy and diversity.of,.DHRl capability. The operating.dats does not 7 indicate -that ther'e has been;an industryJwid(improvement in' loss'-of-DHR .

experience as acresult of actions that were'taken in response to IE

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Bulle' tin 80-12. Referince 18 summ'arizes' actio's n t'aken by utilities in response to the IE bulletin. Our review of the plant-responses does not indicate that there is a statistically significant correlation between compliance with the IE bulletin, and DHR system losses. (For example, some plants which have complied with the bulletin continue-to have DHR loss events; whereas some plants.that have not yet completed actions associated with the bulletin have not. experience'd.any DHR. system losses.).

From Table 3, we also note that Salem 2 has experienced an' unusually high number of DHR losses in a single year. It.had eight losses in 1983 (six

~

during one outage and two duririg another outage). Four of those events involved inadvertent closure of the suction / isolation valves, three events involved DHR pump trip due to problems with the " safeguards equipment control" (SEC). system, and one event resulted from flooding of the service water bay (see Appendix A for details of those events). Subsequently, in 1964, Salem 2 had another loss-of-DHR event which involved inadvertent closure of a suction / isolation valve resulting from a procedural error i

. (, during testing of.the " pressurizer overpressure protection" (POP) system.* l Table 3 indicates that until 1981, the Crystal River plant had nine loss-of .

DHR events. There were between one and three' losses every year for four

. years. Since 1981, there have not been any. The DHR losses at this plant seem to have stopped at about the same time that the plant implemented actions to improve their planning, coordination and management of outage and maintenance activities.

One measure of significance of loss-of-DHR events is the time interval that the DHR function was lost. Table 4 presents a summary of the duration of the loss-of-DHR events which occurred during 1982 and 1983. It also summarizes the duration of ten significant DHR losses which occurred during 1984.

In 1982, there were 18 events, 14 of which lasted from about two minutes to i about an hour. The duration of the other four events is unknown.

Deficiencies associated with plant management, administrative controls.

maintenance and test activities at the Salem station were addressed by !

the NRC subsequent to the 1983 Salem anticipated transient without scram events. We believe that the licensee has taken corrective action in these areas, and a general reduction in the f requency of loss-of-DHR

'h events is anticipated as a result.

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Table 4-Duration of Reported'DHR I.oss Events . ,

Duration 19EE 1983 Ten Selected 1984 Events 0 4 minutes 4 4 0 5 - 9 minutes 3 4 1 1

10 - 19 minutes 0 3 2

20 - 29 minutes 1 1 30 - 39 minutes 2 1 0 4

40 - 49 minutes 2 1 0 0 50 - 59 minutes 1 1 ;

60 - 69 minutes 1 0 1 70 - 79 minutes 0 1 0 0 0 80 - 89 minutes 0 0 0 90 - 99 minutes 0 0 0 100 - 109 minutes 0 0 0 110 - 119 minutes 0 120 - 129 minutes 0 0 1 242 426 Total Duration in 328 minutes (without (4 unknowns) (13 unknowns) unknowns)

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' ~- ' ' 51n'1983,' there sere 28 events.,c.15.of)whichracco'un'teifor' DHRNois'es"tangin~giu '

from under a' minute to 77 minutes. The durat. ions of the other 13 events are ~

(. -unknown.-- ,

Because the 1984 data base was not complete at the time of _ t'his regio,rt, we the 1984 data! .Oure, initial: spreening indicated were unable.to review allaof,'mber of-lossfof-DHR events du' ring 1984,'.tsn' ofc

, that 'although'there were nu those events we're deemed to be significant. ..(Appendix B has descriptions.of

. those' ter) events. )' Those. ten-events, ranged'in/ duration from seven minutes' to two hours:. .Because'of recent changes in reporting requirements'(new LER rule 10 CFR SD 73 effective January 1,1984), it is not possible to make a direct comparison of industry perf ormance by examining the duration and frequency of recent years' loss-of-DHR events. The new reporting require-

. ments are more stringent than those of previous years. Virtually all loss-of-DHR events and their durations are now required to be reported.

Variations in plant technical specifications and licensee interpretation of

. previous reporting requirements have resulted .in many loss-of-DHR events which were not reported, as well as' reports which were incomplete and did not include inforrriation about the durations of_ the' events, for example, NSAC-64 (Ref. 14), illustrates the discrepancy between the number of DHR losses that occurred at the. Zion nuclear station and that which was reported to the NRC. By reviewing control room logs and maintenance records, the authors'of NSAC-84'found that there were 21 loss-of-DHR events at Zion 1 and 2 which the licensee deemed to be unreportable (16 events in which there was autoclosure of the suction / isolation valves while the DHR system was operating, and'five DHR losses which were attributed to inadequate reactor vessel level measurement.during draindown operations)'. In contrast, the licensee only

( reported one loss-of-DHR event through 1983.

In an effort to reduce the. frequency of inadvertent suction / isolation valve closures, Zion plant procedures (as reported in Reference 14) call for deenergi2ing the valves in the open position before conducting setpoint

' calibration and prior to conducting work on the inverters. This has been effective in re.ducing the Zion station's valve closure frequency from one per outage.to zero.

Examination of the data presented in Tables 3 and 4, and in Appendices A and B. indicates that some plants have been having a disproportionate number of long duration DHR losses in recent years; i.e. , . North Anna 1 and 2 have had ,

five long duration loss-of-DHR events in 1982, four long duration loss-of-DHR events in 1983, and in 1984 North Anna 2 had a two-hour loss.

McGuire 2 had three long duration DHR loss events during a three-week period between December 1983 and January 1984 (43 minutes, 62 minutes and 49 minutes). McGuire 1 had three loss-of-DHR events in 1982 and 1983, the longest of which lasted one hour.

The operating experience shows that North Anna and McGuire have experienced i mul-tiple and long duration loss-of-DHR events without apparent improvement.

The Calvert Cliffs plants have experienced multiple loss-of-DHR events

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of in.cr.e.as,ed' mana..g'e. ment attention. '-

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The fact that the number of DHR lo.sses as,well,as the duration of. losses in 1984, continue, to:be.high (>15 events:);tends 'to' fndicate that plant performance is not improving during ashutdow'ris. .However, it 'is 'iinportant to note that' '

. .dur.ing the years- 1982,.1983.and 1984,.the longe.st duration event was two hoursi whereas' there.were five events' during .the yeatM976 4: 1981 which

~were of longer duration. Although many of the 130 loss-of-DHR events' exceeded an hour, plant personnel have always been able to restore the DHR function prior to reaching an unsafe condition (core uncovery). Since none of the long' duration events' occurred immediately after shutdown, there were no ,

serious consequences. However, under slightly different circumstances, some of the loss-of-DHR events could have led to serious consequences.- Figure 4 shows how recovery time varies as a function of the. time after. scram that the loss-of-DHR'eventro ccurs.. i % - -- ' -

While none of the recorded DHR failures a'ffectedi the' health'and safft'y of the public, some of the events caused significant' plan't disruptions, extended downtime, and expensive cleanup and recovery. The Oc6 nee 2 DHR system loss, which occurred on September 18, 1981 (Refs. 19 and 20) is a good example of such an event.*

Our analysis of operating data included categorization of 130 DHR system failures that occurred at PWRs during the years 1976 - 1983. Table 5 presents the results and shows that events involving the" suction / isolation valves and the DHR pumps accounted for about two-thirds of the DHR system i failures.

More than one quarter of all reported DHR system losses which occurred between 1976 and 19E3 involved automatic closure of DHR suction / isolation valves (37 events)** The underlying or root causes of most of the automatic valve closure events were human factors (see section 4.1 for further

-discussion). Only two of the. automatic isolation valve closure events were On September 18, 1981, while at 94% power, Oconee 2 developed a steam generator tube leak (25-30 gpm). A rapid plant cooldown and depres-surization was begun. The plant cooldown and depressurization were delayed when it was found that one of the DHR suction valves was stuck closed, thereby making the DHR system unavailable. As a result, plant cooldown was delayed for more than 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months . The primary to secondary leak resulted in an accumulation of about 2h million gallons of con-taminated water in the turbine building. It took about 60 days to reprocess the contaminated water, and to clean up the secondary system l and the turbine building. i

- This does not include the 16 events that occurred at Zion Nuclear Station which were not reported to the NRC. i s

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Table 5

.Cateoories'of'130 Reported'Totel DHR' System' Failures When Reovired to' Operate -(Loss of Function).

at U.S. PWRs 1976-1983 No. of Events _ (* of Events) i Automation Closure of Suction / 37 (28.5)

Isolation Valves Loss of Inventory Inadecuate RCS Inventory Resulting 26 (20.0) !

in Less of DHR Pump Suction 36 (27.7)

Loss of RCS Inventory Through DHR 10 (7.7)

System Necessitating Shutdown

(' of DHP System Component Failures Shutdowr or Failure of DHE Pump 21 (16.2) 29 (22.3)

Inability to Open Suction /Iscletion 8 (6.1) ,

Valve .

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Others 28 (21.5) l l

Total 130 (100.0)

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jegitimate. respon'ses.t.'o"vA1.'id' signals 1 which" correct 19' detected' ah RC5 4,.,%. ,,': -

pressure exceeding the isolatiori setpoint, i.e. a low temperature overp'ressure event. '

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More than.one-quarter of all reported DHR system losses which occurred between 19,7,6 and 1983 involved loss of RCS.inve'ntory (36.' events).

Twe'nty'-six of the" loss of inventory events resulted in inadequate p' ump -

Many events,'of this type.were.

.. suction, significant,cavitati.on because ofor theirair..lortgbinding.

.re ,dovery time's. ..Reco'very req'uired refilling the RCS and' bleeding off the air or vapor bound pump (s).

Appendix B indicates that in 1984 there were at least six such events which lasted between 25 minutes and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

About one-fif th of the reported DHR losses which occurred between 1976 and 1983 involved DHR system component failure. Twenty-one events involved shutdown or random failure of an operating DHR pump when the other pump or train was inoperable. . 'Eight events involved previously closed . suction / iso-lation valves that could not be opened. ,

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t a l' 4.0 'ANALYS15'AND EVALUATION'0F THE'UNDERLYJNG OR ROOT CAUSE5 0F.IEPORTED

OHR SYSTEM LOSSES ,.

Table 5 presents a listing of the categories'of DHR system' failures. In addition to categorizing the events, our analysis examined the licensee submittals to. determine the underlying or root causes of the events.

Table 6 presents the results of our assessment.

It can readily be seen from Table 6 that the dominant underlying or root causes of DHR system failures are human factors (procedural inadequacies, operator or technician errors, etc.). Human factors account for almost two-tnird; of trie events. Equipment failures were the second major underly-ing or root cause and accounted for aboit one-quarter of the events.

4.2 Human factors The operating data revealed that human factors are the dominant underlying or root causes of almost two-thirds of the DHR system losses.

The major human factors problems that were common to many DHR system losses include: 1 PROCEDURES A. During normal outage activities (maintenance / repair / test /

surveillance), procedures often; f o omit caution statements regarding restoring equipment on completion of tasks, o fail to consider interactions with other tasks or equipment, e are poorly written and cmit steps or contain ambiguous instructions, o fail to identify equipment the same'way it is labeled or referred to by the operator.

I B. During f ailure identification and recnsery activities, procedures l frequently:

o are not available or are no'. applicable for a loss-of-DHR )

event, j o are incomplete or lack specificity, o refer to or depend upor, indicators, instruments, alarms or annunciators that have been optimized for power operation but are inadequate for shutdown operations and/or are improperly ,

I placed for shutdown operations, o do not provide operators with information about times i available for safe recovery, and how to track the course of

. the event.

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' Table 6 Underlying or Root Causes of. Reporte-d DHR System failures No. of Events (% of Events)

Hurr.en f actors inadequate / faulty procedures 49) (37.7))

l Operator / technician errors 23 )83 ( 17.7)'t(63.9) l Inadequate /feulty prccedures 11 (8.5) combined with operator / J technician errors Ecuipment Failures Pumps, valves, re1ays, c'r. 37 ( 2'8. 5)

Unkne., Causes 8 (6.2)

Human Factors Combined 1 (0.7) with Eauipment failures 1

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Others 1 (0.7) >

Total 130 (100.0) i j

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OPER TOR AIDS '(InstYu'medtation? Nohitbri'nh'Endip' ment'Afa'r's,'

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(Irr performing normal' outage ac'tivities and identifying / mitigating'

. failures, man / machine' interfaces are inadequate for ensuring task

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prof..iciency. '.0pe.rator 'aidt, are usually, designed for powe_r .operatio.ns

'and are often ihadequate'for use durin'g shutdown. In general, operator aids _ 'during shutdo'wn:

o' are not available to monitor or track operations or events, o- may be pocrly placed relative to the task being performed,

'o ~ are not integrated into operator tasks (e.g., infrequently monitored levels, temperatures, etc.).

f PERSONNEL ERRORS A. During normal outage activities, erros., of omission and/or

. commission have been caused or affected by:

o misunderstanding of procedures, instructions, and tasks, o unfarniliarity with equipment or tasks, o lack of. understanding of importance of tasks and the interfaces with other ongoing tasks or activities, l u

o accidents (bumping or dropping equipment),

b, o inadequate training.

B .' During f ailure identificati_on and recovery activities, recovery times have been adversely impacted by:

o . operator unfamiliarity with instrumentation used for diagnosis and/or recovery techniques, o operator unfamiliarity with other ongoing activities, o inadequate operator training.

PLANNING For both normal outage activities and event response activities:

o emphasis seems to be on minimizing outage time and meeting technical specification requirements (limiting conditions for operation, etc.), not on equipment or system interaction, o interactions between simultaneous activities may not be j factored into the task assignments or procedures (e.g., ~

jumpering, blocking of circuits, and taking equipment out of service may not be accounted for).

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1 Altho' ugh' the idss-of'-DHR. everit's'as'so'clated with inabequa'te/RCSt inWntory a " {

l usually involved f ailures 'or inadequac'ies ,of' equipment' a,sociatedi with j

I Ll.iquid le. vel measurement,.'we viewed their ss having'been caused by . human

- ifact' ors Since most of:these events repre=ented breakdowns'of the man /m'achine. %

interface. These events typically'resulted from inadequate and/or im-1 properly placed instrumentation,' annunciators, alarms; ina.dequate monitoring '

procedures; inadequate"tPaining 'associatediwith _le' vel' measurement 'sy'st'em'.

j op'eration, or, operator error. ,

Some of. the most sig'nificant even s involving human'f'act' ors occurred whil'e

.the primary coolant-system was partially drained and'the operators were misled on the status of coolant inventory by inaccurate liquid level instru- j

-mentation. In many cases, the level instrumentation devices were incor-rectly calibrated, or were makeshift apparatus which were prone to failure and measurement errors (e.g., tygon tube sight gages). . In most events.of.

this type, the operators did not have advance warning of an inventory problem. A. frequently observed scenario was one in which'the RCS was drained. down to the' point where there was inadequate net positive. suction.

head (NPSH) or air entrapment in the DHR' pump. Ad a result,'the DHR pump' cavitated, co'ld not deliver the design coolant flow, or even became air-bound. The first symptoms were usually increases in pump noise and changes in pump motor current which were caused by inadequate suction- head and cavitation. In many events, the operators diagnosed the problem as a pumo problem when the cause was actually an inventory problem. As a result, in many cases, the operators activated the redundant pumps only to find that they also malfunctioned. It should be noted that continued operation of a ,

'DHR pump with inadequate NPSH or a closed suction valve could result in DHR l pump failure, i Appendices A and'B contain' descriptions of 18 events which occu.rred.in 1982,.

'1983, and 1984 which involved insufficient ~ inventory and subsequent pump problems. Between 1976 and 1984 there have been ten events where the operators required about an hour or more to restore operability of the air s or vapor bound pumps. Most of the longer duration events occurred before 1982. In 1984- there were at' least six low inventory events, two of which lasted more than an hour.

With regard to the lack of information available to operators during DHR operations, we note that many plants do not have (or have improperly placed) annunciators to warn of low DHR pump NPSH or low DHR flow. For example, in Reference 21, Diablo Canyon reported a DHR pump' failure after operating for about an hour with a closed suction valve.

Regarding inadequate procedures and training for mitigation or recovery from a DHR system loss, we note that in a recent LER (Ref. 22) Zion 1 reported a 45-minute loss of the DHR system due to draining of the primary system to a

! level below the DHR pump's suction line. The LER implied that there was no procedure available for responding to this event. The licensee stated that !

"a procedure for loss of RHR will be prepared," and "a procedure for a loss )

of RHR will be written to provide guidance in the proper actions to be taken )

in the event of an indicated loss of RHR." Reference 14 notes that there l were at least five previous loss-of-DHR events at Zion caused by inadequate )

RCS inventory. However, none of those five previous events prompted the licensee to prepare procedures for recovering from such an event.

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Subsequent'discutsi6ns with' ope'rator's"at'several ilant rthat' have" ' V (.E .

experienced significant losses'of DHR have indicated that plant personnel do

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I. not haveracequate information about the i.ime margins ava'ilafil,e for recovery

'. fromLloss-of-DHR events' prior to reaching bulk boiling, core unco'very or some other safety-related threshold (i.e. , tables'or graphs showing time to

. reach bulk boiling or core uncovery as a function of ti.me af ter reactor trip,Jsuch as.that sh6wn in Figure 4).

F A' poignant example',of many ofT the aforementioned human factors concerns'

. (e.g , procedures' a'id man-machine interf ace)'is documented in a recent NRC inspection' report (Ref. 23), from which excerpts are reprinted below.

Licensee Monitoring of Core Cooli.ng Parameters During Mdde 6 j Operation

'The NRC resident inspector expressed concern over the adequacy of ,

surveillance practices provided by the licensee for monitoring I proper core cooling during extended periods in,which-the Unit 2 reactor vessel was partially drained with the closure head detensioned;

. .the reactor vessel was drained to the middle of the hot leg nozzle, which provides approximately six feet of water above the top of the' core.

... Depending on the prior core power history and the length of time between reactor shutdown and head detensioning, core decay f

'\. heat could be sufficient to boil the available water cover Ethin a very short period of fime (several hours) if proper 6 core cooling were not maintaineJ. This vulnerable state 3f plant condition forms the basis for NRC concern.

Considering the plant conditions noted above, the following points were noted with respect to licensee monitoring of important core

cooling parameters:

-(1) There is no reactor vessel water level indication or alarm in the control room. The only current requirement for monitoring vessel water level is to monitor a temporary standpipe installed in containment on a once per-thif t basis.

This could involve up to 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months between water level observations.

(2) There is no core cooling flow alarm in the control room.

Core cooling flow indication is available in the control room; however, it is only required to be monitored once a shift.

(3) There is no direct reactor vessel water temperature indication or alarm in the control room. Available reactor coolant temperature instruments are located in stagnant l portions of the coolant loop and not in the shutdown l

/ cooling flow path. f 1

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~;' 'A~' u,I!Ba' sed in~the71imited availability b1 Linstruhentat' ton of ' " '

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. ' alarms'in theicontrol room anf the infrequently required .

f monitoring of key core cooling parameters by operations personnel,-

4 theT NRC~ resident requested the' licensee to evaluate these concerns and identify what action is consideped warranted.

k "Thelicensee' ackn'ow1' edged'the:NRC concerns and 'took action fo increase the monitoring of core cooling flow to every.,two hou.rs

. .unti.11 therrefueling 'ca'vity was flooded. The licensee aareed to further evaluate the specific NRC concerns from the standpoint of the adequacy of control room instrumentation and surveillance

. ' procedures.

(Emphasis Aoded) 4.2 Equipment Failures Our analysis and eva)uation of the operating data revealed that failure of equipment, such asipumps,ivalves, relays,.etc. , were. the' underlying or root..

causes'of more than one quarter of the reported DHR system failures.(36 of ,

l 130).

The data showed that almost all of these 36 events involved random single failures which occurred while the plants were in modes 5 or 6. The redundant DHR trains were frequently unavailable due to testing,' maintenance or repairs as allowed by plant technical specifications. We believe that there were very few DHR loss events during mode 4 because only a small portion of. shutdown time is spent in' mode 4, and because many of the test, maintenance and repair activities associated with plant shutdown are not normally initiated until the plants are in modes 5 or 6.

.We note that. in a generic letter (Ref. 24), NRR required all operating PWRs .

to modify their plant technical specifications to " provide for redundancy in decay heat removal capability,in all modes of operation." To date, all but two plants have modified their technical specifications to meet the requirements of the generic letter. (Those two plants are Palisades, and TMI-1.) Although, the technical specifications required by the generic

~

letter decreased the likelihood of DHR loss due to single failures, they did not fully assure DHR redundancy during all DHR modes. Implementing the technical specifications that were included with the generic letter do not prevent a licensee from disabling a train of DHR.during times of'high DHR heat load.

For example:

o The generic letter permits the plants to have only one train of DHR operable during periods of higher risk, i.e., high decay heat load during mode 4 and the early stages of mode 5.

10 Licensees were not required to formulate detailed emergency

. procedures for loss-of-DHR and train their staff in their use.

o Standard plant technical specifications allow disablement of all DHR and RCS pumps for up to one hour if the RCS fluid exiting the core is at least 10 F subcooled. However under those conditions, it is

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We' note that NSAC's study'of DHR operating experience.(R'ef 3) s'uggests that

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.both DHR trains should be: operable during.the time of high decay heat (the ~

few hours.that-the plants are'in.^ hot shutdown mode and the. fir,st few. days of

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. cold shutdown) 'We' endorse NSAC's co'ncern for having' such" redundant l)HR.

cepability-available during times of high. decay heat. ,

i 4.3 Technical Specification Deficiencies Our review.of loss-of-DHTt events identified the potential for' increased risk due to inadequate technical specifications, specifically concerning the necessary. conditions required for the determination of " cold shutdown," and v .the absence of'a requirement for. vessel level monitoring equipment.

4.3.1 . Mode . Definition /Early Disabling of Equipment -

An early loss of the DHR system could effectively place the plant in a degraded mode while the plant has'a higher decay heat ger.eration rate, resulting in~ a shorter time available for a safe recovery. For example, the;following scenario is currently possible:*

o During pla'nt' shutdown, mode determination is based upon " aver, age coolant temperature."' However, " average coolant temperature" is not defined in the' standard technical specifications.** Thus, by selecting an inappropriate temperature:to determine " average coolant temper 6ture" f an-inaccurate and premature mode determination could result.

o' Once a plant is declared to be 'in "co.ld. shutdown". (mode 5) plant-personnel may. disable redundant equipment and initiate maintenance, surveillance'or repair activities and'the DHR system may consist of

.only.one operable train.

o _The plant becomes highly vulnerable to losing the DHR system due 1 to.a single component failure.

As noted in Tacle 1, standard technical specification definitions of operational modes depend upon " average coolant temperature." However,

" average coolant temperature" is not defined in the standard plant technical

We are unable to ascertain if this scenario has actually occurred thus far. However, review of operating data leads us to believe that the likelihood of occurrence is high. }

1 No universal convention defines T,yg during modes 4, 5 and 6.

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B&W plants, it was. learned that the B&W plants are not consistent in'the methods they use to determine when cold shutdown conditions.have.been.

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achievel Most of'the plants depend uoon only one temperature measurement' to make a. mode determination. Some of the temperature readings that the B&W

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Y licensees use to determi.ne .when cold shutdown' is achieved are:*

. lo- Cold 1eg temperature .

oD Hot leg' temperature e

o DHR pump outlet temperature o DHR pump suction line temperature o DHR return line temperature However, we note that a reading of only one of these temperatures does not provide a valid indication of " average coolant temperature." Operating experien e has shown that for B&W plants on DHR cooling, the." average primary coolant temperature" can be. much higher than sany temperatures'being used for~ mode determination (Refs. 25 through 29). 'The hot leg tempera-tures,** especially coolant near the top 'of, the " candy canes'," in the reactor upper head region, in the pressurizer, and in the pressurizer surge '

line, may be at substantially hotter temperatures than the temperatures

-being used for mode determination. As a result of a premature designation ,

of cold shutdown, it is quite likely that plant personnel may disable safety l systems and defeat DHR system redundancy in' order to initiate test, maintenance and repair operations which are allowed during cold shutdown, prior to actually achieving cold shutdown conditions. In essence, safety-related equipment which may be required during hot shutdown conditions may be bypassed or disabled prior to actually establishing the conditions required for their disablement.

We.no'te that if some of the loss-of-DHR events had occurred while the primary coolant system temperature was higher, and the decay heat generation i rate was higher, less time would have been available for re:overy and core ,

uncovery could have been reached. From the operating experience and our i calculations, it appears that the risks associated with the loss-of-DHR can be significantly increased by the premature declaration of cold shutdown and concomitant test, maintenance, and repair operations.***.

t

We have not canvassed resident inspectors at other PWRs; however, CE and W standard technical specifications define operational modes in a simiIar manner, and do not define " average coolant temperature."

- Except for the area of hot leg U bends (J 1egs or candy canes) which is unique to B&W plants, most of this discussion is also germane to PWRs

. designed by CE and W.

- - On? of the peer reviewer's comments on the draft case study recommended thAt a grace period be used to determine when DHR system redundancy can be defeated. Specification of such a grace period would depend upon l

DHR recovery time margin such as that which appears in Figure 4.

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4;3.2c Absence 'of Requirements. for *CS* level tleasuremerit'Duri;ng"ShiatdosM,"

In addition to the 26 0.HR lossps which.resulted from inadeq'uate RCS'inven-

'[ tory through 1983, our, review of operating data'has shown that there were at

- least sii additional events in 1984. Those loss-of-DHR events involved

.incorrec_t or faulty lev.el measurement of.a drained RCS. Some of the loss-

~

' f-OHR events were of'long duration '(40 miniites t'o'two hours)?beca'use of the o

time required to restore tt e ! air'or; vapor bound DHR pumps.. 0ur' review.of .

plant specific techn'ical specifications re'vealed that 'there are'rio technical

, specification requirements addressing RC5. level measurement (equipment or i

procedures) during shutdown. We consider the absence of such' req'uire nents, L especially during RCS draindown, a significant technical specification

. def i ci ency.-

l 4.3.3 Omissions Regarding Equipment Operability As.part of.this case study, we reviewed the technical specifications of a number of plants. That review-revealed that the technical specifications at Oconee 1. 2 and 3 are incomplete with regard to shutdown modes and DHR system operability requirements; i.e.,

the technical. specifications for the Oconee plants do not define operation with RCS average temperature between 200 F and 525*F, and do not address DHR system

operability requirements during all shutdown modes.

H ,

Our review of other plants' technical specifications did not identify omis-sions which are similar to those of the Oconee technical specifications.

(

'2 However., since we did not review the technical specifications for all plants, we are not sure that there are no other operating plants which have.similar omissions.

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The Oconee plants' technical specifications refer to the DHR system as the Low Pressure Injection (LPI) system.

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, J "' ' ' 5'.0' flNDINds AND"CONCLU5 IONS' U.S. PWR. experience has shown that dur.ing abo'ut 500 reactor years, 130'

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losses of operating DHR systems were reported (0.25 event per reactor year). N' Some of those events lasted for several hours.

The' operational- data clearly ' indicate that hunia'n ' factors were 't'h e rootJ -

causes.,of most of the loss-of-DHR events.that have been reported to have- ,

+

occurred at U.S. PWRs through 1983." Inadequate procedures.and operator /-

technician errors during testing, surveillance, maintenance, and repair operations were the root causes of almost two-thirds of those loss-of-DHR events.

As noted in section 2.2, review of operational data, licensee submittals, and scoping calculations all indicate that primary system boiloff and core uncovery,can occur during certain events within a few hours after loss of the DHR system. .The situation can be especially acute if.the RC5'is par a tially drained, if the event is initiated by a LOCA shortly after the DHR system is activated, or if the loss occurs during~the first several days of plant shutdown. Fortunately, the plants have recovered from the loss-of-DHR events that have occurred thus far, before sustaining serious consequences.

In addition, under certain conditions,* primary system pressurization could occur at all types of U.S. PWR plants within 30 to 60 minutes after losing the DHR system. Such pressurization could challenge the low-temperature overpressure (LTOP) protection equipment. An extended failure to restore the DHR function could result in a small braak LOCA, with primary-system boiloff at the LTOP relief valve setpoint pressure. Continued boiloff could lead to core uncovery in as few as two to three hours. 'l

- The results of NSAC's probabilistic risk assessment (PRA) of loss-of-DHR events at Zion (Ref. 14) and Reference 15 support our position that a risk assessment of DHR system losses is of little real benefit because of the large uncertainty associated with the quantification of human factors events 4 (human factors being the dominant cause of the DHR system loss events), the l' difficulty in obtaining accurate failure data, and the differences in plant designs. We view the many loss-of-DHR events which have occurred at PWRs thus far (one loss-of-DHR event every four (PWR) reactor years, which t equates to more than a dozen such events each calendar year) as a signifi- !

cant group of precursors. We conclude that corrective actions are required I to minimize the risk associated with DHR losses. )

l As noted in section 3.1, we have been unable to detect a significant '

industry-wide improvement in DHR loss' experiences. If licensees wEre to incorporate the lessons learned from previous loss-of-DHR experiences, especially the recommendations provided in NSAC-52, we would expect to see a significant improvement in their DHR loss experience. However, the absence of such an industry-wide improvement, and the continued occurrence of DHR l loss events, led us to conclude that many licensees may not be incorporating j NSAC 52's recommendations. ,

With the RCS intact and the reactor vessel head on.

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5.1 Human Factors Considerations .

From our analysis and evaluation of operational data', we.~ontlude c that many plants do not pay adequate attention to the human factors aspects of plant operations, testing, surveillance, and maintenance during plant shutdowns.

As shown in' Table 4, and 'as illustrated by the data appearing in Appendi-ces A, B, and C, faulty, procedures and operator / technician errors associated with plant shutdown operations were the underlying or. root causes of almost two-thirds'(83'of 130) of the repor*.ed loss-of-DHR system events.

Based on operating data and discussions with plant personnel and reactor inspectors, we conclude that the techniques used for planning and coordina-tion very widely from plant to plant and are frequently inadequate to prevent the occurrence of many loss of-DHR events. (Loss-of-DHR events have l

frequently resulted from conflicting or interacting outage activities.)

Most plants have. outage planning groups which look atc. outage scheduling from the standpoint of schedule and hardware availability. However, equipment l-.

and system interactions associated with ongoing test, surveillance, and maintenance activities do not necessarily receive adequate planning or attention unless there is a particular technical specification requirement associated with it. Improved outage planning which focuses on the timing of conflicting or interacting activities could significantly decrease the frequency of loss-of-DHR events.

With regard to the man / machine interface associated with DHR system operation and malfunctions, we found that for many plants:

(. o Existing procedures and equipment associated with RCS level monitoring during plant shutdowns are frequently inadequate and are failure prone. Inadvertent and undetected reduction of RCS inventory is a potentially significant contributor to risk associated with loss-of-DHR when the RCS is partially drained (26 events through 1983, and at least.six more events in 1984). We conclude that more reliable instrumentation and procedures should be used to reduce the frequency and, thus, the risk due to inventory problems leading to loss-of-DHR events.

o Dperator aids are not readily available to assist in the detection of abnormal plant behavior while the plant is in modes 4, 5 and 6.

Instrument alarms and annunciators are not conveniently located to enable the operators to integrate them into normal and emergency procedures during shutdown periods. In addition, operator aids are not available to enable operators to trend RCS and DHR system parameters during loss-of-DHR events (e.g., temperatures, pressures, flows, etc.).

We were informed by a reactor operator that during a recent DHR loss event, he had to rely upon his stopwatch and graph paper to determine

. how much margin was available prior to bulk boiling in the reactor.

Time margin information such as that depicted in Figure 4 is not generally available to operators to assist them in recovering from loss-of-DHR events (plot or table of time after DHR loss until bulk boiling or uncovery begins as a function of time after rod insertion at which DHR loss occurs).

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.,5ystem.heatup or pressurizat. ion,.etc. --

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o . Based upon the co.rrective actions taken after loss-of-DMR events,

'.' . ' ve. conclude that p.lant personnel', especi. ally no'n-licensed operations 1 .- and maintenance staff, are not' sensitized'or fully aware of the. risks l associated with their activities during plant shutdownc The risks

l. during times of high decay he'at rate, drain and fill operations, and I

during operations in which redundant equipment is disabled do not appear to be fully appreciated by all plant personnel.

1:

5.2 Desion Considerations - Flow Path from the Reactor' Coolant System i to the Decay Heat Removal System . . . .

1 5.2.1 Double Drop Line Configuration l From our evaluation of the operating data, we conc 1'de u that. adding a second l crop line to provide a redundant DHR suction flow path will not result in a l significant improvement in DHR system reliability and availability.

Furthermore, the double drop line configuration may result in an overall l increase in risk due to the increase in the probability of Event V. As an l alternate to tr.e double drop line configuration, a suction bypass line (as discussed in section 2.1) may provide a less expensive, and possibly safer (when considering Event V) method for improving DHR availability. We con-cluded that the use of a DHR suction bypass line would have contributed significantly to mitigating the September 1981 Oconee 2/ event ~which resulted in significant onsite contamination and an extensive' outage (see section 3.1). The suction bypass line would have introduced an alternate j DHR flow path enabling a more rapid cooldown, thereby reducing the amount of '

leakage, contamination and down-time.

5.2.2 Inadvertent Closure of DHR System Suction / Isolation Valves As noted in section 2.2, closure of DHR system suction / isolation valves shortly af ter initiation of the DHR system could result in an LTOP cha11enga to the RCS at PWRs within 30 minutes,* with core uncovery occurring as j early as about two to three hours after valve closure. '

Operating data has shown that for DHR system operation, removal of power or removal of the autoclosure interlocks to the DHR suction / isolation valves can be a safe, effective method for preventing spurious suction / isolation I

Or within only a few minutes during " solid" operations as described in Reference 13.

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'DHR system is provided by the DHR system relief valve. Since all plants do not'have adequate relief through the DHR system, additional relief capacity

( may be nscessary prior to removing power.or the autoclosure interlocks to the suction / isolation valves.'

This case study. report hasistimulated 'aiu'c h interest.in the subject of auto-closure interlocks.~ Based upon an. earlier (1984) draft of this case study report, Sandia Laboratories' performed a risk assessment'as part.of' Task A-45 evaluating the competing risks associated with DHR suction / isolation valve closures and Event V. Their report (Ref. 30), '? Potential Benefits Obtained by Requiring Safety-Grade Cold Shutdown Systems," was done for the Calvert Cliffs plants' configuration. Subsequent to their quantification of risks, Sandia concluded that:

"The lowest core melt frequency due to the combination of loss of RHR suction during cold, shutdown end V-LOCAs.js obtained when there are no autoclosure interlocks on the RHR suction valves... removing the overpressure interlocks from the RHR suction' valves give's the best RHR suction arrangement for PWRs based upon this analysis.

...when interlocks are present, loss of RHR suction is the largest contributor to core melt frequency for all assumed values of P(CM-LRHRs).** however, when the interlocks are not present, the core melt frequency due to loss of RHR suction is comparable to or less than

- the V-LOCA core melt f requency for the "best estimate" cases.

f Finally, we believe that the "best" RHR suction valve arrangement is to

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have a single suction line without primary system over pressure interlocks on the valves."

l l In. response to the earlier draf t of this case study, NRR reviewed the issue t of "RCS/RnR Suction Line Interlocks on PWRs." NRR performed a prioritiza-

-tion evaluation (a simplified risk and cost assessment). As a result, on August 13, 1985, in Reference 31, the Director of NRR forwarded a copy of his staff's prioritization of this issue, assigned it a'"HIGH" priority ranking, and directed the Director of the Division of Systems Integration to take the actions necessary to resolve this issue. I It is also important to note that Westinghouse has evaluated Kewaunee's proposal for removing the autoclosure interlocks on the DHR suction valves.

Reference 6 notes that Westinghouse's analysis concluded that for Kewaunee, the proposed modification would be a safety improvement. NRR has subsequently approved the modification. As noted in Reference 6, the effects of e"toclosure interlock removal upon plant safety must be The Davis-Besse and Zion plants use this approach, and the Kewaunee plant has recently received NRC approval to remove its autoclosure interlocks.

l P(CM-LRHRs) = probability of core melt given that RHR suction is lost.

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5.3 TFchnical Specification Deficiencies * '

5.3.1 . Mode Definition /Early Disabli.ng of Equipment' We found that most plants' tectinical specifications are imprecise t

'with .

.rega'rd to' the1 designation of plant operating inodes becliuse the aver'age .

coolant temperature's are undefined. 'As a result, a premature' designation of cold shutdown is possible,_and thus equipment can be disabled.or bypassed and DHR-redundancy eliminated during conditions of high decay heat load. As noted in section 4.3, it is possible to enter a condition in which equipment may be bypassed prior to properly establishing the conditions required for the bypass. Thus, the plant is more vulnerable to loss-of-DHR due to a single failure, and with the higher decay heat, improper mode definition can reduce;the time available to prevent core u.ncovery,

~

We ' conclude that regulatory action should be taken to assure the' proper definition of shutdown . mode, and assure Dt:R redundancy during periods of high decay heat load. The use of a grace period ** based upon available DHR i

recovery time may be a viable alternative to th_e present practices.

5.3.2 Absence of Requirements for RCS Level Measurement During Shutdown Our review of technical specifications concluded that the lack of requirements for RCS level measurement and monitoring during shutdown and I

draindown is a significant generic safety deficiency. Considering that:

l l o there have been a significant number of long duration DHR losses involving inadequate RCS level in re' cent years (including six in 1984),

and o the times available for recovery prior to reaching unsafe conditions l are relatively short, we have concluded that regulatory action should be initiated to ensure reliable RCS level measurement.

These deficiencies may be viewed by some as human factors type deficiencies because of their impact upon plant operating procedures, etc.

- Prohibit entering mode S until certain time, heatup rate, or recovery time criteria are met, l

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.As notek/in,section 4.3.3,. our review found three plants' echnical speci-fications to be incomplete with regard to. shutdown modes'and DHR system operability requirements. We are not,sure that,there are'no,other plants

. with similar omissions'. Furthermore, the plantf having'those< deficiencies r- have been determined to meet the requirements of NRR's 1980 generic DHR'

" letter-(Ref. 24). .'We':sre un'a'ble to ascertain why. the defici'e nt technical specification's have not.yet been modified,'and we question if'there are' other plants that have been determined to meet the requirements of the generic letter, but that also have similar deficiencies.

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6.0 RECOMME'NDhTIONS~ 2

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('1) AEOD recommends'that NRR assess the need for~NRC requirements to improve planning, coordination; procedures, and personnel training during shutdown to ensure the availability of DHR.

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Webelieve.that'significant;1 improvements;i$tDHRs:ys't'em-availabilityand reliabi_lity can be achieved by focusing upon human factors aspects of plant shutd6wn.' LWe recognize' the. fact"that NRR.i tinit'ati.ng i a ' generic ma'inten '

ante and surveillance program'to look into some of thesE issues (Ref. 32).

We recommend that, as part'of that effurt, NRR'should review indur.try practice and determir.: if guidelines or specific requirements are necessary to ensure plant safety during DHR system operation. Emphasis should be placed upon detailed planning of test, surveillance and maintenance activi-ties, and the equipment or system interactions which have frequently caused loss-of-DHR systems.

In addition, plant practices regarding the procedures and training of personnel for performance of normal (non-emergency)' operations during shutdown should be evaluated. For example: all operations and maintenance staff (licensed and non-licensed) should receive training to assure that they become senritized to the risks associated with plant shutdown. ~Empha-sis should be placed upon understanding the risks and high vulnerability associated with times of high decay heat rate, drain and fill operations, disabling redundant safety equipment, etc.

(2) AEOD recommends that NRR require PWR licensees to have a reliable method of measuring and monitoring reactor vessel level during shutdown }

modes of operation and corresponding technical specification requirements for operability.

Common industry practice using unanalyzed makeshift devices such as failure prone tygon tJbe sight gages to monitor RCS !evel during plant shutdown should be modified or discontinued. We recommend that NRR require the licensees to use reliable, RCS level monitoring instruments during modes 4, 5, and 6. Consideration should be given to requiring redundant level indication during modes 4, 5, and 6 to ensure availability of trending data, and to warn operators in advance of unacceptably low RCS level. In addition, plant procedures should be modified to assure that the frequency of RCS level monitoring is commensurate with plant status (e.g., as noted in section 4.1, one plant could have monitored vessel level as infrequently as once every 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , whereas fuel uncovery could occur only a few hours after a loss-of-DHR). As a minimum, each plant's safety review committee should review the instrumentation and procedures used for RCS level measurement during modes 4, 5 and 6 to ensure that b high level of reliability is achieved.

(,3 ) AEOD recommends that NRR require the licensees to improve the man-machine interfaces related to DHR operation.

We recognize that all DHR losses cannot be totally eliminated by good planning, goed procedures, well-trair.ed personnel, etc. We believe that if all licensees would perform human factors analyses of their plants' DHR operations, (including normal and abnormal conditions) and modify their l

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. O w plan' epractitus'and t man /stachine3nterfaces:according1pi.the risk's7frosF9HE losses ~would be significantly reduced. A model'to use for,such human

( factors analysesL is one use'd by NRR'(Ref. 33). Reference 33 requires

-licensees to perfotm i specific task analyses, and to integrate instruments-tion, alarms and annunciators into normal and emergency procedures for transients and. accid.ents occurring during power operation. As'a minimum, we recommend that NRR consider requ' iring licensees to perform' human-factors

, reviews as described in; Reference 33, but extend them.to shut'down opera-

' tions, with emphasis on detection an'd mit'igation of 1655-o.f-DHR events.

The operat' ors should be provided with informat' ion (sech as Figure 4).

outlining the time margins available for recovery from postulated loss-of-DHR events as a function of time from reactor trip for a representative set of DHR loss transients. Examples of such transients are: primary system filled at maximum DHR system temperature; primary system drained to minimum level and open to the atmosphere; RCS at refueling temperature, etc.

Information on time margins available would assist operators in recognizing the potential seriousness of the event, and assist them.in choosing apprentiate methods for restoration of the DHR function.

(4) AEOD does not recommend changing the design of DHR systems to include

'redunoant drop lines.

Based upon our analysis of the suction / isolation valve closure logic at plants having redundant drop lines, and operating data, we do not recommend adding a second drop line to plants that now have a single drop line configuration. Such a design change is being considered as part of A-45.

( Howev9r, if N".R's A-45 task concludes that a single drop line configuration is unacceptable, and additional reliability is required, it is recommended that NRR consider a smaller diameter DHR suction bypass line as a possible alternative. The bypass line configuration which we believe worthy of consideration is one with remotely operated valves to which power is locked '

out (actuation to be performed outside containment), with manual overrides (inside containment) to provide additional assurance of their opening in the event of motor or power source problems. This design would represent an

-improvement over the Davis-Besse design which cannot be operated from ^

outside containment. (See sections 2.1 and 5.2.1.)

(5) AEOD recommends that NRR consider removal of the autoclosure interlocks to minimize loss-of-DHR events.

In order to prevent inadvertent DHR suction / isolation valve closures (during DHR system operation) it is recommended that NRR consider either requiring the removal of the autoclosure interlocks to the DHR suction / isolation valves, or requiring removal of power to the DHR suction / isolation valves when valve motion is not required. Prior to implementing this recommenda-tion, it is necessary to ensure that there is adequate relief capacity to prevent overpressurization of the DHR system. (See sections 2.2 and 5.2.2.) ,

NSAC-52 (Ref. 3) had a similar recommendation.

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'.' ~. N i . (6) .AEED' ricommEnds rthat.ERII's tec'hriicaN idecificati6fi" imp'rovem$nCp~rog[am .

" address the issue of DHR. system redundancy to ensure that the DHR system is-

\ 'available during Mode 4 and the early stages of Mode 5. .

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.In section 4.2, we'noted.that even though"NRR's generic letter o'n'DHR 1 addressed.DHR system redundancy, plant technical specifications do not

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req'uire DHR ' redundancy throughobt.perio'ds of high riik;(mode 4'a.nd'the'early j stages.of mode 5). .We also noted.that test, maintenance, and .other, shutdown activitie's can'be init'iat'ed during th'se e ~perdodse As a resu'lt,'.tliere is as

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high likelihood that a DHR loss could occur, at a time when the risk is highest. Upon considering operati.onal data and the plant practices, we believe that regulatory action is necessary to minimize the possibility'of DHR losses during periods of high risk (early in shutdown).

We recommend that NRR's technical specification improvement program address the DHR system operating requirements so that licensees modify plant

. technical specifications to: ,

o assure all planti have' proper shutdown mode definitions-(as' discussed'in sections 4.3 and 5.3); and o ensure that both trains of the DHR system are operable durin'g periods of high decay hedt lo6d, i.e., mode 4 and the early stages of mode 5.

(Presently, the applicable generic letter permits one train to be inoperable during this time.) -

Since tht less-of-DHR experience has not greatly improved following the issuance of NSAC-52'and NRR's generic letter, we believe that technical ,

specification modifications are necessary to ensure adequate redundancy. In addition, we feel that an information notice should be issued to reemphasize to the licensees the overall safety significance associated with the l operation of the DHR systems.

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7'.OlREFERENCES M~ '

1. U.S. Nuclear Regulatory Commission, " Reactor Safety Sludy .An

( Assessment of Accident Risks'in U.S. Commercial Nuclear. Power Plants,"

WASH 1400 (NUREG-75/014), October'1975.*

2. 'JA'. ' Haried, Oali Ridge' National' Labo'rator'y, "Evalua' tion'of' Events Involving Decay Heat Removal Systems.in Nuclear, Power Plants," USNRC '

Report NUREG/CR-2799, July'1982.*

3. NuclearSafety'AnalysisCenter/ElectricPowerResearchInstitute,

" Residual Heat Removal Experience, Review and Safety Analysis,.

. Pressurized Water Reactors," NSAC-52, January 1983. Available from Research Reports Center (RRC) Box 50490, Palo Alto, CA 94303.

4. U.S. Nuclear Regulatory Commission, " Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors," (NUREG-0103 Rev. 4), Revision of Fall 1980.*

5. Letter from H. B. Tucker, Duke Power Company, to H. R. Denton, NRC,

Subject:

. Catawba Nuclear Station Docket Nos. 50-413 and 50-414, dated October 13, 1983."*

6. Memorandum from B. W.' Sheron, NRC to RSB members, " Auto Closure-Interlocks for PWR Residual Heat Removal (RHR) Systems," January 28, 1983."*

( 7. . Tennessee Valley Authority., Licensee Event Report (LER) 50-328/83-101 Sequoyah-2 Nuclear Power Plant, dated August 18, 1983."*

8. U.S. Nuclear Regulatory Commission, Region IV, Daily Report, August 31, 1984.**-

9. Arkansas Power and Light Company, Licensee Event Report (LER) 50-368/84-023, Arkansas Nuclear One - Unit 2, dated October 1, 1984.*"

10. Telephone Discussion between D.B. Lomax and J. T. Enos, Arkansas Power and Light- Company, and H. L. Ornstein, NRC, November 9,1984.

11. Indiana and Michigan Electric Company, Licensee Event Report (LER) 50-316/84-014, D. C. Cook Unit 2, dated June 22, 1984.**

Available for purchase from National Technical Information Service, '.

Springfield, VA 22161.

- Available in the NRC Public Document Room at 1717 H Street, N.W.,

Washington, D.C. 20555 for inspection and copying for a fee.

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. I I2. Letter 'f roni R.-d. eRodfigusf.25acramentofkuni41[allUtifitNiOi.rict? ' '

to J. F. Stolz, NRC,' Subjecti Docket No. 50-312 Rancho Seco Nuclear '

Generating Station. Unit-No'. 1, Low Temperature Overpressurization

, . r Protection.

- , . , , . ... (LT0P)...Setpoint,

.s . , . . . .dat.egi, February.. 15 , .1984.,. *, y

. . , s.

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13. U.S. Nuclear' Regulator Commission, .0ffice for Analysis'.and Evaluation.

'of Operational Data; Case Study-No' AE00/C401 '.' Low Temperature'-

, Overpressure, Events .at Turkey, Point Unit l4," March.1984.*. '

14. Nuclear Safety Analysis Center / Electric Power Research Insti.tute;

" Zion Nuclear Plant Residual Heat kemoval PRA " NSAC-84, ' July 1985. ,

Available from Research Reports Center (RRC), Box 50490, Palo Alto, CA j 94303.

15. Letter f rom A. D. Rossin, Electric Power Research Institute, to C. J.

Heltemes, NRC, September 3, 1985.

16. U.', S Nuclear Regulatory Commission, " Report to Congress on Acnormal 1 l Occurrences, April'- June 1980," NRC - (NUREG-0090, Vol. 3,'No. 2),

November 1980.**

17J U.S. Nuclear Regulatory Commission, Of fice of Irispection' and' Enforcement, Bulletin No. 80-12: " Decay Heat Removal System Operability," May 9, 1980.*

18. W. J, Foley, R. 5. Dean, A. Hennick, Parameter Inc. , " Closeout of IE Bulletin 80-12: Decay Heat Removal System Operability," USNRC Report NUREG/CR-4005, June 1985.** 1

19. Duke Power Corporation, Reportable Occurrence Report R0-270/81-17, Oconee 2, dated November 13, 1981.*

20. Institute of Nuclear Power Operations, " Analysis of Steam Generator Tube Rupture Events at Oconee and Ginna,"82-030, November 1982.

21. Pacific Gas and Electric Company, Licensee Event Report (LER) 50-275/

84-004, Diablo Canyon Unit 1, dated February 2, 1984.*

22. Commonwealth Edison Company, Licensee Event Report (LER) 50-295/84-031, Zion Unit 1, dated October 16, 1984.*

23. U.S. Nuclear Regulatory Commission, Inspection Report No. 50-206/84-04, 50-361/84-27, 50-362/84-28, San Onofre Nuclear Generating Station, December 21, 1984.*

. Available in the NRC Public Document Room at 1717 H Street, N.W.,

Wcshington, D.C. 20555 for inspection and copying for a fee. !

- Available for purchase from National Technical Information Service, Springfield, VA 22161.

J l

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1

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. - . 4 5 .- -

54. U'..S'.fNuclear" Regulatory C6mmiss'ioii,4" Generic 'LettSr ' AllOpIritfhh '

. Pressurized Water Reactors (PWR's)," from D. G. Eisechut, June 11,

[ t.1980.*

H.,!

. , . . , . . , . . . . . , .....i. - s. - .c ~ . . . ~ r s . .. -

'25. ~ U. S. : Nuclear' Regtilatory -~Commiss'lon'insp'setion Report" _ 50-269/81-34, 50-270/8,1-14, and'50-287/.81-14, Oconee Facility, July. 23,:1981.*-

. 26. Letter f rom. WLO. . Parker, Duke Power Company, to J. . P. O' Reilly', NRC ,

Subject:

Oconee Nuclear Station," Docket No. 50-269 . July 31,'1981.*

27. Nuclear' Safety Analysis Center / Institute of Nuclear Power Operations,

" Steam Voiding in the Reactor Coolant System During Decay Heat Removal Cooldown," Significant Event Report 91-81, October 26, 1981.

28. Florida Power Corporation Inter-office Correspondence - Operations Advisory from P. F. Mckee (Nuclear Operations Superintendent) to Licensed Operators, April 21, 1981.

'29. U.S. Nuclear Regulatory Commission, Office of Inspection and !

Enforcement, Circular No. 81-10: " Steam Voiding in the Reactor Coolant. !

System During Decay Heat Removal Cooldown," July 2, 1981.*

30. D. R. Gallup, D. M. Kunsman, M. P. Bohn, Sandia National Laboratories, Potential Benefits Obtained by Requiring Safety-Grade Cold Shutdown Systems," USN"C Report NUREG/CR-4335, July 1985.**

f 31. Memorandum from H. R. Denton, NRC to R. M. Bernero, " Schedule for

\ Resolving and Completing Generic Issue No. 99--RCS/RHR Suction Line Interlocks on PWRs," August 13, 1985.*

32. Memorandum from H. L. Thompson, Jr., NRC to H. R. Denton, " Maintenance and Surveillance Plan," August 2, 1984.*

33. U.S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Requirements," II.F.2 Instrumentation for Detection of Inadequate Core Cooling, (NUREG 0737), November 1980.**

Available in the NRC Public Document Room at 1717 H Street, N.W.,

Washington, D.C. 20555 for inspection and copying for a fee.

- Available for purchase from National Technical Information Service, Springfield, VA 22161.

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.A .' . , * * -

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-- Loss of- Decay' Heat Removal Systems - .

,at U.S. Pressurized. Water Reactors. .m.~. .

..,,s..,...,.

.. . . , . . '" ' ' '*~'

During 1982 ahd T983 '

' Plant ,

Date' Docket # LER # Description .of Event ,

-j '

, ' Ginna 04/12/83 50-244.83-015 Air binding of = RHR pump (1zmin. loss)

Ginna 05/01/83 50-244 83-017 Filling reactor refueling cavity - low RWST. Secured "A" RHR pump - Suction valve on operating "B" pump was closed.

' { Duration of event ' unknown)

- Turkey Point;3 .10/07/83 50-250 83-018 Flow restriction"on component cooling water discharge valve on RHR heat exchanger.

(Duration of event unknown) ,

Turkey Point 3 .10/08/83 50-250 83-019 Procedural. error during surveillance testing resulted in closure of suction / isolation valve (6 min. loss). '

l-Salem'1 03/16/82 50-272 82-015 Vital bus tripped. C& 4onent:

' cooling water and service water.

were lost. Redundant trains were out for maintenance.

(45 min. loss).

Surry 1 05/17/83- 50-280 83-024 Inaccurate standpipe level indication - low RCS level, RHR pump cavitated. (Duration of event unknown)

Zion 1 03/17/82 50-295 82-01) Inadvertent (contractor person-l nel) opening of inverter output l

breaker caused closure of the RHR pump suction valve. (3 min. loss) l' 50-311 83-024 RHR suction valve closed during Salem 2 05/14/83 05/15/83 operation. (Duration of events (2. events) unknown) ,

The 05/14/83 event was triggered by f i

a vital instrument bus which was de-energized for maintenance; the l

05/15/83 event was triggered by a failed comparator.

a 9

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- .Date" ' Docket- # c LER #s , Description..of.. Event. . .m Of. s Plant .-

05/24/83 50-311 83-025 - RHR pump trip caused by. logic /

Salem 2-circuitry problem on'the " safe-

, guards equipm,ent control" (SEC) system. (Duration of event '

unknown) 50-311 83-031 Loss of RHR pump due to spurious Salem l 2 06/23/83 actuation of the.SEC system.

- (Duration of event unknown) 50-311 83-032 Failed. gasket in joint down-Salen 2 l 06/ 23/E3 stream of' check valve flooded, the service water bay., Lost all service water, and thus cooling water to the RHR pumps, diesels, etc. (Duration of event unknown) 50-311 83-062 Vital bus transfer caused voltage belem'2 11/28/83 spike which resulted in closure of suction / isolation valve. (Duration of event unknown)

( Sale, 2 12/20/83 50-311 83-066 Loss of vital bus - due to personnel error resulted in closure of suction / isolation valve (22 min.- loss).

06/2"/82 50-312 82-015 Simultaneous test and maintenance Rancno Seco caused failure of bus, closure of the suction / isolation valve, and loss of DHR flow. (Duration of event unknown) 50-317 82-026 Spurious opening of breaker Calvert-Cliffs 1 05/17/82 f rom tne operating DHR pump (2 min. loss).

Calvert Cliffs 1 10/12/F3 50-317 83-061 Inadvertent isolation of shut-down cooling - caused by not '

deactivating isolation system when performing a hydro test '

l; on instrument sensing lines o,' ..

(30 min. loss).

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g IDate Do'cket' # 'LER E ' Descri ptLihn ' of ' Event Plant -

.vU MCalverti$.liffN2f.n .1)./22/S2'. ,.550.-3)8,n.7. 82:053,..j,; Technic,ian ~'

a power supply' pane 1;' caused

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~

closu're.of a DHR return. valve. '

(4' min. lois)[ ' '

50-318.82-054, D'HR lo'st due to' .a .f ai ed power.

Calvsrt Cliffs 2 15/24/82 supply. (Durat. ion of event unknown).

50-318 82-055 Vital inverter failed, caused Calvert Cliffs 2 12/28/82 an isolation of the DHR return line. (Duration of event unknown).

Calvert Cliffs 2 01/04/83 50-318 83-001 Inverter tripped during surveillance testing - caused isolation of the DHR. return .line; (15 min. loss).

50-318 83-005 Test procedure er g . Operating Calv'ert Cliffs 2 01/07/83 DHR pep sto.pped due. to test' of recirculation actuation signal

-(9 min. loss).

Sequoyah 1 09/16/82 50-327 82-116 Power was removed to allow modification work on solid state protection system; RHR suction valve closed. (Duration of event unknown) 50-328 83-101 False RCS level indication by Sequoyah 2 08/06/83 makeshift tygon tube and rubber hose level instrument. RCS temperature rose from 103*F to 195af in 77 min. Plant had been shut down 18 days earlier.

Beaver Valley 1 50-334 82-018 Failure to start RHR pump due to 05/12/82 circuit breaker problem. RHR pmp that had been operating was erroneously secured prior to attempt to startup idle pump.

(2 min. loss)

Beaver Valley 1 06/29/83 50-334 83-020 Construction worker made an error in making a design modifica-tion. De-energized bus feeding RHR pump - faulty procedures and communications between shifts (92 sec. loss).

- t.

Plant ,. _ , Date. ,, ,. Docket # .LER #, , ..pescrip_ tion.of. Event.

St. Lucie 1 03/29/83 50-335 83-021 Construction workers shorted a power' supply caus'ing closure of DHR suction / isolation valves

.(10 min. loss)..

Millstone 2 01/06/82 50-336 82-002 Technician error during a pre-ventive maintenance test resultet in loss of a vital instrument panel, and autoclosure of the suction / isolation valves (7 min.

loss).

North Anna 1 10/19/82 50-338 82-067 . RCS . drained to .below centerline (2 events) 10/20/82 of hot leg nozzles. RHR suction was lost because of low RCS level and incorrect level indication.

(10/19/82, 36 min. loss; 10/20/82, 33 min. loss).

North Anna 1 01/22/83 50-338 83-003 Failed inverter, ca'used RHR suction / isolation valve to close (4 min. loss).

North Anna 1 02/18/83 50-338 83-009 Both RHR pumps were cavitating.

Cause not deterinined -(5 min.

loss).

North Anna 2 05/20/S2 50-339 82-026 Lost suction to RHR pumps due (3 events) to draining of RCS and erroneous level indication (8 min., 26 min. , I hr. losses).

North Anna 2 07/30/82 50-339 82-049 Lost suction to "A" RHR pump due to draining. Diagnosed as a pump problem. The "B" was then started and it also became airbound (46 min, loss).

North Anna 2 04/14/83 50-339 83-023 Operator inadvertently opened a breaker, causing RHR suction /

isolation valve to close (<1 min.

loss).

North Anna 2 04/29/83 50-339 83-036 Loss of vital bus. RHR suction /

isolation valve closed. Caused by maintenance personnel con-ducting a test as loads were being transferred (<1 min. loss).

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:xDocVet'4 Y:tERf:i *:Descr1ption of Event *. :..s.y .* :

. .; North' Anna 2~ 05/03/83 50-339 83-038' Inadequats monitoring of -RCS level.

B .

' Loss of RHR pump suctiori. (Duration

.of event unkn,own) .,,

' Fa rley . 2 09/28/83' 50-364'83-042 Operati.ng RHR pump failed while redundant pump was secured.

.(Duration of event' unknown)

McGuire 1 03/02/82. 50-369 '82-024 Low RCS level due to vessel draining and inaccurate level indication. Operating RHR pump started to cayitate, the other pump was undergoing main -

tenance. (Event lasted 50 min. -

a licensee analysis indicated that 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .were available prior to the onset of boiling.)

McGuire 1 06/24/82 50-369.82-053 Inverter failure caused closure.-

of suction / isolation valve (6 min. loss)

McGuire 1 04/05/83 50-369 83-017 Low RCS level due to vessel drain and valved out-level sensor. Both RHR r rps cavitated. (Durationof event unknown) 50-370 83-092 Low RCS level due to draining

' McGuire 2 12/31/83 and inadequate level indication.

Running RHR pump had no flow (43 min._ loss).

11/12/84 50-395 83-136 Bus transfer during plan't modifi-Summer cation caused an interruption of power to an ESF instrumentation bus. An erroneous overpressuriza- i tion signal resulted causing suction / isolation valve closure, and interruption of DHR flow (5 min. loss). .

l 9 ' e 4

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.,,. . ~. .

.s : P ..Appendi x ; B : . c p v

>,s..:. ., 9; p ;y.,. \ .

b ..

Selected Loss of Decay Heat Removal System Events

.at .U.S Pressurized Water Reactors During 1984 t

Pla nt Date D:aet # 'LER # Description of Events

~

Zion 1, 09/14/84- 50-295 84-031 While draining the .RCS in preparation for primary -

secondary leak testing, the RCS level dropped below the

r. DHR suction line. The . liquid level was.being read from a manometer type arrangement.

Incorrect level measurement resulted from the fact that the manometer reference leg-was pressurized by nitrogen purge gas. RCS temperature increased from 110 F to 147'F (45 min. loss).

(84-002 While testing the pressurizer Salen 2 '02/09/84 50-311 overpressure protection system a procedural error resulted in automatic closure of a suction /

isolation valve (17 min. loss).

'D.C. Cook 2 -

05/21/84 50-316 84-014 Procedural error with a partial-e ly drained RCS. Simultaneous operation of two DHR pumps caused vortexing at the loop suction. Both pumps became airbound (25 min. loss).

North Anna 2 10/16/84 50-339 84-008 Clogging of a standpipe used for RCS level monitoring resulted in a 64" error. Upon introduction of air, the operating pump cavitated.

The redundant pump was started and it also cavitated. Both pumps

, became airbound (2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loss).

_ _ _ _ -- - - - . _ ___ . - - - - d

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~

Date Docket #' LER # Descrip'tio'n' of'. Event '

, ' Plant

c. ..c?- . '. .. . . .

Trojan 05/04/84- 50-344- 84-01 0 ' During RCS draindown fau'lty. -

level measurement led to

- air binding ~of the RHR pump.

The RCS was vented to atmosphere, ~

s A tygon. manometer configurati.on was being used to~ measure RCS level, however, " crud blockage" of the manometer tap led to !

erroneous. level measurement.

RCS temperature went from 105"F to 20l*F (40 min. loss).

ANO-2 08/29/84 50-368 84-023 During RCS draindown, faulty level instrumentation led to air binding of the DHR pump. A tygon manometer config-uration was being used -

however, the. operators did not-account for reactor vessel pressurization due to the presence of nitrogen purge gas.

RCS temperature went from 140'F to 205'F (Approx. 35 min. loss)

I McGuire 2 01/09/84 50-370 84-001 During draining operations, a procedural deficiency led to inadequate NPSH/ air entrainment of the DHR pumps (1 hr. 2 min.

loss).

McGuire 2 01/15/84 50-370 84-002 Personnel error during testing.

Re-energizing power to the breakers for the suction /

isolation valves caused automatic closure of the suction / isolation valves. Valves were opened manually (49 min. loss).

Summer 10/18/84 50-395 IE 1 DHR loop was out for surveil-Daily lance testing. An inverter Report failure caused closure of the operating loop's suction isolation val ve. (25 min. loss).

Summer

~

11/06/84 50-395 IE A procedural error in testing Daily relays on the bus supplying the Report DHR pump caused the bus to strip.

The associated diesel was out for maintenance (7 min, loss).

________________-__-_____-____m._______ ____ -- _ _

B l'

- 5?'-

.. , ,1 .

, .. - ; Appendix C ,

a , .,.. ; ,, - , ,

Decay ' Heat ' Removal System ' Loss'es 'at Davis-Besse

~

- Some of the most striking data on DHR losses 'comes from a. r'eview of the l Davis-Besse pl' ant'so' perating experience. From 1918 - 1981', thh Duis-Bessp- ~

L plant 1 accrued the. largest. numbe' r of reported DHR losses of 'any PWR: 16 .

events.* Seven..of those. events involved automatic closure of the suction /-

isolat' ion valves. There had be~en 'sev'en previous closures of the 'stiction/-

isolation valves during plant 'startup and testing; however,' those seven events are not included in the tally of 16. Subsequent to 1981, there have been no DHR losses at the Davis-Sesse plant. A detailed review of that plant's experience is quite enlightening.

Table Ck presents a listing of DHR suction / isolation valve events which have taken place at the Davis-Besse plant. Table C2 presents descrip-tions of all loss-of-DHR events wnich occurred at the Davis-Besse' plant from startup testing through 1983, Most of:the inadvertent closures of the suction / isolation valves were due to human .f actors (operator errors, incorrect procedures, lack -of. procedures, ~

etc.),and' resultant failures of power supplies to the safety features actuation system (SFAS) channels. Most of the events which occurred sub-sequent- to power operation were of short duration (four lasted four minutes ,

or less and one lasted 18 minutes). Recovery from most of those events !

required only ' clearing the perturbing signals and reopening the isolation

' valves. There were two events which. lasted much longer:

f o 'On' April 19,1980, a 2 }-hour loss-of-DHR event was initiated by a f ailure of an instrument bus, which eventually resulted in the closing of the suction /1 solation valves. Restoration of the DHR system was impeded by air binding of the DHR pump. The licensee's lack of procedures.

for restoring the air bound pump and the extensive modification and maintenance activities tnat were Deing conducted at the time contributed to tnat event, o That event was reported to Congress as an Abnormal Occurrence (Ref.16).

The abnormal occurrence determination was made on the basis that the event represented a " serious deficiency in management or procedural controls in major areas."

o On July 24,1980, a 50-minute loss-of-DHR event occurred as a result of suction / isolation valve closure. That event is of considerable interest because restoration of the DHR system was accomplished by using the suction bypass line to establish a flow path for the DHR system.

The 16 loss-of-DHR events were complete losses of the DHR system function when the DHR system was required to remove decay heat.

(

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3 ;, . .

b - *~ '

4 4 '

\g

'Af ter 4he 14th inadvertent' suction / isolation val've clohre event, the' NRC

+

~

.ap'p roved'an amendment to Davis-Besse's technical. spec.ifIcations allowing H'

removal ^ of power from the DHR. suction / isolation valves 'dur,ing plant shutdown.

(To preclude returning the plant to power with'.the .DHR suction /. isolation . ,

val'ves Open, the. pressurizer heaters are interlocked with the'9HR suction /

isolation' valves. The pressurizer heatbrs 46%ut be sctivated above the setpoint of the 'DHR' re' lief valve if both' of the DHR isola' tion valves are open. The pressurizer heaters are also interlocked so that if only one DHR isolation valve is closed, the heaters ~ will shut off at a pressure

% below the DHR system design pressure.)

Subsequent to the implementation of the aforementioned technical specifica-tion amendment,' the Davis-Besse plant has not experienced any further inadvertent DHR suction / isolation valve closures, it appears that.the Davis-Besse plant's solution to the spurious DHR suction / isolation. valve closure problems which have led.to many loss of DHR system events has been-effective. Furthermore, we note that as a result of the April 19, 1980 event (Zi-hour DHR system. loss), the licensee.took action to modify operating-and emergency procedures to minimize the possibility of a recurrence.

Additional guidance was given to the plant staff on how to recover.from loss-of-DHR events, including the venting of DHR pumps, and the implementation - '

offbackup cooling sources. Steps were also taken to improve administrative controls during shutdown.

^

NRR's June 11, 1980 generic letter on DHR (Ref. 24) requested all PWRs to '

amend their technical specifications to provide for redundancy in DHR capacity. In response to NRR's generic letter on DHR, the Davis-Besse plant submitted an amendment to their technical specifications, indicating that an operable system will always be kept in a standby state during modes 3-6 in order to. assure continuous DHR in the event that the operating heat removal system should f ail.

Subsequent to implementing the aforementioned improvements in DHR system ;

. operation, there have been no similar. losses of the DHR system at the Davis-Besse plant, it appears that the corrective actions that were taken at the Davis-Besse plant have resulted in a substantial improvement in DHR system operation.

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.l; DHRSucYion/Isolitihri

p ',Ns: Clo'suFe'Ivintsa't

- Davis-Besse Causing a loss of the DHR System .

L '

W. ,

' Duration of, DHR ' System 1.oss DEvent-Date . LER~#

f LMay 14,(1977 .77-006 Not stated 'during plant startup and ' testing May.14, 1977- 77-007 'Not stated - during plant startup and testing May 27, 1977.77-002 Not stated - during plant.startup and testing

. May.28 ,.1977 77-003 Not' stated -during plant startup and testing June.12; 1977 77-005 Not stated '- during plant startup and testing

' July 22, '1977 - 77-009 Not stated - during plant startup and testing June 28,.1979- 79-067 18 minutes l April 19,1980 80-029 23 hour2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months s-May ' 28, 1980 80-043 '2 minutes

~ July 24,' 1980 80-058 50 minutes Augus't'8, 1980 80-058- - 3 minutes August 13, 1980 80-060 5 minutes t

l l

- Fr -

Table 'C 2.' . Losses of,th'e Dhh System at. Davis-Besse, ,

.LER # 1 ate ' Description of Event 77-006 May 14,1977 During plant startup and testin'g', an IAC mechanic ;

caused.a sh' ort, .thereby tr.ipping .an SFAS and an ,,

RPS channel. 'While trying .to replace the blown fuse, an operator.de-energized the wrong SFAS and RPS channels, thereby caus'ing-SFAS Tctuatibni closing the.DHR isolati'on valves. -(Duration of event.. unkn'own ) .j 77-007 May .19, 1977 During plant startupl and testing, operator error caused a loss of essential power to an SFAS channel.

!- An error in restoring the power resulted in de-energizing another. SFAS channel. SFAS actuation resulted, causing the. DHR isolation valves to close.

(Duration of event unknown)

During plant startup and testing, while replacing

~

77-002 May 27,19 77 -

a cover on a junction box containing an SFAS channel, a. loose output lead shorted, resulting in closure of a DHR isolation valve. (Duration of event unknown)77-003 May 26,197 7 During plant startup and testing, a procedural

/- error in recalibrating RCS pressure bistables on

\1 an SFAS channel resulted in closure of a DHR isolation valve. (Duration of event unknown)77-005 June 12,1977 During plant startup and testing, operators did not follow their procedures -for SFAS monthly tests. As a result, a DHR isolation valve closed.

'Deatier. of event- unknown)77-009 July 22,1977 During plant startup and testing, while inspecting (2 events) for loose electrical insulation, an 1&C mechanic caused a current surge, which resulted in closing a DHR isolation valve. About 15 minutes later, after restoring the DHR flow, he caused another (identical) event which resulted in DHR isolation valve closure. (Duration of event unknown)78-060 May 28,' 1978 DHR flow was lost for about 2 minutes. An operator accidentally bumped a control switch de-energizing the bus supply power to the DHR pump.

L( 1 l

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, . Table C,2. (Continued)l - -

'

l

'p ~

LER# 'Date ' '

'De.scribtion of Event'78-067 June 15,1978 Three loss:of DHR events lasting & total of about (3 events) 2 minutes. Power was interrupted to the operating

' . DHR pump. .The other pump was inoperable .at.th'e time. . Maintenance personnel . accidentally ~ bumped-a relay tripping the operating DHR pump. An operator made two errors while trying to transfer poner to an essential bus (resulting in two other power interruptions to the pump).79-067 hoe 28,1979 18-minute loss of DHR. During surveillance test-ing, a slipped alligator clip caused a short circuit and failure of power supply to an SFAS channel. As a result, DHR suction valve closed. 6 80-030 April 18,1980 .29-minute loss of DHR. Leakage of RCS water through a partially closed valve resulted in inadequate DHR pump NPSH and erratic pump flow operation. The pump was secured until the leak was stopped and RCS level restored. During the event, RCS temperature rose from 93*F to 103*F.80-029 April 19,1980' 2 j-hour loss of DHR. Vibration from construction

( work actuated a ground . fault relay. Due to an .

abnormal electrical lineup associated with outage {

Jctivities, loss of power resulted in SFAS actu- '

ation. Control power to the DHR suction valves was lost, causing the suction valves to close. The SFAS actuation transferred the DHR pump suction

' to the BWST and then to the empty sump. The pump became airbound. RCS temperature increased from 90 F to 170 F while the vessel. head was detensioned ;

(140*F is the maximum temperature allowed while the '

vessel head is detensioned).80-043 May 28,1980 2-minute loss of DHR due to an inadvertent closure of a DHR isolation valve. An I&C mechanic was checking out a plant modification. Due to a test procedure inadequacy, the isolation valve interlock circuit was actuated, and the valve closed.

l 80-044 May 31,1980 8-minute loss of DHR flow. The operating DHR pump was secured by a control room operator.

i (An I&C mechanic took a DHR flow meter out of l- service to perform surveillance testing. Control )

room personnel were unaware of this. Upon '

l seeing that the DHR system flow had dropped offscale, a control room operator stopped the

[ pump.) i l

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. Table C 2. -fContinued)

LER#- Date- Description L' 80 049- June 14,1980 DHR pump. flow loss for about 2 minutes. ;

Inadvertent 5FAS actuation caused DHR pump realignment to the BWST and_ BWST isolation. An 18C mechanic was restoring containment pressure inputs to SFAS following an Integrated Leak Rate Test. Because of a procedural inadequacy, SFAS was actuated and the DHR pump was realigned to .,

deliver BWST water to the RCS and the refueling !

canal. .When BWST level dropped to the low level limit,'SFAS level 5 actuation took place closing the BWST. isolation valve causing a loss of i suction-to the DHR pump..

\

1 80-058 July 24,1980 DHR flow was lost for. 50 minutes because of an

- automatic closure of an isolation valve. An electrician blew a fuse while conducting wire pulling operations associated with a plant design change. As a result of the blown fuse,-

, an automatic closure of one of the DHR isolation valves took place, and the pump became air-bound.

The DHR flow path was restored by opening the manual bypass valves. During the event, the hottest in-core thermocouple temperature rose from 104'F to 111*F.80-058 July . 24,1980 DHR flow was lost- for about 2 minutes due to an inadvertent closure. of one of the DHR isolation valves. Subsequent to making a plant modifica-tion, an I&C mechanic performed restoration work out of sequence. As a result, one of the isola-tion valves closed.

^

l80-058 August 8,1980 DHR flow was lost for about 3 minutes due to an inadvertent closure of one of the DHR isolation valves. Valve closure occurred during maintenance when a bistable in the valve circuit was removed !

due to a procedural error.

80-060: August 13, 1980 DHR flow was lost for about 5 minutes due to an inadvertent closure of one of the DHR isolation valves. Valve closure occurred during SFAS

, channel modification work. The 18C mechanic

' f ailed to fully defeat the automatic isolation valve trip prior to performing SFAS channel l moc'ification work.

l

.___-___A

4 4

t

.[ . Table C 2. < .(Continued) ~

LER#. 'Date Description 81-004 . January;7, 1981 DHR pump failed to start due .to a breaker -

problem. Electricians were able to restart the pump after a 15 minute delay.

L. 024-Apr il -.18, 1981 -

2-minute' loss of DHR flow. In response to "two-burning potential devices" on a bus, the bus was isolated. An error was made in-the sequence of transferring power and isolating the bus. Power' .i

- was lost to the operating DHR pump.

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ENCLOSURE 2 ,

CALCULATIONS' i i

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Verrnor t L< r d ee SFP - Part I - 12 Month (Normal Heat Load Calculati on)

Betch No Buri Decay Heat Cumm Heat Cumm 44 Bun Ts

. ----- ------- ------ ---- -- --- --- ---.-- -- ------ - .=----- -

1 104 .11113 .12113 104 19.00 2 108 .11267 .22380 212 20.00

~

120 .12223 .34604 332 21.00 4 80 .07956 .42560 412 22.00 5 96 .09322 .51882 508 23.00 6 96 .09102 .60984 604 24.00 7 84 .07776 .68'759 658 25.00 0 136 .12292 .61051 824 26.00 9 328 .28944 1.09995 1152 27.00 10 30 .(C585 1,12590 1182 29.00

+ 8. 9 BWf lo, is est As sustep t i ot IE r Power p eer Dundle = 15.44197 MDT U/ Hr Irradiation Time -- 35040. Hours Reec t at Power a 1665. I1Wt h Ref er ence. NURErc GOO, Standard Revi ew Flan (1) Scttson 9.1.3 Sp r>n t Fuel Pool Cooling and Clean-up Systems, Rev. 2 (2 Branch Technice.1 Position ASB 9- 2 Petidual Decay Energy for L.i yh t-Wat er Reactors 4or Long Ter m Cooling, Rev. 2 9

i a

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' Ver rnon t . Yarikee EF P - Pert II - 18 Month. (Normal Heat Load Calculation)

~ '

i-i 1

l

.Datch NO Bun Decay Heat. Cumm Heat Cumm # Bun Ts i ..

.1 132s 6.46233 6.46233 132 .02 1

' 128 .45769 6.92002 , 260 1.52 3 132 .27578' 7.19500,' 392 3.02

4. 129 .23802 7.41323 ' 520 4.52

l. S 152. .20677 4 7.62060 652 6.02 6 l128. .39101 7.81160 780 7.52 7' 132 .18942 S.00102 912 9.02 8 '128 .17706 8.17808 1040 1O.52 9 332 .17612 8.35420 1172 12.02 iO 328 .36475 8.51895 1300 13.52 11 132 .16391 8.68296 1432 15.02

'12 336 .16293 8.84579 1568 16.52 13- 120 .13869 8.99448 1688 18.02 Anumptiens:

Power. per. Lundle

15.44197 MBTU/Hr Irradiation'Tirne = 39420. Hours.

F% actor Power ' 1665. MWth Reference ~ NUREG-OBOO, Standdrd Review Plan (1) Sec t i on 9.1.3 Spent Fuel Fpol Cooli ng and Cl ear-up Systems, Rev. 2 (2) Drench Technical Position ASB 9~2 Residual Decay Ene gy for Light-Wster Reactors for Long Term Cooling, Rev. 2 4

0

( , , . , , , . . - ..,.,_. ,. . . . . ._

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Ab i Ver v.on t Yenice EFP - Part la - 12 Month (A Nor tnel Heat Load Cel cul ati on)

Betch No Pun Decey Heat Cumm Heat Cumm # Bun Ts l

1 104 .11798 .11798 104 16.50

'. 108 .11962 .23759 212 17.50 J. 120 .1297'7 .367~6 332 18.50 4 80 .09447 .45183 412 19.50 5 96 .09896 .55079 '508 20.50 6 96 .09663 .64742 604 21.50 7 84 .08255 .72997 688 22.50 8 136 .13049 .86046 824 23.DO 9 ~.,2 a . aO'728 1.16773 1152 24.50 10 30 .02744 1.Ic/517 1102 25.50 4 t,o."ho d u,WG (esumptians:

-_... ~___

F- owe r per Dundle = 15.44197 MBTU/Hr Ir redi ati on Ti nn-. e: 35040. Houru Tsent t o. Fouer '- 1e65. MW t li Referorme: l .lUR E G - O '9 00 , 5tendard Review Plan (1) boc. t i on 9 . 1. ~,

Opent F uel Pool Cooling and Clean-up Systems, Rev. 2 (2) branch Technical Position ASB 9-2 hesidual Decay Ener gy for Light-Water Reactt,rc for Long Term Cooling, Rev. 2 s

_____________________--__._-a -

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c, a.

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Ver mont Yankc-e SFP - Part II - 18 Month (Abnormal Heet Load Calculation)

Th> t c h No Eun Decay Heat Cumm Heat Cumm it Bun Ts 1 36L 18.01620 18.01620 368 .02

.' 128 .45769 18.47389 496 1.52 3 132 .27578 18.74967 628 3.02 4 128 .21802 18.96769 756 4.52 5 132 .20677 19.17446 888 6.02 6 128 .19101 19.36547 1016 7.52 7 132 .18942 19.55489 1148 9.02 8 118 .17706 19.73195 1276 10.52 9 132 .17612 19,90306 1408 12.02 10 129 .16475 20.07282 1536 13.52 11 132 .16391 20.23673 1668 15.02 12 20 .02396 20.26069 1688 16.52 Assumpt i one :

Fower p er Fundle = 15.44197 MBTU/Hr 1rradiation Time = 39420. Hours Resctor F'ower = 1665. MWth

Reference:

NURE G -OBOO ., Standard Review Plan ,

(1) Section 9.1.3 Opent Fuel Pool Cooling and Cleen-up Syst ems, Rev. 2 (2) Branch Technical Pos,ition A5B 9-2 Reu dual Decay Energy f or Light-Water Reactors-for Long Term Cooling, Rev. 2 l

I i

l

_____.-____--_-____D

n:. ; .. .a .:e.; ^ ;:: c w ;vu:. ;; M. ;;.~. ~ - ' -- m . v.:

l s, ,"

Ver mont Yankee Spent Fuel Pool Expansion Batch- No Bun Decay Heat Cumm Heat Cumm # Bun Ts 1 13 . 6.4(233 6.46233 132 .02 2 128 .46216 6.92449 '260 1.50

'3 132 .27680 7.20129 392 3.00 4 128 .21833 7.41961 520 4.50 5 132 .20691 7.62652 652 6.00 6 128 .19110 7.81762 780 7.50 7 132 .18950 8.00712 912 9.00 8 128 .17713 8.18425 1040 10.50 9 132 .17619 8.36044 1172 12.00 10 128 .16482 8.52526 1300 13.50 11 1:; 2 .16398 8.68924 1432 15.00 12 136 .16299 8.85223 1568 16.50 13 120 .13875 8.99098 1688 18.00 14 104 .11741 9.10839 1792 19.00 15 103 .11904 9.22743 1900 20.00 16 120 .12914 9.35657 2020 21.00 17 80 .08406 9.44063 2100 22.00 10 96 .09049 9.53911 2196 23.00 19 96 .09616 9.63527 2292 24.00 20 84 . OE.215 9.71742 2376 25.00 21 136 .12986 9.84728 2512 26.00 22 329 .30215 10.14944 2840 27.50 23 30 .02698 10.17642 2870

28.50 6s sumpi 2 co ee .

Power per bonUI e = 15.44197 MBTU/Hr irr adi atit n 1itur 2 39420 Hours Reec t or Peer u 1665. MWth Time Atter Shut down = 150. Hours or 6.25 Days Ref er enc e: NUREG-0800, Stander d Review Plan

-( 1 ) Section 9.1.3 bprnt Fuel Pool Cooling end Cl een-up Systems, Rev. 2 (2) Brench Technical Posi ti on ASB 9-2 Resi dual Decay Energy for Light-Water Reactors for Lont Term Cooling, Rev. 2 Spent Fuel Pool l'aoling System ser vice Water System

)

i Out l et Temper etus et (Fi 153.61 143.15 Inlet Temperatures (F) 198.84 (Pool Temp) 85.00 a

Wei er F) ow hatte (Mlb/Hr) .225 .175 Heat Exchanger Surface Area (SqFt) 500.O Heat Exchanger Overall Conductance 375.0 BTU /SqFt-Hr-F The two pass correction factor, F= .873

o. . . . . .. - n .' : ' .% . ;..: :.;; ;: ' ' --- :. .

Ver mon t Yankee.8 pent Fuel Pool E>:p an si on l-l Decay Tir.a Decay Heat Pool Temperature F factor l (Hours) (Days) 010TU /Hr ) (degrees F) 1 150, 6.25 1O.17642 198.84260 .8734 158. 6.58 10.03828 197.24130 .8731 l 166. 6.92 9.91109 195.83780 .8732 174. 7.25 9.79359 194.51760 .8732 ,

182. 7.58 9.68468 193.30180 .8732 1 1

190. 7.92 9.58341 192.16860 .8732 '

198. 8.25 9.48892 191.11210 .8732 206, 8.58 9.40049 190.12320 .8732

.214. 8.92 9.31746 189,19470 .8732 222. 9.25 9.23927 188.32030 .8732 230. 9.58 9.16542 187.49440 .8732 j 238, 9.92 9.09546 186.71210 .8732 246. 10.25 9.02901 185.96910 .8732 254 1O.59 8.96574 185.26140 .8732 262. 1O.92 8.90532 184.58580 .8732 270. 11.25 8.84751 183.93930 .8732 27C. 11.b9 8.79206 183.31930 .8732 286. 11.92 8.73E77 182.72340 .8732 294. 12. 2"~, 8.68745 18'2.14940 .6732 302. 12.58 8.63793 181.59570 .8732 31O. 12.92 8.59007 181.06050 .8732 318, 13.25 8.54374 180.54230 .8732 326. 13.50 8.49882 180.04000 .6732 334 13.92 8.45520 179.55230 .8732 342. 14.25 8.41281 179.07820 .8732 350. 14.5e e.37154 178.61s70 .e732 358, 14.9? 8.33134 178.16710 .8732 366. 15,25 8.29212 177.72850 .8732 374. 15.SE 8.25383 177.30040 .8732 3E2. 15.92 8.21641 176.88200 .8732 4

M.

5 e

.. . ., . . L .. . " .?= 4.c 2 ' ; I w wcuc - , . . . .

.Ver%ont YanFee-Spent Fuel Pool Expansion (Normal Heat Load Calculation)

Datch No bun Decay Heat Cumm Heat Cumm M Bun To 1 2 ~> 2 6.46233 6.46233 132 .02 2 128 .46216 6.92449 260 1.50 3 132 .27680 7.20129 392 3.00 4 128 .21833 7.41961 520 4.50 5 132 .20691 7.62652 652 6.00 6 128 .19110 7.81762 780 7.50

7. 132 .18950 8.00712 912 9.00 8 128 .17713 8.18425 1040 10.50 9 132 .27619 8.36044 1172 12.00 10 128- .'16482 8.52526 1300 13,50 11 132 .16398 8.68924 1432 15.00 12 136 .16299 8.85223 1568 16.50 13 120 .13875 8.99098 1688 18.00 14 104' .11741 9.10839 1792 19.00 15 105 .11904 9.22743 1900 20.00 16 120 .12914 9.35657 2020 21.00

~17 80 .08406 9.44063 2100 22.00 i' 18 96 .09849. 9.53911 2196 23.00 19 96 .09616 9.63527 2292 24.00 20 84 .08215 9.71742 2376 25.00 21 136 .12986 9.84728 2512 26.00 22 329 .3021D 10.14944 2840 27.50 23 30 .02698 1O.17642 2870 28.50 Assumpti one:

Power p s. r D6ndir = 15.44197- MBTU/Hr 1rredietion Time u 39420. Hours Reac t or Power r 1665. MWth References NUREG-OBOO, Stendard Review Plan I

(1) Section 9.1.3 Spent Fuel Pool Cooling and Clean-up Systems, Rev. 2 (2) Branch Technical Position ASB 9-2 Residual Decay Energy f or Light-Water Reactors for Long Ter m Cooling, Rev. 2 Spent Fuel Pool Cooling System Service Weter System Out l et Temperatur es (F) 153.61 143.15 lolet Temperatures (F) 198.84 (Pool Temp) 85.00 Water Flou R6t es ( til l; / Hr i .225 .175 ,

. Heat Exchanger Surf ace Ar ea (SqFt) 500.O Heat Exchanger Over all Conductance 375. O BTU /SqFt-Hr -F The two pass correction factor, F= .873 ]

- - - - - - - - - - - - . - _ _ - - _ - - _ _a

y c ', : .. -~.;a; . . .:. ; ,..

^^~~~*~"~~ =

~

Wr'Toon t Yani:ee Spent Fuel Pool Expansion ~ ( Abnor mal Heat Load Calcul ation) i Datch No Bun ' Decay heat 'Cumm Heat' Cumm it Ban Ts t 1 368 18.01620 18.01620 368 .02 2 128 .46216 18.47835 496 1.50 132 .27680 18.75515 628 3.00 4 128 .21833 18.97348 756 4.50 5 132 .20691 19.18039 880 6.00 6 128 .19110 19. .~6714 9 1016 7.50 7 132 .18950 19.56099 1148 9.00

.O. 128 .17713 19.73812 1276 10.50 9 132 .17619 19.91431 1408 12.00 10- 128 .16482 20.07913 1536 13.DO 11 132 .16398 20.24310 1668 15.00 12 136 .16299 20.40610 1804 16.50 13 120 .13875 20.54485 1924 18.00 14 104 .11741 20.66225 2028 19.00 13 108 .11904.. 20.78129 2136 20.00 16- 120 .12914 20.91043 2256 21.00 17 80' .08406 20.99449 2336 22.00 18 96 .09849 21.09298 2432 23.00 19 96 .09616 21.18914 2528 24.00

.2 0 54 .08215 21.27129 2612 25.00 21 136 .12986 21.40115 2748 26.00 22 122 .I1239 21.51353 2870 ,

27,50 Assumptions:

l -----------

Power per Bundle = 15.44197 MBTU/Hr i Irradiation Time-== 39420. Hours Fieac t 6r Power e 1665. MWth

Reference:

NUREG-0800, Standard Review Plan (1) Secti on 9.1.3 ,

Spent Fuel Pool Cooling and Cl ean-up Syhtems , Rev. 2 (2) Brunch Technical Position ASB 9-2 Residual Decay Energy for Light-Water Reett or s for Long Term Cooling, Rev. 2 Sper't Fuel Pool Cooling Syst em Ser vice Water System '

Outl et Temperatures, (F) 229.93 207.93 Inlet Temperatures (F) 325.55 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr> .225 .175 Hu e t- En ch ancter Surface Area (SqFt) 500.0 l Heat Ex c.han g er Overall Conductance 375.0 BTU /5qFt-He-F The two pass correction factor, F= .873 l

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' Page 4 of 7 TABLE 1.7.6 (Continued)

Core Mechanical Design Vermont Yankee Dresden 2 l In-Core Neutron Instrumentation Number of in-core neutron detectors (fixed) 80 164 Number of in-core detector assemblies 20 41 Number of detectors per assembly 4 4 Number'of traversing-incore probe neutron 3 3 detectors Range (and number) of detectors Source' range monitoring subsystem Source to Source to 0.001% power (4) 0.001% power (4)

Intermediate range monitoring 0.0002% to 20% 0.0003% to 10%

subsystem power (6) power (8)

Local power range monitoring 0.01% to 125% 5% to 125%

subsystem power (80) power (164)

Average power range monitoring 2.5% to 125% 5% to 125%

subsystem power (6) power (6) 7 Sb-Be Number and type of in-core neutron sources 4 Sb-Be i Core Standby Cooling Systems h).

% High pressure coolant injection system (No.) 1 1 l Number of loops ~ 1 1 Flow rate (gpm) 4250 5600 Automatic depressurization system (No.) 1 1 i

Low pressure coolant injection (No.) 1 1 Number of pumps .

4 4 Flow rate (gpm/pum)) 7200 at 20 psid 4833 at 20 psid Auxiliary Systems Residual Heat Removal System q i

Reactor shutdown cooling (number of pumps) 4 34 Flow rate (gpm/ pump)ll) 7,200 5,3504 Capacity (btu /hr/ heat exchanger)(2) 57.5 x 106 27 x 106 (4)

Number of heat exchangers 2 3(4)

Primary containment cooling Flow rate (gpm) 28,000 RHR Service Water System Flow rate (gpm/ pump) 2,700 3,500 4 4 Number of pumps (1) Capacity, during reactor fooling made with 3 of 4 pumps running.

~

(2) Capacity during post-accident cooling mode with 1650F shell side inlet (J? temperature, maximum service water temperature, and 1 RHR pump and 1 RHR service water pump in operation.

(3) For~all 3 units.

(4) Separate shutdown cooling system.

_ - - - = _ . _ _ . _ . -

l l

J 1

VYNPS ,-

~Page 5 of 7

'9 '

TABLE 1.7.6 (Continued)

Auxiliary Systems Vermont Yankee Dresden 2 Reactor Core Isolation Cooling System Flow rate (gpm) 400 None Fuel Pool Cooling and Cleanup System capacity (Btu /hr) '2.37 x 106 3.65 x 106 Turbine-Generator i Design power, Mwe 1665 2527 Design power, MWe 564 809 Generator speed, RPM 1800 1800 Design steam flow, Ib/hr 6.721 x 106 9.945 x 106 q Turbine inlet pressure, psig 950 950 Turbine Bypass System Capacity, percent of turbine design steam flow 105 40-Main Condenser Heat removal capacity, Btu /hr 3605 x 106 Circulating Water System Number of pumps 3 3 Flow rate, gpm/ pump 122,000 Condensate and Feedwater Systems Design flow rate, 1b/hr 6.4 x 106 9.725 x 106 Number of condensate pumps 3 4 Number.of condensate booster pumps 4 Number feedwater pumps 3 3 Condensate pump drive ac power ac power Condensate booster pump drive ac power Feedwattr pump drive ac power ac power l

Transmission System Outgoing lines (number-rating) 2-345 KV 5-345 KV 2-17 KV Normal Auxiliary AC Power Incoming lines (number-rating) 2-345 KV 5-345 KV 2-115 KV 6-138 KV 1-4160V Auxiliary transformers 1 1 1 -

1 Startup tra,nsformera ,

Standby AC Power Supply 3 (for 2 units)

')'

Number diesel generators 2 Number of 4160V standby busses 2 2 Number of 480V standby busses 2 2

h~ f

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I t-g Verwoni YanFee Spe:nt Fuel Pool Ex paresi on o

t' .

.. Decay 12 hunth 18 Month Total Heet -SFP Weter Timecin; Cyr l e . Heat Cycle Heat Loed in Temper atur e Days; Lad Loed .MBTU/Hr- ( Deg r ect s F)

6.6 1 12500 9.09610 10.22190- 199.35

-7.O 1.11D72 S.65G99 9.81471 194.74 8, O 1.12G6G 8.36741 9.49306 191.14 9O '3.12D58 E.10DG1 9.23139 180.21 1O.O 1.12570 7.09696 9.01246 j OL 77 1'.~ J ' . I ' '5 /I '., 7.6991es 8.82459 183.68 l'2,. 9 1.12535 7.53439 8.65973 151.53 13 . ,U . .125"20 : 7.SE706 8.51234- 180.16 1-4. 0 1.12G21 7.25331 8.378S2 178.68 lJ b . . i 1.1;"U 13 7 1303D '8.2554G 177.30 av e 1.12506 7.'01619 8.14125 176 03

.O 1.12499 6.90439 8.03439 19 4 . P;'

13m0 1.12491 6.80836 7.93377 173,71 1 % O. 1.12484 6,'71379 7.03863 172.6D 2? . O '1.12476 6 6735/ 7.74G32 171.63

,.! 1. J. 1.,124 c 9 6.b3766 7.6623'. 170.67 22.6 '.12462 6.4"*571 7.500' .169.70 2J.O 1.'12454 6.37738 7.501; 168.BB

"' 4 D '1.1'M "' 6 30240 7.426E 115.04 25.u. ').12440 '6.23054 7.35494 167.24 2h O 1.12432 6.16161 7.28593 166.47 2 e 1. 12 v", 6. 0% D4Fi 7.21968 135.73.

T 1.124)2 6.O-'156 7.15602 165 01

'.a J . l41 L 97v70 7.09490 164.35 JL' . 12403 '5.7 1E5 7.03591 163.e7

v. 1.1230E 5. GOO 76. 6.92464 162.43 3 4. 0 -. 1.12370 5.69a61 6.92234 161.25 M. . o . j i .e. W 5.60!69 6.72528 160.19

36. n 1.12244 1 51234 6.63578. 159.20 A:2..O 1.12 Wi 5. 4 2PP7 (.55226 6 158.26 4..O 1.12314 5.35105 6.47419 '157.39 a n . .a 1. 3 ;. . '.0 0 '.27810

. 6.4911n 156 ' ~,

4 n. t :. 12205 5.20969 6.33254 151 EO C.<> . 1 2;'7 0 114541 6.2681: 155.0B 9.. 1.122 % 5.084f2 6.20748 154.40

5. . D 1.12241 5. O'I.7 89 6.1DO30 153.77 5 4 '- 1.1222e 4.97403' 6.09629 153.le So. O 1.1 M 1 : 4.92307 6.04518 152.D9 58.0 1.12197 4.87478 5.99675 152.05 60.O -1.121E2 4,82892 5.95074 151.54 62.O 1.-12167 4.7E531 5.90693 151.05 64.0 1.12153 4.74376 5.86529 150.59 66.O 1.12138 4.70411 5.82549 150.13 68.0 1.12123 4.66600 D.'7E743 149.71 75 3 J .1210c, 4.6.2991 5.75100 149.30 7/,0 j,1; Ova /' . 5 % 2 :s 5.71604 145.71

14. , i .1 N70 4.56i67 1 6E246 148.54

t. e 1.I'100 .- / 4.!290: 5.65010 140.17 76,o 1.120'.O 4.49854 S.61904 147. GTJ E0.0 1. ; 20."; D 4.4666e U.56901 147.4' i

.~,..-..w,, l

,, ., . n. t v. ~ %.m s... - - .m -ww .s

, . ' 1 1202s 4.43980 5.56000 147.17 0 7, 1.11 W- . i . ". ' m i' e 9. D1 EC 4 .146,70 100.0 '.1184' 4.'1160 5.33448 144.67 115.i.- 1.11 0 , 4. 06 S D'.) L.18159 142.9i 130.0 1.11o23 3.9340b 5.05030 141./7 145.0 1.11D).3 3.E1914 4.93429 140.17 149.O 11466 3.77069 4. 9055'~, 139.33 Rwt De t o: 23 .b n--87 N .. ; y il M a n '.. h .16 1onth Total Heat SFP Water ilm a ii L y c. , e ibt C, y c. l e Herit Loed ii i ic np er r.itr e De ye Lo ici Los cl MDTU/He (Degr ees, F) 6.O 1.09961 9.0311Et 10.13076 199,33 7.0 1.0995'l E.62300 9.72'I49 102.7:

' 1. M V46 8.3J.226 9.40172 190.13

'J . 1. 0"9 .' 9 8.040SD 9.13994 187.20

, . 1. > 9 9 3 ~' ~7. F.115 9 8.97090 164.,75 11 G 1.0992'.. 7.6336'/ 8.7 292 182.61 J0 1.r9915 7.465"E' O.567V6 10. 90 f e

1. 05 i l(- 7.32135 8.4204L 179.Je j 1a .0 1. 0"'?O7 ".1G7c8 8.29651 177.6v l'...'- 1. C%; 6 7. 06 4e' 1 E.16337 176.~i

' 1 , '4 " U 6.0501S c .04901 'i70.09

.I ~.o 1. 0; E Ei' 6., U/! 3: ~ ,.94205 J 7 7. 80 1 B. 0 1.09574 6,74260 7.64134 172. 67 17 .' !. MOW n . 6tl 7 % 7.7460t 17 61 1.09E60 6.50703 7.65569 179.60 c 1, < > i . Q ~/ 8. , 6.47107 7.D6960

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7.U 1. 09F.3D 6.31057 7. 4 0E'E 1 :. 7 . E"

.l.' 1.09031 6.23549 . 33380 167.C0

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26. < > 1.04817 6.09449 7.19266 .

165.63

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i 164 .

26.O 1.09B02 5.964bl 7.06253 16 ~J . 9 7

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2.09774 5. 73~' O;' 6.E307e 161.3E 3 2 . ' .' 1 . i. ,. . ..,.r-e . 7,

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,6 3 '.. . O 1. O V"< 4 5 5.53355 6.63100 159.14 39.0 1. 997 ~' O 3.44400 6.54130 '55.14 1 l

o . 'u7 e t.

4C.U 1 -5 . . , . . /. :

s . 6 v..ne 15 < . a... , I 42.o 1.09702 5.29252 6.37934 156,37 I 1

44.O i.09597 5.20918 6.30605 1DS.50 f 46.0 1.09o77 5 14057 6.23730 154.74 48.O 1.09659 D.07611 6.17270 104.O:'

50.O 1.09h44 5.01S4*.' 6.11187 153.34 D2.0 1. O'> e 3C 4,95821 6. 05/! D:' 152.70 54. 1. O?6 il 4.9041E 6 . 0< O-'" 152.09 l be. O,60' 4.00304 5.94912 151.52

. G 9'~.0 7 1:.80456 D.900/1~ 1 D . 97 o . E7 l. o . 7W . L . E D 4 2 '. . 150.4t -

1 e . i. 2.005S'; 4.71474 5.31032 1i:9.96 i 1

64.d 1.09544 4.6730 L . 7 63 4 :. 142. 49 i I

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66.O 1.09529 [4.65319, .5.7284S - 149.05

~60.O 1.09515 4.D9511 - 5. 6 ti O26 148.62 l '-

70.O 1.30501' ti . 55865 5.65366 - 148.22 l

< 72.O' 1.CY486 4.32367.. 5.61853 147.83 74.O- 1.09472 4.49008 U.58480 147.45 76.0 1.094SE 4.45776 0.53234 147.OE)

L '78.0 1.09443' 4.42662 U.52105 146.73-

'EO.n j . ov42c; 4.39659 D.490Ei7 1/16.4 0 G2.O 1.09414 4.36755 5.46169 146.07-EU C 1.07393 4.32570 0.41951' 145.60 200 0 1.092D5 4,14.?42 5.23527 143,,53 alb,

2.0i18 0.969d7 0. 0E,12'2 141.E1

E.O 09028 3,,8583A 4.94P64 140.34 J45.: .08921 3. 742T 4.83173 139.03 149,,0 1.05592 3.74303 4.E0275 135.70 4

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Ver tr.on.t - Yania.e la Month 16/18 Cycl e F4ir6- Nb ihr s Iwray Heet Cun,m Fleat - Curnm ti . Due T s.

1 1.

137 6.55339 6.DD339 132 .02

_. 125 .47166 7.02506 260 1.42 1:12 .27642 7.30148 392 2.82

'4. 128 .21278 7.51426 520 4.22 S 132- .19929 7.71355 652 5.62 6 123 .19349 7.89705 750 7.02

'/ 132 .1G20D 6.07909 912 8.42 3 '12L .l'7047 B.24957 -1040 9.82 9' 132 .16954 8. 41(?DO 1172 11.22 10 )7 . J '" 9 _"J. - G.57855 170^ 12.6'.

11 17 2- .15f t91 8.73770 1432 14.02 12 10e .15033 8.59608 1568 15.42

'11 1 '. O .13510 9. C 3110 1508 16.E2 A uwa:p t i en m :

Poiwr pw leunuir e- 13.41197 t1BTi>/Hr

) t r adi cti % ~1 i m r "J 5040. Hours l Reetter Powtn a 166D. t1Hl h

'Ti r ? .A f Lu; Chut a :mn - 1 ".4 . Houco or 6. Of' Days ,

Ref er erite; bluHEG- MNlv , St unda d F ev a s.w F l emn (1) Section C.1.3 Epen! Fuel Fool Coo]in; and Cl ean-up S y s t erno , hev. 2 (2 Prw cli Tetlinical Pos i ti on ASE 9-2 Rs bidtud Decey Ener gy for Light-Weter React or t.

f or Laiig Ter m Cooling, Rev 2 i .

Opent Fuel Pool Cooling Sys t ern Ec.r vi ce Water Syttern ;

Outl et Tenger stur c e (F' 145.89 136.61 f i nl et T eo.p er a t ur ev (F' 19'. 03 Wool Temp) E5.00 Water Flow R at ta 011 b /Hr ) .225 .175 Heat Exchanger Sur f ece Aree (SqFt) '500.0 Heat'. Exchanger Over 4 Il Conductance 375.O BTU /3qFt-Pr-F The tw:a pass correction factor, F= .873 i

en-. ..w -

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. Vernent Yankett 18 Munth~16/18 Cycle t

.e :f a n . .

'Decef Heat Paul Temperature- F factor

, I Hch > r t.) . : (Deys) 01DTU/Hr) (degrees F) l L

..7 _ _ . . _ . - _ . _ _ _ . _ . _ _ - --- . _ _ . _ _ _ _

j id 6. 00 9.03118 'l86.03090 .6734 .j 16 6 7.00 8.62395 181.42730 .8731 j %> 8.00 8.30226 177.84590: .5732 216. ,9.00- 8.04050 174.91400 .8732

'24O. 10.00. 7.82158 172.46710 .8732 264. 11.00 7.~63367 170.36510 .8732 238. 12.00 7.46878 168.52140 .8732 312. 13.00 7.32135 166.87270 .8732 "M. 14.00 7 .187 4 E- 165.37570' .8732 J60.1 1G.00- 7.064A1 163.99940 .8732 3N+ .16.00 6.95015 162.72160 .8732 l 403. 17.00 6.84323 161.52610 .8732 432. 1 E*. 0 0 6.74260 160.40070 .8732 4D6. 19.00 6.64742 159.33630 .8732 48v. 20.00. .6,0D708 158.32600 .8732 504 21 00. 6.47107 157.36430 .8732 S28. 22.09 6.38901 1D6.44660 . E72 !

552, 23,.00 6.310G7 155.Se950 .8732 j s

S?6 -24.AO 6,23D4V 154.72Y8,? .8732 i eaOO . 25.00 6.14353 153,92510 .8732 l 624. 26.00 e.A9449' 1D3.15310 .873?

643, 27.00 6.02820 152.41180 .8731 o 7 . N ,00 5.96451 151.69950 .R7:2

696. 29.00 5.90327 151.01470 .8732 720. 30.00 5.84435 160.35580 .8732

744. 31.00 G.78764 149.72160 .5732 768. 7"N M) 5.73302 149.11080 .8732 7 9.. . .a3. 00 - 3,68039 148.52230 .8732 bi6. 34.60 5.e2967 147.95510 .8732 84O. 35.00 5.58074 147.40300 .8732 GM. ?h.Op .L.53355- '146.88020 .8732 698. 37 W- 5.48799 146.37070 .8732 912. 38.00 5.44400 145.87880 .8732 436. 39.00 5.40131 145.40360 I .S'732 i I

960. ~4O.00 D.36044 144.94440 .8737 98 t, . .41.00 5.32073 14/.50030 .8732 100E- . 42.00 L.28232 144.07080 .8732 1032. 43.00 5.24516 143.65520 .8732 leb6. 44.09 6.20918 143.25290 .8732 10E 45.00 D.17434 142.86320 .8732 1104.. 4L.00 5.14057 142.48570 .8732 1126. 47.00 5.10785 142.11970 .8732 115' .J , 48.00 D . O'7611 141.76480 .8732 1176. 49.00 5.04532 141.42050 .8732 1200. 50.00 5.01543 141.03630 .8732 1224. 51.00 4.98641 140.76170 .8732 124B. 52.00 4.95822 14O.44640 .8732 1272. 53.00 4.93082 140.14000 .8732

~129e . b4. : O 4.9041E; 139.84210 .E732 1320. SS.00 4.87826 139.55230 .8732 134'. S e. . O ? 4.853v4 139,27020 .8732 1J.64 5?.00 4.82848 138.9956& .8732 1392. b8.00 4. E,04 56 138.72810 .8772 1416. 59.00 4.78125 138.46750 .8732

1440. 60.00 4,758D3 138.21330 .8732 1464. 61.00 4.73c37 137.96550 .87Z2

'1485. 62.00 4.71474 137.72370 .8732 .

I fd ., - ,.'.s:. a..~ ..- n r..,u ..- : . , ,g;. 4 .

[ ',, 1U36.

.15's n 64.00 6 0 ., O 'i 4.67302 4.65297

'137.25710 137.03190 8732 8732 1564. 66.CO 4.63319 136.81170 . 6732 l bor . 67.00 4.61394 136.59650 . 8732 1632. 6P. 00 4.59511 136.58590 . D732 16M. 169.00 4.S7669 136.17990 . 8732 1(90. 7D.00 4.03860 135,97820 . 07 32 i104. - 7.1.00 4.54098 135.*/8069 . 8732 17 2 A .72.00 4.52367 135.58710 . 8732 i 7 % ,. 73.00 4.50671 135.39740 . 8732 1776. 74.00 4.49008 135,21130 . 8732 1800. 75.00 4.47376 135.02890 . 8732 1824, 76.00 4.45776 134.84990 . 8732 1 1

lb48. 77 00 '4.44205 134.67420 . 8732 l 1E72. 78.00 4.42662 134.50170 . 6732

-1896., 79.00 4.41146 134.33220 . 8732 l b. . EO.DO 4.!V658 134.16570 . G732 1944. 81.00 4.38194 134.00210 . 8732 m 1969. 82.00' 4.36755 133.84120 . 8732 ,

8732 ~d 19Y; , 83.00 4.35340 133.68290 .

2016. 84.00 4.33948 133.52730 . 873 l 2040, 85,.00 4.32578 133.37400 . 8732 20e4. SL.03 4.31229 133,22320 . 6732

.:05 ,, 87.00 4,29901 133 07470 . 8732 2i12. 80.00 4.20593 132.92040 . 8732 )

,E l 36 89.00 4.27304 132.78430 . 8732 2160. 90.00 4.26034 132. e>4 230 . 6732 2184. 91.00 4.24782 132.50220 . 8732 cit.w. .. ,, m .- ,. -

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, .tvar. . .u . c 1..

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2232, 93.00 4.22330 132.2281O . 8732 2256. 94.00 4.21129 132.09360 . 8 7 3',-

.280 95.00 4.19944 131.46120 . 9732 2304. 96.00 4.18774 131.83040 . 8732 2328 97.00 4.17620 131.70130 . 9732

. 6 2 ,, 96 Os 4.164b0 131.57380 . 0732

??)6. 99.00 4.15359 131.44790 . 8 73 240v. 100.00 4,14242 131.32360 . 8732 2424, 101.00 4.13143 131 20070 . 8732 244 R 102.00 4.120S9 131.070'.50 . G732 2472. 103.00 4.109E5 130.95930 . 8732 2496. 104.00 4.09924' 130.84070 . 8732 2ho. 105.00 4.08875 130,72340 i . 8732 i

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,- . . . . . . . . . ._;. a...... . . . . . . . . .

- . .a_.,--.

Ver rnant Yeail:ee F ar t II 18 Mon 16/18 bat.ch No Bun Decoy Hest Cuntm Heat Cumm M' Bun is 1 ' 1 ~', 2 1.76039 1.76039 132 .27 2 128 .40770 2.16009 260 1.67 3 132 .26040 2.42849 392 3.07 4 125 . 20~/ 7 4 2.63624 520 4.47 5 132 .19702 2.83326 652 5.87 6 12E .18209 3.01535 700 7.27 7 132 .18085 3.19621 91? 8.67 0 128 .36940 3.36561 1040 10.07 9 137 .26989 3.53450 1172 11.47 10 12'8 .15D36 3.69207 1300 12.07 11 132 .15792 3.85079 1432 14.27 12 116 .15736 4.00815 1568 15.67 13 t20 .15427 4.14242 1688 17.07 Ar turnpi 1 on ti :

Poweer per Dundle - 10.44197 M51(1/Hr 1 r r e d i i.it i or . T3me = 350 c O ,. Hours Re.sc t or Power = 1665. NWLh Ti me After Shutcown n 2400. Hours or 100.00 Days heference. IJURE Gr- 05v0, S t. aid e r d h e v a t.. w F1en (1) Section 9.1.'

Spe r,t Fuel Fool Looling and Clean-up Syst emn, Rev. :

(2) Branch Techr.ical Position ASD 9-2 Residual Decay Ener gy f or Li ght-Wat er hoactors for Long Term Cooling, Rev. 2 Spent Fuc1 Pool Cooling System Service Water S yst ern Outlet Temperatur es (F) 112.93 105.67 2n1ti Temperatures (F) 131.34 (Pool Temp) 85.00 Wet er Flow Rates (Mlb/Hr> .22o .175 Heat Enthanger Surf ace Area (SqFt) 500.0 Heat Enchanger Overell Conductance 375.0 BTU /SqFt-Hr-F The two pass correction factor, F= .873

__________.___________9

L

,; , ;;.r r;,ri - , , ,,m _ a ; 2 4 7- - . . . , ,.

~., ,

'.br roon t Yankee Pert II 18 Mon 16/18 1:

1

~

Dc.c s y 13 ms Decay- Heat Pool Temperature F fact.or l

' iHour s ) ( Day s ) (MBTU/He) (degrees F) 1

. . ...~_ ____ _____ ___._ ____ _ ____ ___ _ _ _ _

2400. 100.00 4.14242 131.34090 .8734 2424. 101.00- 4.13143 131.19490 .8731 )

2448. 102.00. 4.120b8 131.08130 .8732 <

2472. 103.00: 4.10985 130.95870 .8732 2496. 104.00 4.09924 130.84090 .8732

-2520. 105.00 4.08875 130.72340 .8732 2544. 106.00 4.07838 130.60750 .8732 2S68. 107.00 4.06913 130,49280 .8732 259;'. 1 Orb 00 4.05798 130.37940 .E732 2616. 109.00 4.04795- 130.26710 .8732

.2o49. 110.00 4.03802 130.15610 . O'/32 2'64, 6 111.00 4.02819 130.04620 .E732 ;

,2698. 112.00 4.01846 129.93740 .8732 I

- 271' . . 113.00 4.00884 129.82970 .-8772 2736. 114.00 3.9Y931 129.72320 .8732 2760. 115.00 3.9998/ 129.61760 .8732 2764. Il 6. 00 3.980D2 129.51310 .0732 2808. 117.00 3.97127 129.40960 .8732 2352, 11S.Ou .a.96210 129.30710 .8732 2856. I19.00 3.c?5302 129.20560 .8732 2020. 120.00 3.94402 129,10490 .8732 2904. 121.00 3.93511 129.00530 .8732 2 MJ . 12 00 3. 42d8 128.906DO .S732

-2952. 123.00 3.91752 , 128.80860 .87'22 i 2976. 124.00 3.90385 128.71160 .8732 i 3006. 126.00. 3.90025 128.61550 .8732 CO24. 126.00 3.89173 128.52020 .8732 128.42570 .8732 I 3048. 127.00 3.88328 30/2s 125.00 3.87490 128.33200 .8732 3096. 129.00 3.86660 128.23910 .8732 3120. 130.03 3.85836 128.14700 .8732 3144. 13).00 3.85020 128.05570 .8732 3168. 131.00 3.84210 127.96520 ,

.8732 3192. 113.Ou e. 83407 127.87530 2

.8732 J 3216. 234.00 3.S2610 127.78630 .8732

"' ? 4 0 .

. 1 E. ( O 3.81820 127.69790 .5732 3264. 136.00 3.81036 127,61020 .8732 3 3288. 137.00 3.80258 127.52330 .87~2 l 3312. 138.00 3.79487 127.43700 .8732 3336. 139.00 >.7E722 127.33140 .8732 30eO. 167).60 3.7796? 127.26650 .8732 3384. 141.00 3.77209 127.18220 .8732 3408. 142.00 3.76461 127.09860 .8732

'3432. 143.00 3.75719 127.01570 .8732 3456. 144.00 3.74983 126.93330 .8732 4' 3480. 145.00 3.74252 126.85160 .8732

.3504. 146.00 3.'73527 126.77050 .8732 3528. 147.00 3.72807 126.69000 .8732 3 3552. 145.O() 3.72092 126.61010 .8732 3576, 149.00 3.71383 126.53080 .8732

, _ . . . . , _ , , . _ . . . . ~ . , , , ,

., . . h. . .. :: + .; .- - . - --

- ; ~ .c Vermord Yan kee '12 Mont h 11/12 Cycle L . .Betch No han Decay Heat Cuimn Heat Cum:n L 4i Bun Ts 3 101 .10733 .10733 104 19.00

/ -108' .10908 .21640 212 19.90

l. '3 120 ,11862 .33502 002 20.80 l 4 80 .07739 .41242 412 21.70 L 5 96 .09090 .50331 508 22.60

-6 96 .08896 .59227 604 23.50 7 64 .07610 .66945 68S- 24,40 8 136 .12071 .78917 824 25.30 9- .329 .48493 1.07410 1152 26.20 10 Jo .02G51 1.09961 1'102 27.10 Assumptiorm j

n. .-...n-..-

Power. per . Dundle = 15.44197 MBTU/Kr Iri. cus 2 a t i on lira u 32412. Hour t:

heac t or Power m 166D. MWth i Time Aftnr Shutc!own =166440. Hours or 6935.00 Days

. Ref erence: NUFEG-OSOO, Etandard Fevit w Plan

( 1 )- Section.9.1.3 Spent Fuel Fool Cooling and Cleen-up Systems, he.. '

(2) Bratich Technical Posi ti on AEB 9-2

.Rcsidual Dec ay- Ener gy f or Li ght-Water Reactors for Long Term Cooling, Rev. 2 Opent F uel Fool Cooling System Be.r vi ce Water System.

I Cut 1 t;t TOf14r t/r 0 L ur ces> (F i 72.41 91.28 Inlet Temperatures (F) 97.30 (Pool Temp) 85.00 Water Fj ew Rites (M1b/Hr) .225 .175 Heat Exchanger Lurface Area (SqFt) 500.0 Heal Enth, anger Over ell Conductance 375. O DTU/9qF t-Hr -F g The two pass correction factor, F= .873 s l

l 1

l

~,.m..

_ _-_.____m_ _m_-_____-_--_-______-m_ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _

a, . , , . , .

W r enor J. Yeni.ee 12 Honth 11/12 L yt.l u vu 4 .1n- De c .> y imd Fuel Tumpterature F ir or (Hw e ti />s <> 01ET U / He ) (cegtces F)

,i+.*+++ +-++++e C 1.099e1 97.30120 .8734

,, o + a a #o+u '1 1.09954 '

97.29427 .9731

++ + . + , - ++.+++ y 1.0:i946 g 97.;9353 .e 32

u .e + **n+* % 1.09939 $ 97.29404 .G732

+ , + * + + ++o +4 ,, 1.O?932 < 97.29346 .E732 o*u* oew+ si 1.09925 e 97.29253 .8732 n++++ ++.+v+ fa,1.07925 v 97.29179 ,073:

oo+ ++-*_+ ,3 1.09910 ely 97,29099 .E732

+w+2 >+ +-+ 4 , y 1, c.v c;O3 ,q r?7_2ccle , E- -

t n ***v ***w ,f 1.09896 , p- 97.28938 .3732 4,,+++4 +x.,r 3 1.O m v <L 97.2bE57 . cG O.J

. #. e, ..<e .

v .; 1. 09ET t1 9/.26776 . E,7 7

, + 4. + ,c 4 #o~++ ,y 1.098ia c7.28696 r .6132

+++ 4v , ++%. ti 1 098o7 91'.29615 .377"

+ +&x,+ + , + , * , s, 1.05800 9/.2853S .ET

++s- * * -. m ,, y . Oya; ; 97 28454 ,a732

+*wg," +""&> w 1. OW4 6 97.283?? .H7: -

++ +a- 4+ r u , o 1.0953F 97.28293 . 8 71.'

+*v.,, .+ s w 1. 0 :, E 3 i ' , . 2 8. 1 ' .mi o+++, .+-n, u i.09524 v7,201% .8732

+ , , * + .. >+,-, a 1.09617 ;7. Jer 2 i /3;

+**++; w**++ o 1.09810 97.27971 .573-

++# +w ,# +. , y 1. O'.,0 0. 97.27e90 . L ' 32

+++**m ++++w 6 1.097o5 37,27910 ,g772

**+++ **a++ n 1.09789 97.27731 . EO 32

+**ot +++w& 1 097G1 97.27649 .8732 t u ,: a+ .++.*+ ss1.09774 97,27569 .d7; wo) ---++ 1.09766 97.27459 .8737

++++4 ++xn+ yy 1.04759 v/.27439 .E73/

o+o* w + .. v n j,0975? 97.2732c .8732

+ +n,+ ++++e h 1.Ow745 97.27248 .E732

+*++n , . * , + , + .09738 97.27167 .3732

+ ++.*4+ o.+++ Jy 1.09730 97.270S6 .8732 wu*u- m~+ 1.09723 97.2700n .5732

++ + + . - + + t +- + o ya1.09716 97.26926 .8732 p p p ,4 er p- . . + s 3 ,, ( . y ; n,9 ~.-

r .. -

~1 l . ,,,, C; c.,. Q. r,y g

.C._..-.

j 1

n+++* +*+u+- 46 1.09702 97.26765 .8737 :

o+*4 **.,o 1.09694 97.26685 . G73: i

e*n ++++++ vv1.096S7 97.26604 . E73: I

{

wn+* e*w++ 1.09660 97.26523 . 8732 i n - , . . . l s + , , :. , ++.,> ..

g i . .. - -ni. . , - 1. . .n , tA... . t- ? .. ,

1

++++** *****+ 1.09666 97.26363 . 8732 1

++n++ +++ *+w yy 1. 09659 97.26292 . 8732

+>*na *u*++ 1.09651 97.26202 . 8732

,o** no*+ ro1.09644 97.26121 . 8732 i

- - +w no++ 1.09637 97.26042 . 5772 )

++++++ +++++F f'1.04630 97.25960 . E732 )

- - - n *****+ 1.09623 97.25881 . 57"2 j 1

o+a+ +wo+ rY 1.O?615 97.25800 . E72

]

- +**+ * * + *

1.09605 97.25720 . 8732 l

+n#++ ov++ , rt 1.07601 c;7. 25 639 . 57'S 1

+**+++- **oi. 1.095;4 97.25559 . 0722 i

+n n +w+s eP1.OY587 97.2509 . E032

++,*** *+++ . 1.09580 9'/.25399 . 8732

++#+++ n*o+ so 1. On572 97.25318 . 873T l l

- - - - +++++-- 1.09565 97.25238 . S'732 i j +u+++ +++*n 6wi.0955G 97.25157 . 8732 l

l~

L _ _________-__________- - __-_ - _____~_ _____O

. _ _ . . ~ . . . . m..,.,...

a . . 1 ., . o u .t -

' ~

i;.s.zura -- ~ . a r : ,.

+c+o++

+

++++++ <,v1.09544 97.24998 .E732

A>a* * * *

1.09536 97.24917 .6732

++r+- +++ +; s.41.09:29 97.24537 .S732

k***' "**** '. 09522 97.24755 .8732

++t.++ +-++++ c, P l . 09 51 D 97 2el677 .8732

,. . ~< + ~#.=> ;,09nop 97 . :, a 39 (, , grp

++,.44+ ++v+** ,o 1. 0c. 501 07.24516 .A732

+ .x -. .. u *.-s,-+ 1. O t; r,5 3 97.24435 .8732

+e . + 4. +: H +*++ ,w1.09496 97.2435S .0732 u ,, 4 m. *

o +: + 1. 0 9 e; 7 9 97.24274 .0732 '

+++,++ n+++4 391.09472 9'/,24144 .E732 e+++* "+++* 1. O'i4t$ 97.24115 . 8 7 "'m~

+ o +.+:+ . ++++** 1t1.0942 ~;7. 24 035 .87:'.

e, x . , ., ++3.+,+ 1.094DO 97.23954 . 8732

++<+n +++m pr 3.09 34; cr.23974 c ,8732

.* w w e n+*** 1.09436 97.23793 . 8732 i

++ *n+ <++ +4 to 1. ra ',.>,c- c<7. 23714 ,. 6732 ;

+ r. s + ~ , , . . , . - , ,

,, . - + 1 . , . .,. 9t 9 / . 4a.v;, . c;, f .;.t !

4++++, ++++5+ y s.1. 09/' 14 c7.23S53 . Dyr

++3o +-+++++ 1.Oc407 r 97,23473 . 873;_

+*4.+,* + + + - + , 1.09400 97,23 H2 . U/at

++++*v +k+&W+ i[1.09393 97.2331[ . 873," ,

++n++ +o+++ 1. Oc/3 b L 9/."3232 . 87aa

- +o, * * * * , . + 1,09379 97.23152 . G732

+ + , . . . - + ++,.++,e 3.n937j 97,23072 . 873:-

+e++*r ++ m + 1.09564 97.22992 . a732 ;

+ %-4+ n+++4 1.09357 97.22912 . EO :;

om 4 +-,* r + 1.09350 97.22831 . 5732

++o++ +++**+ 1.093a3 97.227Dj . 0732

- +.+. * *o+,+ 1.09336 97.22672 . 8731

+ + + + -+ +++*-- j.09??8 c;7.21591 . A772

- ++,* w+-+++ 1.09321 97.22511 . G732 l

- - - + 1 *&+**+ 1.09314 97.22430 . 6732

+*,-*s. n n+ + 1.09307 97.22350 5732

+oo+ ++++++ 1.Ocn y, 97.22271 . 877:'

+u+** ++ **+ 1.09293 97.22190 . 3732 q

++,-4 *-++^- s oo 1. 09283 9'<.22110 . E 3 :

++-#, + + - + , .

1.09276 97.2203.1 . 8732

+ S , 4. + ,. +++..+ 1.0927i 97.21951 . cO L.2 ,

I n.+.,* +-+-+++ 1.09264 T7.21B70 . 8732

++,.++ ++++++ 1.09257 97.21790 . 8732

+.++++. 4,.++* 1.09230 97.21710 I . 87T2 ,

1

)

i l

l l

l i

l I

1 l

1 1

i l

t L

. . ... ~. .w , r . . .a ; .w w v . . :.- - - . .

Ver mont Yankee Part 11 12 tion 11/12 Be t c h. No-Bun Decay He.st Cumm Heat Cumrn # Bun Ts 1 104 .10662 .10662 104 19,27 2 108 .10837 .21499 212 20.17 s 120 .11784 .33283 332 21.07 4 80 .076D9 . 0972 412 21.97 S 96 .09030 .50002 508 22.87 o 96 . OSB"J.6 .58840 604 23,77 7 84 .07368 .66409 688 24.67 k 136 .11993 .7E401 824 25,57 9- 328 .28307 1.06709 1152 26.47 10 To. .02554 1.09243 11 H;c 27.37 As s. ump t i on s,:

Fower per bundle r- 15.44197 NBT U / rir Ir r acji ati on 11 w " "2412. ' Hours Reac t or F oe>ut -

1660. MWth Time 6.if t c" Sheldo>n -169D40. Hours or 7035.00 Days Fmi t r enc c - NuAEG-%oO, St alid ar d Revi ew F1an (1) Buc. tion 9.1.3 9pcrit Fuel Poul Coclinc ant! Clean-up 5ytt emn , Res. I' C) E,r a n ch Technical Pocition ASB 9-2 he"s i du al Occ.a y Ener yy 4 or L2ght-Water React or o for Long Term Cooling, Rev. 2 Spent F ver Fool Cooling bystem Screitt' l'at er S ys t ern 8

Out l et T etopc t M e* er (F) 92.37 C 1. 2 'l j 1 r. ) ce l Tetigcr o t.ur es (F) 97.22 (Pool Temp) 55. ::'O i i,

Weter F1ow Rott.s (Mi t;/Hr ) .22D .175 Hea+ Erchanger Eurface Area (Sqrt) 560.O .

Heai Exchanger Oserell Conductance 375. O BTU /SqF t-Hr -F )

Thu t wo pew correction f ac t.or , F = .873 ll l

i i

l l

l-l l

l

___________.___________m

.. . .. . . . . ..; a . = :. : : -

Ver nrarit Yardee Par t 11 12 tjori 11/12 Dc ca v T ring Decry Heat F'c o l Temperature F fetter (Hum e) (Days) (ttBTU/Hr ) (degrces F )

+ + + + + , +*+++t rob 1.09243 97.22086 .8734

- +s. .

- - +++ 1.09235 97.21397 .8731

++*+++ +*+*++ 1.09223 97.21522 .0732

,.+n **+-n* 1.09221 97.21372 .87.32

++++++ m +*+- 2.09214 97.21316 .8732

o .1 4 #-n** 1.09207 47,21228 .8732

+*,,.*4d *+++*+ 1.09200 97.21150 .8732

*

w r- oo** 1.09192 97.21069 .8732

++ f. + + n+ m j.09isp 97.20990 .8732

,,+4 s ++++** 1.09178 17,20909 .8732

+n+,a *+++++ 1.09171 97.20830 .8732 ;

+nxa ****** 1. 0916il 97.20750 .873< l

,5++v w+++t 1. O Y 15'/ 47.20670 . 87 3:- l

- +i** *+ + + 1.09149 97.20099 .8732

+u*++ +++++* 1.09142 57.20509 .3732

- ++o *+M+* ,,g ,1.09135 97.20430 .8732

+*,.-**+

w a v +.- 1.0911a 97. 20v 9 .9732

- - - - + *cs<+ 1,09:21 97.202'71 .8732

++++u *++ne i , Oi/111 c;7.20190 .9732

o- **+*** 1.09107 97.20110 .8732

++nts m +w 1.0cO99 e 97.29029 .8737

- - m +: *

u t 1.090 % 97.19'750 . 87%

+ n+++ mo+ j . 090Ei, 47.19870 ,8732

+++#+w ***u* 1.09078 97,19791 .8732

++x+*k **nH 1.09071 97.1971; .8732

+w**C ****** 1. O'Y Oi,4 97.19630 .8702

++*--*+- +++++* 1.09057 97.19550 . 83 2

- - - w +*+*++ 1.09049 97,19470 .8732

+#+,.,+ n,u+ 1.09042 97.19391 .8732

+***** **+ n + 1.09035 97.19310 .8732

++**++ +-+++4 /30 1.09v28 97.19231 .8732 m +* ***+++ 1.09021 97.19151 . 8 7-' 2

+*n ,* 4w+++ 1.0Y014 97.19070 .8732 non *+u+a 1.09007 97.18991 i .9732 i

- - - - + **a.+++ 1.089'9 97.18Y11 .0732

.+ *** +4 + n + 1.08992 97.19831 ,9732

+.++,.+ ux+e 1.Ouv80 97.18751 .8732

- - - - H **** 1.08978 97.18671 .8732

++*v++ *++-*++ 1.08071 9/.18D91 .8732

+oo + w**+ 1.08964 97.18513 .9732

+ +,

u + + o +++ 1'.08957 97.19432 .8732

.**,+, **m+ 1.03949 97.183S; .8732 o***+ ****u 1.08942 97.18272 .8732 ]

L

u*** ***++* 1.089'".5 97.18192 .8732 1 1 4 u+++, n++++ 1.08928 97.18112 .8732 1 o**o **w*+ /v r 1.08921 97.18034 .8732

- - - ++ *++n+ 1.08914 97.17953 .8732

- - *+ ****++ 1.C8907 97.17873 .8732

+ou+ w++++ 1.08899 97.17793 .8732

++***a *++**+ N9 1.05892 97.17714 .5732 l ,

i l

l l

l

I

//

gy '

. . WT-' ~ ~ - " ~ ~

- f-< m " ~

gL 7 Ver mont Yankee- SFP E.npanei t,n 11/12h16/16 Spsnt Fuel' Fool Cooling Syctem Service W ter System 1

' Outlet' Temperatures (F). 153.31 142.89 Inl et ?l emper'atur en (F)' 198.33 (Peol Temp) 85.00

- Wa ter F1' ow Rates (M]b/Hr) .'225 .175 Decay Heat Land . (MBTU/Hr-) 10.131

~ Hm-.t Enchanger Cur 4 ace' Aree (SqFt> 500.O Heat E~nchareger - Ovc-et11 Conduct ance 375. 0 Btu / 5qF t-Hr -F The-.two pass cor t ecti on factor, F= .573 Ver rnent Venkee SFP En pansi on .11/12b16/1R Spent Fuel Pool Cooling System 5arvice Water Syst?m Out1et .Tempeetur es &) 1150. 50 1/10. 5 6 In) et T r rnpr:r +tt ur rz e (F) 193. 7.~ (Pool Ternp ) 85 CO Water F1 ow hEtc+ (M) h/Hr ) ,223 575 Ducay Heu.i Lc.ad- (MOTU/H1 ) 9.723 Huat Exchenque bur + ace Aresa (SqFt) 500.0 Hutt E% hnnger Dver4J 3 . Conductant e 370.O btu /sqFt-Hr-F 1he tWo pw.n correct 1un factor, F = .873

'Vernient Yenkce PFP Expc'nsia: 11/12h16/19 Spent Fuel Foal Cooling System Service Water.Syutem

' Out l ui- ledpersturee W) 148.34 133.73

'Inl et Teenpdr et ures - (F > 190.13 (Pool Tevop > S0.00 Water.F1ci i R e t. -- . 411 b / Hr ) - .225 .175 Dece.y Heat L.o e ti (NDT U/Hr ) 9,402

. Hunt.Exchene;er,Surfece Aree (FqFt 500.O I Heat Exchanger ' Dver c11 Conductance 375.O Btu /SqFt-Hr-F The two p me corr ection fmctor, F' .873

% rnant. Yenkee EFP Enpt,ncion 11/123<16/16 Epent Fuel Pool Cooling System Service Water System Outlet Temperatures - (F) 146.50 137.23 Inl et. Temperatures (F) 107.20 (Pool Temp) 85.00

.Woter F1ow Rates (M1b/Hr> .225 .175 l

Decay Heet Load (MBTU/Hr) 9.140 Heat - Enchanger Sur it.co Area (SqFt) 500.O Heai Enchanger Ovet v11 Londuttante 075.O Btu /54Ft-t ir -F Th r- 1.wo pass correction factor, F = . S

l.

l l

l' .. ... .

4 + . . .~._w. .~w-,.

=~ .

, n . - . .

l l

j

l. .

W: r nen t Yent:ee SFF Ex p an t,i on 11/12M 6 /19 Spent Fuel Pool Cooling System Service Water Sycten l

\

0 ie l et icmperaturos 4~T 145.10 '.35.98

.N et . e,aaer atur c r ( T- ) 184.75 (F ool Ten p > 85.00 Wat er F]cw Rates (hlb /Hr) .225 .175

1. cec o y Heal Loed (MPIU/Hri L4. 921 Hest Er.chenger surfete aree ( 5 ;jF t ) 500.O Ncot Eattmgu Det r e]1 Cc nduct ance 375. 0 1:t u / 9 qFt -Hr -F Tne two pass cor recti on factor, F = .873 W'r@ont Y a r. n e., EFr' W pension 1.1/1 2 M 6/$ 8 5 p t.n t F u .21 Pool Coc12ng Sgt. tea F er s i c e Walt. r S y c t ru i Lutlel l emp ut w tur n (F) 143.81 134.E7 In1ct remperatureo (F> 192.61 (Pool To rip i ES.00 Watcr Flou F .4 t e t 'M] n / R- ) .225 .173 Dec. 7 y H. r t unac ( I F L', W i

. G.730 Heat E> chteigt r iurfcee Area ( 5c;F t )

. SOO.O Haai & ( hang et n y,4 - Il c c,n d t " . t m ic e 37 D. 0 13t u / SqF t-Hr -F The t wa pc e correction factor, F = .87~

Ver uut t Yo '.ac SIP Cu panti on 11/1281c /18 Spent Fuel Pool Cooling System Serv 2co Water System J

b c L .L u .' e op a r t u r e s tF 142,72 133. %

In]et l emper e t ut ( v- (r' 180.60 (Fool Temp) ES.00 Wa ter F .L otv rw t e o (M1 b /Hi ) -

.225 .175 Decay H M L occ: (hbTU/Hi) 6.568 Heei Er c h cu ag e r suriete Area (SqFt) G00.O Huut Euchengei Ovt r ell Conductance 375. 0 14t u / SqFt-Hr - F Ti m tu pecc c or r 1 etion fortor. F' .873 Ver munt Ventrec SFP Eapansion 11/125.16/18 Spent Fuel Fool Cooling Sy=*em service Water System Outlet T emp erat ur es; (F) 141.75 133.12 Inlet Temperature (F) 179.16 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Datdy Heat Loed (MI:TU/Hr > E.421 Hee.t. Ex changer Sur f see Area (SqFt) 500.O He.it En t hanger Over all Conduc t anc ~~-277.66 (Pool Ternp ) 85.00 ,~~

Water Flow Rates (Hib/Hr) .225 .175 Decoy Heat Load (MD f U/ Fir) 8.287 Heet Erchenget- Sur f acs Art ( S q Ft. ) 000.0 He.ct F.n cha g er Over td } Concluct ante 375. O Bt u /EqFt-Hr _F The two pase correction.fector, F. = .873 Ver n:ont Yen t;ee FFP Enpant i on 11/12616/18 l- ,

~~Spent rue; Prol.Cuoling Syt. tem 5er vice Weter Sycteni I>~~

l i . Ou tl el. l empuc at.ur t t . W) 139.99 131.65 Initt 7 t roper ai ne' ve. (F) 176.27 (F ool T erop ) 85.00 Water Flow twtos (M1b/Hr) .225 .175 l .Le e ny Hu L Louc: ' MHT U'il+ ) 5.J65 Heat E>: c h se g er buriace Aiw (54Ft> 500.O Heat EPch mger Det r all Conductance 375. O Blu/5qFt -Hr -F

~~j. The two pees correction factor, Fu . 37;.~~
l Wr (M 'nt Wrri.ee EFr Enpencion 11/12F416/10 h
Spent Fuel Fool Cooli ng Syt. ten Service Water System
. ._ _ ....__ _ _ ...._____.._ _ ~ - . . . . _ _ _ _ . _ _ _ _ . _
l l
\
l QutleL lemperaturta Wr 139.23 ~ 100.99
.initet Temper si ur ee ( F 's 175.00 (Poo) Temp) 85.00-Weter F1 ad liates W U s Hr > .225 .175 Decay Heat Loed (MDTU/Hr) 8.049
~~Heal Exchan pr SuNecu'Aree (EqFt> 500.0 i~~
~~3 Heat Exchanger'Dverall Conductance 370. O Btu /5qFt-Hr~~
' The two paen..correcti on fector, Fa .873 Vermont Yani:ee SFF Expansion 11/12016/18 Spent Fuel Pool Cooling System Service Water System Outlet Temperature (F) 138.50 130,38 ,
L Inlet temperatures (F) 173.80 (Pool Temp) 85.00 l
Water Flow Rates (M)b/Hr)' .225 .175 Dec.sy Heet Load (MBTU/Hr ) 7.942 !
. Hest Exchanger Surf ace Area (SqFt) 500.0 l
.Hi-et , Exchanger Over all Conductance '375.O btu /5qEt-H -F The two pass correction fettor, F = 873
o ,
. , . .:a ;; . m '. . '" ~ " -
.. ..> .. :v Ver mcso l. Yenket E P 'Eu pe,nsi on 11/12&l6/1E Epeist. Fuel Pool Cooling System Service Water System
'Out J ei -. Temperatures (F); '137.82 129.81 In)69 l emper:atur ew IF) 172.67 (Pool Teng ) 85.00 ele..M ow Fbtes '( M1 b / H- ) .22'3 .175 Decny He-et . Load j (tPTU/rir ) .
7 841 l Heet Ex c list.ger Sve. f acs Ar e4- (5qFt) 500.O l Ht id En c, hanger ' Ove al1 Conductarice- ' 370. 0 Bt u/EqF t -Hr -F The two pees correction f actor , F n. .673 l Um mor;L Yerekte BFP Enpansion 11/1316/18 5 pent' Fuel Pool Cooling Syctem Service Water Systerr !
I
---..- ------..-__-.-- -- -- - -.- --~~- . - ----.. ;
I 1
Outl et . Ymper etur em . (f-) 137.18 129.26 !
.lniet Tengeralerev (F)' 171.61 (Poo) . Temp) 85.00 .,
1 I
We t er. Flou iietes :hl'J /Hr ) .225 175 '
.Dec h Hent Locd ( tF:~i U / Hr ) 7.746 Hest . E x t.henger Suri et.e Area (5qFt) 500.O
Hent Exchancier Overed 1 Concluctance' 375.O Btu /SqFt-Hr-F The two: pawn correction 4 ect or', F = .873 Ver rnoni Yankee SFD Expension 11/12&1.6/19 Spent Fuel Pool Cooling System Service Water System
. OU t let Tt u ipero tu% 0> 136.D5 12 8. ~*5 101et Temper A ures (F) 170.60 (Pool Temp) SL.00 W,e t e> Fl ow iMe$ blis /Hr ) . "' 20 .175 De:c a) Hect Lasci (MBTU/Hr) 7.656
Heat E9.i u n n

> Sur f ect Are. (EqFt) 500 Q Heitt Enchange: Omr al l Conductance 375.O btu /SqFt-Hr-F Tne t wc; p a s.c c c.;- r ect a an factor, F= . 8 7-'
Ver mont y'ard:ec SFP Expontien 13 /1M:16/1E 5 pent Fuel Pool Cooling System Ser vice Water Systera i
- Out l et . Temperature =f. (F) 136.00 128.26 Inlet Temperatures (F) 169.64 (Pool Temp) 85.00 Water Flow Rates (M]b/Hr) .225 .175 Det. a y Heat Load (MBTU/Hr) 7.570 Hest Exchanger Surface Aree (SqFt)- 500.O heat. Exchanger D era;) Conductance 375. O bt u/5qFt-Hr -F
,The two pass cotr.crtioie f actor , F = .873 l
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~~. , Vet.tont . Yankee 5 P Expansion 11/12&16/18~~
, cpent Fuel Poc1 Cooling System ser vice Water- Sy: tem
'- ;Ooklet'ismperatures IF) 125.44 127.78 1, J u t 16 taper at u? E b W) 160.71 (Pool Terap ) 85.00 6tet -Flow' Rates (M)b/Hr). .225 .175-DstetyL Hest Load L(MDT U/Hr > 7. 4 F7
-). eve E>: changer . Surf see At ea (SqFt) 500.O Hetil En hangnrr O mrall Cunductancu 37b.O Bt u/SqFt-Hr -F The two pen correction factor, F-= .873 Vier ttiari t Venken grep n p en ni an 11/12gq3/ip 5pont Fuel Pool Cool i ng Systea 5.wrvice Water Ey' stem G Uutiel T emperatur em (r 134.91 127.34 Ird et ~i e ~nptar t.t ar et (F* 167.84 ( F ool T em;-) 8S.60
. We.t er - F1 ou F.s t % (M1b/H-) .223 .170 Decey Hv e i. oi 0 ' U 'BT U1H,- ) 7.409 Heet Ex t. henge Surfwcv Area (SqFt) G00.0 Huat Ewt b.s ige- 0"e*all-Conduciante' 375.O btu /SqF1-Hr-F 1he t.wo pats cotrection faster, F= .873 Var munt Yt nLee ' FF"P E::pentsi on 11/12t 16.'1 S Spent Fuel Fool Cooling Syston. St?rvice Wetcr 5ystem
. -- -..-- -..--...~ - --.- - ---.----- --.----
Out i et .' i rmper r t w et @) 134.41 126.9.1 -k In]et T ensper at ur er. (F) 167.00 (Pool T einp ) ED. 00 W as eu Flow ~Re m ( M1 ti 'Hr ) .225 .175 Ducay HoeL L bat' 01DTU<Hr) 7. ~33 4 Het Ehchinger Sv f ir c e Arec (SaFt) 500.0 Heat Exchenau Over ell Conductance 375. 0 Dtu/SqFt-Hr -F The two ps.- co recti ,n fa: tor, F = .873 Verruont Wnkee 5FP E,:pansion 11/12h16/13 Spent _ Fuel Fool Cooling System Earvico Water System M
~
. Gullet'Temperetu es (F) 133.92 126.50 Inlet Temperatures (F) 166.20 (Poul Temp) E5.OO Water F1'ow Rated (M1b/Hr) .225 .175 Decay Heet Load (tTTU/Hi) 7.262 Heat E.achanger Surface Area (Oqrt) 500.0 Hest Enchai.ger Cver c l i Coiidutiante 37S.O Blu/54 Ft-Hr-F T h c: two pecs. cur r ection factor, F = .873
. m .m.
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. ri
_.____.___m _ _ _ _ - _ - _ _ _ _ . _ . , _ _ _ _ _ _ _ _ _
'1 y , . . a . .. . ;. . : 7 "' ' " ' ^ ~ ^ ' * %;.".M :: * *~:: * .. .u Ver ruut el W n t ee E F- P E.r. p an ui en 11/12D16/1h Spent Fuel Pool . Cooling SyEtem Star vi ce Water ' Sysl.em u..---...._.-...--__----.. - - . - - - - - . . . . . . - -
. Outlet Temperatures ( F )-
133.46- 126.10 Inaot 7 temperatures !F) 165.43 (F'ool %mp ) 85.00 Wei. t o-
. Flow Rates (M1b/Hr) .225 .1.75
. Dec'ay Heat Loed . (MbTU/Hr.) 7.193 1%et . Exchmger B u r f a c r- Ar ea . (SqFt > 500.0 He n'. h r.h tag er Dverall Cuiiductence 37D. O ~ blu/ SqFt-H~r.-
The t wd p.n s corcection factor, F,= .873 Vermont Yankee EFF Eupanci on 11/12t:14 /' t
' Spent Fuel Pool . Cooling System Service Water System Uut 1ka le o per et urus (F' 2.33.01 125.7 In 3 t t T t ml cr e.it to e. XF) 164.68 (F ool Temp) EU.00
. Wc tet Flow Wi i.u s (Mla/Hr) .225 .175 Decay Hez:t Loed (MT TU/lir i 7.126
~~Hear Ex changcr .6ut f c.: t.:e . Area (ScFt) 500.O +~~
Heen Eachoneer Uverv1) Conductance 370. O Blu/SqF t-Hr -F The-two r e. correc. tion factor, F= .873 Verner.! Y..: .n f ee FrP Enpenu un 11/12.% 2 6 '1 F 5 pent Fuel Pool Cooling System Service Water System Dutivt '1umpeieturet. 'F)
. 132.58 125.56
) ni et T emperet ur e+. (F)' 163.97 (Pool Temp) 85.00 Water F1N R ele a (M1 b /H!-) .225 .175 Irce.y Heat Lcyd (MHT L'/Hr > 7.063 ,
H e c. ' & .a ' es.q ce Sw f eu- Ar es (SqFt> 500.O Heal D c nenign Owrell Conductance 375.O Btu /SqFt-Hr-F The bya pe: e corr ection f actor , Fe .973 Veruunt Yenkee BFf' Expension 11/12&16/10 Spent Fuel Fool Cooling System Service Water Sys. tem
. Outl el Temperature, (F) 132.16 125.01 In) et Temperat ures (F) 163.28 (Pool Temp) EL.00 Water Flow Rates (M1b/Hr) .225 .175 l
Decay Heat Load (MPTU/Hr ) 7.001 Heat Enchanger Surfete Area (SqFt) 500.O
' Heat-Enchanger Overall Conductance 770.0 Btu /5qFt-Hr-F
.The t wo past correction factor, F= .873
.-___.______-__m__ -
, , , , .,;..~...-... y..g . ., ,,
.,- . p Weonchi Yenhet SFP Er peinsi ce , 11/12P16/18 Spcmt Fuel Pool Cooling S gtem- Service. Water System
- - - - - _ - - . - . . - - ~
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L Outlet temperatures (F) 131.77. 124.67 j inj e,4 l emper atur es . (F r 162.62 (Poc1 Temp) 50.00 wM ut 'Fiow Rates (M1b/Hr) .223 .175 In ay Heat Load OGTl.1/Pr ) - 6.942-
- Heat Ersche.nger Surfaco Arr.; - (SqFt ) 500.O Heat Ex cfienger Ove-eell Conduc t arite 37D. O T;tu/ 5qF t -Hr . -
1he two pass correctio1 fector, F= .873 Wr noon t Ytu i hre SFP Enpant, ion 11/12'.u 4 /10 Spent Fuel Fool ' Cooling System Service Water System uutlet Temper at ures (F) 1?1.02 124.03 Jid et T eniper t t t- es 'F) .
161.38 (Puol TempF 85.00 Wutor I' l oc. Rates (M10/Fr) .225 .175 bet a y _ H ..w I Load ( MBT U 'Mr ) 6.831 Heat Extheoger Luri (c e nrca (SqFt) 500.O Heat Ex c hancier Ove all Conductance 37U. 0 Fit u/SqFt-Hr --F
ihe two paws ccrrection factor, F = .873 Ver ment Wr+ec SFP Expant.ico 41/12M 6/18 Spent Fuel Fool Cooling System Ser vice Weter System Ou t l e.: . Tempe a$.orew ti 130.3.! 17.3.44 1olet Tenper at ur e-a (F 160.22 (F oul Tenp). 85.00 Wat o r Flow Rut es (!11 ta / H* ) .223 .175 Detay Heat'Lued (flBT U< br ) 6.727 Heat Enchangrr Surf act: Aree (SqFt) 500.0 i Heat Ex chang er Overell: Conductance 675.0 Btu /SqFt-H -F T hu t wo ' pac e., cc-r ecti on factor, F= .873 Ver rison t hnkee SFP Expant.iun 11/12Sd 6 /18 Epent Fuel Fool Cooling System Service Water Sy tem Out l et. . T emp er atur es (F) 129.67 122.89'-
Inlet Ten 4peratures (F) 159.14 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Deccty Heat Load (MBTU/Hr) 6.631
- Heat Exchanger Surfaco Arse (SqFt) 500.0 Hect Exchangt- Onr all Conduci ante 375,0 Dtu/SqFt-Hr-F Thc- two-past c.or r u c t i ori factor, F = .873
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W , n, .c L Yarmee EF F E.: p eru.i ur. 11/1: M a '18 Spenc Fuel Pool Coaling System Serv ; ce Water System 4
D..i .t. T emp er t. t ur e s, (F) 127.07 122.38 i n .) e l . Te m,3c r ator ce (F> 15~.14 (%ol Temp) 85.00 16 t er Flow Reteu <M10/Hr! .225 .175 Det < .y Heet Lotd (MF 1 U/ Hr > 6.541 H, : t Enchangrr Furfa<e Are; ( S q F 4. ) '00.0 Hc ai Eacht ran t Dver all Conduct ance 375. 0 Pt u /SqFt-Hr -F Ti e two pc.n c or rc ct i on f actor , F=
, ,873 W. r,,on t Yaiii:ee EFP Enpane.1on 11/12016/16 Spent Fuel Fool Coolang Eyat em Service Wat"r Systea Du t. l e t 1 en..p er u t e ir o u .F).
123.51 121.90 J r.1 et Temper oi ur es (f ) 157.21 (F col Ternp ) 85.00 Lu t or- Flow Rates (hlb /Hr) .225 .175 De c c. He e t L owJ ( ME T U .' Hr -
6.4DC Heat En cin..'icn r vtriece At eu (SqFt) SOO.O Hc st Erchange- Over all Conduttante 37ti. O bt u / EqF t -He -F li'e twu p <. s s correctioit f ector , F =
.673 Wr n:osil Yor i'et 9F F E < pent.i on 11/12fle/20 Spent Fuel Fool Cooling System Service Water System Liu t I e i. 7 enup t. r . L ur % (F . 127,97 121.45 I r,1 ( t Ten per e t u< e s @> 156.33 (Poul Ttmp) 55.00 W+ t er Flon hetee i M] u .'Hr . .223 .175 Deceq Hee.t Luto .MblU/Hr 6.379 Hos L Enclianur- br e sco Area (ScFL) 500.0 Heat Enchalige Overull C on d uc t eo ic e 375.0 Btu /SqFt-Hr-F The two p s ea torrection factor, F = .873 Wr nivi st Y e n i . e c. bFP Ex p an s.,i c:ri 11/ 2 2td 6/18 S p en t. Fuel Fool Cuolin.] System Service Water System Outlet Temperatures (F) 127.48 121.03 Inlet Temperatures (F ) 155.50 (Pool Temp) 85.00 Water Fl ow Rates (Mlb/Hr) .225 .175 Decay Hect Loed (MBTU A le ) 6.305 Hest Exchanger Surface Area (EqFt) 500.O Heat Exchanger Over al1 Conduct aric e 375. O Blu/SqF t-Hr -F lhe two pmss correction fattor, F = .873
~
o. - - -
h o,on t Yankee 5;P Expinsion 11/12b16/38 Spent Fuel Fool Cooling System Service Water System Uutle f erop er a t ur e n i. F s 127.02 120.64 li,i t ! lemper atur et (;> .154./4 (Pool Tenip ) ES.OO Water Fluw Ra t ero (M1b/Hr) .225 .175 Da o f Heat l.oed (MF TU/Hr ) 6.237 F4 .t E ,c c l . m n g e r Surfeco /ir t . e (SqFL) 500.O heat E c c h avigt - Om s11 Ctr duct t nu. '75. O E: L u /SqF t-Hr -F-ltw two past. c co' r e t t i cu i f actor , F = .873 Ver rm.,n t Y.u n it ee SFP Expansion 11/12616/1b Epunt Fuel Pool Cooling System Ser.' ice Weter Syster Ou t l ei. 1 e.op -n vi - (F/ 126.59 120.27 i ni t .l Temp. utw ee (F ) 104.02 (Fool T L>mp ) 65.00 We t ei F l an. R :,1 e - ., I n L / a r ) .225 .175 lite w Hue i owl ( ,11G d / Hr > 6.173 ,
Htet L xche nger c.ur4a..e Arca ( E q F~ t i SOO.O l He t- t E :: r. h a i e r O vi.-r d 1 Lonductente 375. O B. 2/ SqFt-Hr -F line two p&n= s.or r t-t t i on +ector, F = .872 V + p a n i t- Ye :l: ef t BC F ' [nparnion ]!/}2Dj6/19 Epont f uel Pool Cooliny Systeen St'r vi c e Wat er E,ys t en: 1
. - - . . . - - . - . . - . . - - - . . . . . . . . - - - . --- .-- --.-... - j i
l u .s t i e '. i ., ,
4# e 3 ilo.1f 1 15 . r .2 i i n) e t 1 e < .,4 i atur e- ti 103. ~2 4 (Pool Tt%p) E5.00 4 l
l W,LL F i o.. hett. , L ! . 'Hr .225 170 I Decoy rim cl _t-d (Ni:TU/ Hr ) 6.112 ;
H...t r.4 vny < v. . t -at. vw i. x,F t ) 500.C -
Heat Enthanger Over d 1 L t.nduc t ante 370. 0 Otu / St,F t-Hr --F 11.e twa p e, - c,rrectit a 3 ector F = .873 Vu to u d 't m a k ee EF P Expwiw2 uti 13/1 h16'18 S p e n ;. Fuel Pool Cooling System 5crvice Water S y s t en.
Uutlel Ten.peratur en (F) 125.79 119.60 Inlet Temperatures (F) '52.70 1 (Pool Temp) 85.00 Water Flow Rate =, (M1b/Hr) .225 .175 Det .: y Heat Loeu (NE.TU M ir > 6.055 Hest Exchanger Suriece (ir e- (Sqc t ) 500.O huat Eu t t mnger O ct ' c1 Conductance 37D.O Dtu/SqFt-Hr-F Thc- two p 32, cyr tic . 4ector, F .573
. . ~ . . - . . _ . . _
, .;. 7,__._.....,.
V Wri tount V; nhee EF F Eipesu on 11/i.%16/18 Spent Fuci Pool Cooling System Servico Water Systo.m e . . - - - _ _ . _ . . . . . - . . . _ . . . - _ _ _ - _ _ . _ . . _ - - - _ . .
k' 1
' Out l e ; ( e.mpc" atures (F) 125.42 119.29 I r. ) e t l eraperetut en (F) 102.04,(Pool Ternp ) 85.00 Watur Fl ow Fiates (M1b/Hr) .225 .175 DutN y Heat. Load' 01 BTU /Hr) 6.000.
He.at F.n c h en g e r : Sur i sce fir ea ( So,F t ) 500.O Hunt Tw c h alg a Ov H a) 1. Ccodu: i ence ~'70. O Bt u/54Ft-Hr-F The t wo pass corrcrcti on factor,: F= .873
' Uc+ aiunt Yant ec EFF Ex pensi on 11/12&36/18 Epsnt Fuel Fool Cocling S y s-, t un Ecrvice Water System Out l ei 'i m;per ilut w (Fe 125.OE .
110.99 Jolet Tempc ctr u /. F ) 101.52 (Pool T c rep ) Sie. Os Wei.et F1 ov. Eetwe ' (M1 b /l;r ) .225 .175 Deem % i i. (,00 (NET U/Hr > L.949 Heat En clsigor Sur ia e Arua ( E;qFL ) 500.O Heet Ot t h e: eq u nmrtil Contlu c t snce 375.O Btu /Sq't-He-F '
1he t wo p a s t, correction f ctor , F = .673 L Ver toont v .' n h < e EEF Enpansion 11/12516J1F 5 pent Fuel Pool Cooling Systein Service Water System Uutlet koptraturum (F) 12?.73 118.'72 Inlet Aaps atures T) 1SO.97 (Pool Tempi 85.00 i
1 . . . . -
Watee F) ou F:a le s ( Fi b /Hr > .225 .175 Decay He at Load '. MBT U/ Hr )- 5.900
. htm t Ed. c h en g o# - Evrf ea Area (SqFt) 500.0
~'
Heut Encher ger Oven al 1. Conduct ance 375. O bt u/SqFt-Hr -F The twc pau correc.tirn factor, F = .87 Ver tue n t Yankee SFP Expansion 11/12 Lie /16 Spent Fuel Pool Cooling System Service Water System
'Out bd ' Teenporatures- (F) 124.44 118.4D Inl et Temp eraturer., (F) 150.46 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Detry Hest Load (tibT U/ Hr ) 5.BS4 Heat Exchanger Surface Are6 (5qFt) 500.O l Heat E:<chenger Over ell Ccad.tcLence 375.0 blu/5qFt-Hr-F l; T he ' t.wo . pe s s corr ec t i on factor, Fa .870 l
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. . _ . . ~ . .
7,, , , ,
Vert-ic nt uni.ce SFF E u p an t>i or 11/12& 16 /1 F, Epeist. Fuel Fool Cuoling System Service Wat t er Sy' tem Ou i . r t. Temperatures, (F) 124.14 118.20 loj el Ternper at ur en (F) 149.V6 (F ool Temp) 85.00 Water Flow Ratew < !11 b /Hr ) .225 .175 Decoy Hect Load (NOTU/Hr) D.810 Heal F.x c t io n g er G.4 fecu Arca (5qFt) C00.0 Heat E ; c.h r ny tc, Deere11 Conductance 375.0 Et.u/bqFi-Hr-F The tn3 pasu cotrection factor, F= .87!
V ' m <. r i t i m o r. e e 9FP En p s.n t i on 11/12016/1E hpent Fuel Pool Couliiig Sy sstem ner ex cu Water System OutleL o op a -Lur es (F) 127.86 117.'?6 In)(1 Tt spe t a t u: ie (Fi 149.47 (Paul T et-ip ) E5.00 We.t er Fl uis Fei ci I Mi tt / Hr ) .225 .175 Di c e- Hi,i i im (rWTU'Hr) D.760 HuL E n c h s,- i i w - Euriacu Area (SqFt) 500.0 He et En d o nqp. c1] Cen W.tctarv e '7D.O Blu/SqFt-Hr-F liie two pen ccrrection factor, F u .873 er mci rt W il:e. c 5' P E;pantirn 11/1 T":16 /1 P Epent Fuel Poo; Locling Sy sst em Service Water System
.~----- --.------- - .- .- - --- -.-.-- -- ---- ..
Uv t. l e t ~i p o u t 4 sLure. (F) 127.59 117.73 Inlet Tt mpe. et ur c; ( r- ) 349,og (poo) 7 ega;, ) 65.OC W r.1 o r Flow rsa; e. 'hii,/Hr .225 .175 Decc) Heat oad (METU/Hr ) U.728 hwt E m t h -; n c, e r surfece fir e c (Sqft) 500.0 Heat Enc.hangu Over a) 1 Conductente 375.O Utu /SqF t--Hr -F ilie tw p e: tr c o~ r o . . on f ac tor , F = . C T' Vermont Yankee EF F' Expantion 11/126.16./16 Eper t Fuc1 Fool . Cooling Syrten Service Water System l
l Outlet Tenperotures (F) 123.33 117.51 In1et Temperatures (F) 148.62 (Pool Temp) 85.00 Weter Flow Rates (M1b/Hr> .225 .175 Decey Meet L ost! (MbTU/He) D. 6M Heat Exchanger Surface Area (SqFt) 500.0 Hest Em hnogti Over el] Conductance 370. O Ftu/EqF t-Hr -;
1hu twu p a s +.; c or r ec t i on factor, F = . 9 7 ~~
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.. . . ; , . ; , . . :=,; - ~~; .: h.-~m?m~:m- . . . . - >
l a
W'r mon t Vankee SFF Enpano2 uti 11/12"16/10 bpont Fuul Pool Coo;ing Systoai 5ervice Water Sys t em Ou t 1 o 1. . Temperatures (F)- 123.09 117,31
' I n L_.t Ternperatur eu (F) -148.22 (Pool T ernp . 8D.00 Water.F]ow Rates (M1b/Hr) .225 .175 Decey' Heat Load ( ME'T U / Hr ) ,
D.654
'Hea t E>t chang er Sur f ece Area (SqFL) 500.0 Heat Ein chhn ger Cvrtall Conductance 375. O Btu /S.;Ft -Hr -F
The two pass correction factor, Fe .873 Vermont Yankee SFP Expantion 11/12&16/18 Fool Co.21 i ng Syst ein Ser vi ce Water System Spent Fuol. _ _ _ _ _ _ _ _ _ _ _ ' _ _ _ _ _ _ _ _ . _ _ _ ____ __.. _ . _
Outles ~eniper r3turne (F). 122.85 117.11 In]Lt Ten.per a1 ur et. - ( I- ) 147. 53 '.(Pool Ternp ). 09.00 Uuter Flow F .e t n w -- ( t11 p / Mr ) .223 .175 Decay-Heet .uad (MO.U/Hr> D.619 Heal Exchangur Bur +aa Areci (SqFt) 500.O H#at ' E,x cheuiger Over a) 1 Cat aiuct once 370.O Ftu/SqFt-He-F
~~The Luo pe.a currection f a t t u r ., F.= .673 W r niont Venkee EPP Enpe rmion 11/12L 1.6 /10~~
~~5 pent Fuel Pool Cooling System Survice Water Systen:~~
Outlet i m ipe e tur e= (F- 122.63 116. ci l Inlet Temper atures (F) 147.15 (Pool Terop ) E *. 00 Weter Flow Netc, (N b/et' .225 .175 Decay Hed L oa t.' (t1E TU/Hr > 5.555 Heat Erthanym' Surface Area (Sqrt) .gog,o i Heat Encr angr.e- Ovei all Conductance 370,0 Btu /SqFt-Hr -F The two pm cor recti on f actor , F = .873 Ver man 'i an k ee SFP Expancion 11/12&16/16 Spent Fuel Fool Cooling System Service Water Byctem 9
Outlet T eraper a tur es (F) 122.40 116.73 Inlet lemperatureb (F) 147.08 (Pool Temp) 80.00 Water Flow Rates (M1b/Hr) .225 .175 Decey Heat Loed (MBTU/Hr ) 5.052 Hud Enchanger Surface Area (5qFt) 500.0 H.m t Euthanger Dverall Ccoductance 375.0 Btu /SqFt-Hr-F The two pass correction f actor , F - .873 l I
I
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W t trent Yankee SUP Enper. bion 11/12 Lie /18 Spent Fuul Pool Cool i'n g S y st e.n oer vice Water Eyt.tu.n D.utl e t. Temperature '(F) 122.19 116.55
?nlet Temper atur es W) 146.73 (Poul Ter p) 8D.00
'LLs Flow Ratus U1);/Hr .225 .175 lhtey Heat Lead (MF:7 L1/Hr ) D . S. 1 1 He t t 'Es ch eng;er Sur i aco W ee-(EqFt) GOO. O Rn L t.schunger . U 'e a) 1 CcoductAnce 3 70. O Btu /Sq:t-Hr -F
-E e tw.c, pa h :or r ec.t i on factor, Fu .873 Uet t..un t Y ard.ee EFF E v.p en ::,i un 11/12.H L/18 Spent Fuel Foci Cauling Eystem Eurvice uter Syst or.
Ou tl el Teo.per4turti M~) 121.99 116.38 Inl et 'i ving e 6 L u n b (F )' 146.40 (Pool Temp) 8S.OD Water F 2 06v t. oles 01] /He) . 223 .175 Decey ek,. c t L o .: tl G1ET!!/Hr ) 5.491 Huet En:hengtt- 5.a f acc Area (SqFt) 500.0 Heat & chanut.r Over all Cofiductance 370. 0 Btu /St:F L-Hr,-F The two p r.t w i w rection fM.tur, F a. .U73 Ver mont Yankee EFP Enpo.nti on 11/12L16/18 S p t.:n t Fuel Fool Coo 11r g Systeu Service Wter System
_.. .______ _-___---_---_ _ -_ _ ~.-__.. -__ --_
Latlet To,op er #. . m .Ff 121.80 116.11 Inict ienperaturis (F; 146.07 (Pool Temp) EL.09 Wbter Flow O l t, , (MiU/H .223 . 175
.Decc a y Heal Lokd Ulbl U/ -Ir l ) U.462 '
Hea t Enc ha m .<r Eurfece Ar es ( S q r l. ) 500.0
. Heat E xc.h m .g t r Dverall ~unduttance 370. 0 Blu /SqFt-Hr -F The tuo pers :u re ction f cc t.::c , F= . .873 Mctmoni Yani.ee SFP Exponi.eion 11/1 L L '. 6 /1 E I I
S,.> e n t. Fuul Pool Cooling System Scrvice Water Systen /
1 Outlet Temper atures (Fi 121.51 115.97 Inlet Temper atures (F) 14D.60 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .223 . 175 1)ectiy Heat Load (MBTU/Hc) 5.420 Heat Exchanger Curface Area (SqFti 500.O Hest Exthenger Over c.11 Conductance 375. O Fit u/ Sq:t-Hr-F ThC two pass correction factor, F = .873
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Ver snen t Yamkee SrP Eapension 11/12&l6/10 Spent Fuel Fool Cooling System Service Water Syttum Out l ot Temperatures (F) 120.27 114.91 Inlei lemper atur et (F) 143.53 (Pool Temp) EU.OO ,
I
' bate- Flow Rates' (M1b/Hr ) .225 .175 Decey Heat Load (MBTU/He> 5.233 Hect Exchanger Sif ec e Ar ec . (SqFL) i 500.0 Heat Enchanger Dve* v11. Conduct,.<nce 375. O Blu/SqFt-Hr -F The two pass -correction f actor , F= .873 W rniont Yanh + SFP Expanrnion 11/12016/18 Spent Fuci Paul Cooling System Service Water System Outlet Temperature, (F) 1 19. 2"~ 114.03 Irel et T e oper v u., m (F , 141.81 (Pool Temp) 83.00 Watei Flow Ralet i. H1 ;.3 / Hr ) .225 .173 Dec. a y Heut Lued (MF.TU/Hr ) 5.091 L H, rat Ekhengier Sutface Area (SqFt) G00.0 Heat Exchanger Ovtr 4.] 1 Ccoductante 375. O Bt u/SqFt- Hr -F lhe~two pass cor rection factor, F = .373 Verment Yankee SFP ExpenLion 11/12!16/18 Spent F uci Pool Cooling System Service Water S yste:m Outlet Tungeratur es (F) 118.34 113.26 Inlet Temper 4.stut et (F) 140.34 (Pool Temp) 80.00 Wat er Flow Rutee 'Mih/Hr) .225 .175 Decay Heat Load (MBTU/Hr) 4.949 Hoet Enchanger Eurfaco Area (SqFt) 500.0 Heat Enchanger Overal1 Conductance 375. O Btu /SqFt-Hr -+
The too pens cor t ectior- factor,'F = 873 Ver mont Yankee SFP Expansion 11/12&l6'la Spent Fuel Pool Cooling System Service Water System
.Outlut Temperatures (F) 117.55 112.61 Inlet Temperatures (F) 139.03 (Pool Tempi 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (MBT U/Hr ) 4.832
'Huat Exchanger Surface Area (SqFt) 500.0 Hest Exchanger CNet e] 1 Conductance 370.0 litu/SqFt-Hr-F The tuo pas- ct. **ecc t i on f actor , F = .873 I
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Spent Fun] Pool Cooling Systern Service Water S /w lui.
Ou t l e+: T ernp ur a t ur e s, (F) 117.36 112.45 Inlci Temper a t t tr n, IF) 135.70 ( F'oc.1 Teng ) 85.00 Water Flow Rat eu (M1 b /lir > .225 .175 Det c ,- HL et L asti (t9TLF/ fir > 4.GOT 12 u i. E,che<nger Sur4 ace (,r es (5qFt) 500.O Lie. c; E; - tisi Jt v l' ,~.'r 4 1 1 Co' .tlu t.t #:n c e 37D. O B t u /Dar t -Hr -F Tire unu p .wa cuirection f ac tur , F 2 .573 i
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Ver mont Yankee SFP - Par t II'- 18 Month
. 8,e t c h No Eun Decay Heat Cumm Heat Cumm 4t Bun Ts l
1 132 6.57369 6.57369 132 .02 2 128 .45787 7.03156 260 1.52 3 132 .27582 7.30738 392 3.02 4 128 .23804 7.52542 520 4.52 5 132 .20677 7.73219 652 6.02 6 128 .19101 7.92320 780 7.52 7 1!2 .10942 8.11263 912 9.02 8 128 .17706 8.28969 1040 10.52 1 9 132 .17612 8.46581 1172 12.02
.1 0 128 .16476 8.63056 1300 13.52 11 132 .16391 8.79448 1432 15.02 12 13e .16293 8.95740 1568 16.52
'13 120 .13869 9.09610 1688 18.02 As s u top t i on n .
Power per Dundle s. 15.44197 MB1U/Hr 1rradiatio. Tinm a 39420. Hours Reactor Fower = 1665. MWt h lime (4fter Shutdown m 144 Hours or 6.00 Days Reieronte: NUREG-CEOO, St ande d Re vi cw Plan (1) Section 9.1.3 Spent Fuel Peol Cooling and Clean-up Systremc, Rev. 2 (2)- Fratich Technical Position ASB 9-2
' Resi du al Detsy Ener gy for Light-Water React ors for Leng Term Cooling, Rev. 2 i
Spent Fuel Pool Cooling System Service Water System Outlet Temper atur ec (F) 146.33 136.98 Inlet Temperature (F) 186.76 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Heat Exchanger Surface Area (SqFt) 500.O Heat Exchanger Overall Conductance 375.0 BTU /SqFt-Hr-F The two pass correction factor, F= .873 l
wa-__ -- a. --
c +. . :. ..s .. :; .m m :~: n :~. V ::~:: T ~ ~~ ~ ~v Ver n.ont Yankee SFP - Part 11 - 18 Month De r T 2 tre Decay Heet Pool Temperature F factor (Hours) (Days) (MBTU/Hr) (degrees F) 144. 6.00 9.09610 186.757iO .8734 156. 6.50 8.87982 184.28820 .8731 160. 7.00 8.68899 182.17070 .8732
'160. 7.50. 8.51934 180.26820 .8732
1 192. B.00 8.36741 178.57100 .8732 204. 8.50 8.23035 177.03760 .8732 ;
216. 9.00 8.10581 175.64510 .8732 228. 9.50 7.99187 174.37100 .8732 24O. 1O.00 7.88696 173.19770 .8732 252. 10.50 7.78975 172,11070 .8732 264. 11.00 7.69916 171.09760 .8732 276. 11.DO 7.61428 170.14850 .8732 288. 12.00 7.53438 169.25490 .8732 300. I2.50 7.45881 168.40990 .8732 312. 13.00 7.38706 167.60750 .8732 324. 13.50 7.31860 166.84290 .8732 3 ~.,6 . 14.00- 7.2D331 166.11180 .8732 34b. 14.50 7.19062 165.41080 .8732
-360. 15. % 7.13035 164.73680 .8732 372. 15.50 7.07227 164.08730 .8732 384. 16.00 7.01619 163.46020 .8732 396. 16,50 6.96195 162.85360 .8732 4OC. 17.00 6.90939 162.26590 .D732 420. 17.50 6.85340 161.69570 .8732 432. 18.00 6.80886 161.14170 .8732 444. 18.50 6.76069 160.60300 .8732 456. 19.00 6.713'/9 160.07860 .8732 468. 19.50 6.66811 159.56770 .8732 480. 20.00 6.62356 159.06950 .8732 4 9:.: . 20.50 6.58009 158.58340 .8732 504. 21.00 6.53766 158.10890 .8732 516. 2i.50 6.49621 157.64540 .8732 528. 22.00 6.45571 157.19250 .8732 540. L. 50 6.41611 156.74960 .8732 552. 23.00 6.37738 156.31650 .8732 564. 23.50 6.33940 155.89280 .8732 57e. '24.00 6.30240 155.47800 .8732 588. 24.50 6.26609 155.07210 .8732 600. 25.00 6.23054 154.67450 .8732 612. 25.50 6.19572 154.28510 .8732 624. 26.00 6.16161 153.90360 .8732 636, 26.50 6.12818 153.52990 .8732 648. 27.00 6.09643 153.16350 .8732 660. 27.50 6.06332 152.80450 .8732 672. 28.00 6.03184 152.45240 .8732 684. 28.50 6.00097 152.10730 .8732 696. 29.00 5.97070 151.76870 .8732 I 708. 29.50 5.94101 151.43670 .8732 720. 30.00 5.91188 151.11100 .8732 732. 30.50 5.88331 150.79150 .8732 744. 31.00 5.85527 150.47790 .8732 756. 31.50 5.82776 150.17030 .8732 768. 32.00 5.80076 149.86830 .8732 <
780. 32.50 5.77425 149.57190 .8732 792. 33.00 5.74823 149.28100 .8732 804. 33.50 5.72269 148.99530 .8732 816. 34.00 5.69761 148.71480 .8732 -
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.D.' . ow. :.:. 4 '. av - * ' J .' b ? N/" ~ 1 *+ b . '*t D Y W ~ '~ * " ' . ' c / M 840.' 35.00' 5.64878 148.16890 - .8732 852. 35.50' 5.62503 147.90320- .8732 l .- 864. 36.00 5.60169_ 147.64220 .8732 876. 36.50 5.57876 147.38580 .8732 8ea. 37.00- 5.55623 147.13390' .8732
/<. ' . 37.50 5.53409 146.88630 .8732 912. 38.00 5.51234 146.64300 .8732-924. 38.50- 5.49096- 146.40390 .8732 936. 39.00 5.46994 146.16890 .E732 948. 39.50- 5.44928' 145.93790 .8732
.960, 40.00 5.42897 145.710eO .8732 972. 40.50 5.40900 145,48740 .8732 964. 41.00 5.38936 145.26780 .8732 996. 41.50 5.37005 145.05180 .8732 1008. 42:00' '5.'35105 144.83940 .D732 1020, 42.50. 5.33237 144.63040 .8732 1032. 43.00 5.31390 .144.42490 .8732 1044. 43.50 5.29590 144.22260 .8732 10D6. 44.00 5.27810 144.02360 .8732 1063. '44.50 5.26059 143.82780 .8732 1080. 45.00- 5.24335 143.63500 .8732 1092. 4S.50 5.22639 143. 4 fl530 .8732 1104. 46.Ou 5.20969 143.25850 .8732 143.07460 .8732
~
1116. 46.50 5.19524' 1128. 47.00 6.1770D 142.89360 .8732
'1140. -47.50 5.16111 142.71530 .8732 1152. 48.60- 5.14541 142.53970 .8732 1164. 48.SO .5.12995 142.36680 .8732 1176. 49.00 5.11471- 142.19650- .8732
'1180. 49.50 5.09970 142.02860 .8732 l
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Ver rnont Yankee SFP - Part I - 12 Month
'Dctch- No Bun Decay Heat Cumm Heat Cumm 46 Bun Ts I 104 .11113 .11113 104 .19.00 1
2 109 .11267 .22380 212 20.00 3 120 .12223 .34604 332 21.00 4 80 .07956 .42560 412 22.00 5' 96 .09322 .51882 508 23.00 6 96 .09102 .60984 604 ,24.00 7 84' .07776 .68759 688 25.00
'8 136 .12292 .81051 824 26.00 9 328 .28944 1.09995 1152 27.00 10 30 .02585 1.12580 1182 28.00 Assumptions:
Power- per Dundle ' 15.44197' MBTU/Hr Irradiation T i tr c = 35040. Hours Fw c.c t or Power c 1665. MWth Time After Shutdown =166440. Hours or 6935.00 Days
~~Reference:~~
NUREG-0800, Standard Review Plan (1) Section 9.1.' 3 Spent Fuel Pao) Coo)i ng e nd C1ean-up Bystems, Rev. 2 (2) Branch Technical Position ASB 9-2 Residual Decay Energy for L i ght-Wat er Reactor s for.Long Term Cooling, Rev. 2 Epetit fuel Pool Cooling System Servi ce Water System Out I rt Temper at ur es (F) 97.59 91.43 j inlet lemperaturce (F) 97.59 (Pool Temp) 65.00 l Water Flow Rates (M1b/Hr) .225 .175 Heat E>: changer Surface Area (SqFt) 500.0 Heat Enchanger Over all Conductance 375.0 BTU /SqFt Hr-F The two pesd correction factor, F'= .873
____m_. _ . _ _ _ _ _ . _ _ _ . _ _ _ - _ _ _ _ - _ -
~~a. . . , . . . . .. ..: .~:~~
Wr mont Yankee SFP - Part 1 - 12 tionth
(
DF s y Time Decay Hcnt Pool Temp er atur e F factor (Ham +) (Dayw) (t1 BTU /Hr ) (riegr ees F)
+++*++ *+++++ 1.12560 97.59418 .8734
~~+++w ****** 1.12576 97.58749 .8701~~
~~,,+ *+++++ 1.12572 97.58920 .8732~~
+*,+** *****+ 1.12569 97.58807 .8732
++,+++ ++++++ 1.12565 97.58791 .8732
+++n* * * + + + 1.12561 97.58740 .8732
++**** **n+* 1.12558 97.58702 .8732
+*.** ***** 1.12554 97.58660 .8732 oo+* n+*** 1.12550 97.58619 .E732
~~+**+*+ 1.12547 97.585'78 .8732~~
++++++ ++++++ 1.12543 97.58537 .8732 n++n ***++* 1.12539 97.58495 .8732 na + + ***+++ 1.12535 97.58454 .8732
+4 + + o +.+++e 1.12532 97.58412 .8732
++u++ ++++u 1.12528 97.58372 ,8732
+.+ok +*+n- 1.12524 97.53330 .E732 i
+++++.* +oa+, 1.12521 97.50289 .E732
.++*** *,*u
~~1.12517 97.58248 .8732 5~~
~~+n+*+ &,+++* 1.12513 97.58206 .8732~~
n+ + * * " + 1.12510 97.58165 .8732 p+n++ **+ w+ 1.12506 97.D8125 .8732 o o ** *n+** 1.12502 97.58003 .8732
++++" #- + + n + 1.12499 97.58042 .8722
+**n* +++*+* 1.12495 97.58000 .8732
~~n++ ++++++ 1.12491 97.57059 .8732~~
~~+*n++ ++++++ 1.12408 97.57918 .8732~~
~~n+n +*++++ 1.12484 97.57877 .8732~~
~~+ + * * * ****++ 1.12489 97.57835 .8732~~
~~>++ ++++v+- 1.12476 97.57794 .8732~~
+++ * + * *
~~1.12473 97.57753 .8732~~
~~+++ n.*+, 1.12469 97.57712 .8732~~
+++++* n o .* + 1.124e5 97.57670 .8732
++++++ ++**u 1.12462 97.57629 .8732
+**++v ++++-a* 1.12458 97.57580 .8732
+***++ ++++++ 1.12454 97.57547 .8732
+++++* +++++6 1.12451 97.57506 .8732
+++++* ++++++ 1.12447 97.57465 .8732 m **+ n**++ 1.12443 97.57424 .8732
++++*+ ++++n 1.12440 97.57382 .6732
~~+n** ****** 1.12436 97.57340 .E732~~
~~"" *++**+ 1.12432 97.57301 .8732~~
~~+**6 **+++* 1.12429 9'7.57259 .8732~~
~~++*+++ ++++++ 1.12425 97.57218 .8732~~
~~+** ****** 1.12421 97.57176 .8732~~
~~++++u ***n+- 1.12418 97.57135 .8732~~
*"*** 1.12414 97.57095 .8732 n+n+ *****+- 1.12410 97.57053 .8732
****** 1.12406 97.57012 .8732
~~+-**+++ *+++++ 1.12403 97.56970 .8732~~
~~+=*** ****++ 1.12399 97.56930 .8732~~
++++*+ *++u- 1.12395 97.56888 .8732 !
+++++r *+++++ 1.12392 97.56847 .8732 j
+**+** ++++++ 1.12388 97.568C6 .8732 j
+*-*** ***+++ 1.12384 97.56763 .8732 i i
~~+n+n + n o +- 1.12381 97.56723 .8732~~
~~+,* *****+ 1.12377 97.56682 .8732~~
+*++++ +n+++ 1.12373 97.56641 .E732
.m
....:.o/ - r. -m m o ~; .. . .
r > r , >,, ,,,-ae.
s . 1 c .i .s. r i . acaw . . 1
+++++* .8732 I
+ ,
~~s *+ 1.12366 97.56558~~
+n + wr +++-+** 1.12362 97.56518 .8732
+**+,* .+++++* 1.12359 97.56476 .8732 i
wn ++n ** 1.12355 97.56435 .8732 j
~~++++++ **++n 1.12351 97.56394 .8732~~
~~, + ~ , "*"* 1.12340 97.56353 .8732~~
++nn *- , o *
~~1.12344 97.56312 .8732~~
~~+++ o* ****** 1.12340 97.56270 .8732~~
~~++4** +++*** 1.12336 97.56229 .8732~~
~~+ 1.12333 97.56158 .8732 5 I~~
+>*
u+*** n**** 1.12329 97.56147. .8732
.*****+ ****** 1.12325 97.56106 .8732
~~++++ ++*+++ 1.12322 97.56064 .8732~~
~~+o* ***-***- 1.12318 97.56023 .8732~~
++++o + + + " + 1.12314 97.55981 .8732 ;
i
+++*o **+*** 1.12311 97.55941 .8732 i
+,n* nuu 1.12307 97,55900 .8732
+u*w * * * "
1.12303 97.55859 .8732 u+4u o + *u 1.12300 97.55:17 J .8732
+++++ +**+++ 1.12296 97.55777 .8732 ,
i
~
,* n u o*+++. 1.12292 97.55736 .8732
* * * " *to++ 1.12289 97.55693 .8732
+**o+ +++*o 1.12265 97.Sb654 .8732
++*4*+- *u*n 1.12201 97.55611 .8732
+ * + + " ++**+" 1.12278 97.55572 .8732
~~++** +++*** 1.12274 97.55330 .8732~~
~~++w** ++++++ 1.12270 97.55489 .0732~~
~~H*** ****** 1.12267 97.55447 .8'732~~
~~++++* ***-+" 1.12263 97.55406 .8732~~
****** 1.12259 97.55365 .8732 9
m.,
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l Wr rnont Yank.ee Per t II - Second Batch Batch ' No' . Bun > Decay Heat Cumm Heat Cumm 44 Bun Ts l
l l 1 132 6.57369 6.57369 132 .02
. 128 .45787 7.03156 260 -1.52 132 .27582 7.30738 392 3.02 4 126 .21804 7.52542 520 4.52 0 132 .20677 7.73219 652 6.02 6' 128 .19101 7.92320 780 7.52 7- 132 .16c742 8.11263 912 9.02 8 12e .17706 8.28969 1040 10.52 9 .132 .17612 8.46581 1172 12.02 10 12C .16476 8.67056 1300 13.52
11. 132 .16391 8.79448 1432 15.02 12 136 .16293 8.95740 1568 16.52 13 120 .13869 9.09610 1688 18.02
'(1stunip t i ons :
Power per bundle = 15.44197 MBTU/Hr
~ I r r e.d i at 2 on . 7 i era u . 39420. Houre a Reactor Power ' 1660. MWth
'Ti me M ter Ehutdown = 144. Hours or 6.00 Days ,
Rc4 orent c :. N! M C4-OE 00, St ander d Pcvit>w Flan (1) Section 9.1.3 Spent Fuel Pool Cooling and Clean-up Eyctems, Rev. 2 (2) Br anch Technicel Position ASB 9-2 Recidual DecEy E.ner gy f or Light-Water Reactors for Long Ter m Cool ing, Rev. 2 Spent Fuel Pool Cooling Syst eni Service Water Syst ern i
Outlet Ternper st ur es (F) 146,33 136.98 Inlet Temperatures (F) 186.76 (Pool Temp) 85.00 Water Flow Retes (M1b/Hr) .225 .175
. Heat Exchanger Surface Area (SqFt) 500.0 Heat Exchanger Overall Conductance 375.0 BTU /SqFt-Hr-F The two pass correction factor, F= .873 l l
l I
l.
L-
.,m . . . . . . . _ . :. .
i o. , m._, . ._ . . .. . . - . . . ....
Verniont Yankee Far t II - Second Batc.h 1
l-imcw Tina Decay . Heat Pool Temperature F factor
( Hou r .-a s - (h ys) (t1 BTU /Hr ) (degrees F)
'1441 6.00 9.09610 '186.75710 .8734 l' 168 7.00 8.68899 182.15440 .8731 l 192. B.00 8.36741 178.57450 .8732 216. 9.00 8.10581 175.64380 .8732 l 24O. 1O.00 7.88696 173.19820 .8732 264. I1.00 7.69916 171.09750 .8732 280. 12.00 7.53430 169.25500 .8732 312, 13.00. 7.38706 167.60750 .8732 136. 14.60 7.2D531 166.11180 .8737 l- ,360. 15.00 7.13035 164.73680 .8732
-3E4. 16.UO 7.01619 163.46020 .8732 400. 17.00 6.90939 162.265(70 .8732 432. 18.00 6.80886 161.14170 .6732 456 19.00 6.71379 160.07860 .8732 4D0. 20.00 6.62356 159.06950 .8732 504. 21.00 6.53766 158.10890 .8732 528. 2;.v0 6.45571 157.19250 .8732
'55?. 23.00 6.37738 156.31650 .8732 576. 24.00 6.30240 155.47800 .8732 600. 25.00 6.23054 154.67450 .8732 624, 2 6. nO 6.16161 153.90360 . 873:/
64a. 27.00 6.09543 153.16350 .8732 c;72 . 2b.C0 6.03184 152.4b240 .873.4 696. 29.00 5.V7070 151.76970 .8732 720. 30.00 5.91184 151.11100 .8732 744. 31.OO ".85527
. 150.47790 .8732 768. ~J 2. 00 5.80076 144.86830 .8732 792. 33.00 5,74823 149.28100 .8732 016. 0.a . 00 5.69761 148.71480 .8732 G40. !5.00 5.64878 148.16890 .8732 .
864. 36.00 5.6016t/ 147.64220 .0732 888. 37.00 5.55623 147.13390 .8732 912. 38.00 5.51234 146.64300 .8732 936. 39.0C 5.46994 146.16890 .3732 960. 40.00 5.42897 145.71080 .8732 984. 41.00 5.38936 145.26780 .S732 1008. 42.00' L.35105 144.83940 .8732 1032. 40.00 5.31398 144.42490 .8732 1056. 44.00 5.27810 144.02360 .8732 1080. 45.00 5.24335 143.63500 .8732 1104. 46. OC 5.20969 143.25950 .8732 1128. 47.00 5.17705 142.89360 .8732 1152. 48.00 5.14541 142.53970. .8732 1176. 49.00 5.11471 142.19650 .8732 1200. 50.00 5.08492 141.86330 .8732 1224. 51.00 5.05599 141.53980 .8732 1248. 52.00 5.0278(? 141.22560 .873?
1272. 53.00 5.00058 140.92020 .8732 1296. 54.00 4.97403 14O.62320 .8732 l 1320. 55.00 4.94S20 14O.33440 .8732 1344. 56.00 4.92307 14O.05340 .8732 j 1368. 57.00 4.89861 .139.77980 .8732 '
1392. 58.00 4.87478 139.51330 .8732 1416. 59.00 4.85156 139.25370 .8732 144O. 60.00 4.82892 139.00060 .8732
.1464. e1. 00 4.80685 138.75370 .8732 14 % A7.00 4.78531 138.E1280 , ,, 8'f 3 2 ,
. . . . . . . . . . - . ~ . . _ . . , . . . ..._. .._ _ , ,.
3# 7; n g. e,3 . v u -- - ' C 7 6 4W ' ' " 1 Sd . 2 N /O , ~ ~"*%/32 .
' ~ ~
1036. 64.00 4.74376 138.04820 .8732 l"/6 0 . 65.00 4.72370 137.82390 .8732 1584. 66.00- 4.70411 137.60480 .8732 1608. 67.00 . 4.68494 137.39050 .G732 1632. 6E.OO~ 4.66620 137.18090 .8732 1600. 69.00 4.64786 136.97380 .5702 1680. 70.00 4.62991 136.77500 .8732
- 17v4. 71.00 4.61232 136.57840 .8732 q 1728. 72.00 4.5952O' 136.38580 .8732 i 1752. '73.00 4.57822 136.19700 .8732
.1 '/ 7 6. .74.00 . 4.56167 136.01190 .8732-
'1600. 75.00 4.54544~ 135.83040 .8732
'1824. 76.00 4.52951 135.65230 .8732:
1848. 77.00 4.31388 135.47760 .8732
~1872. 79.00 4.49854 135.30590 .8732 .
1896, 79.00 4.48347- 135.13740 .8732
.1920. 90.00 4.46866 134.97180 .8732 1944 61.00 4.45410 134,80910 .8732 1969. 82.00 4.43980 134.64900' .8732 1992. 83.00 4.42572 134.49170- .8732 2016, 84.00 . 4.41188 134.33690 .8732 2040. E3.OU 4.39826 134.A8460 .8732 i
se . *
- . . _ _ . - . - -. ~ . . - . _ --..--.-_.___.____.___.____.__-__._-__a --
g c g . . . ; -j ~- .-~ ~;v.~: :w,= . : ~-~ ~ :: - ~ ~ .:- ~-.::n ::
~
Verrnant Yankee Part I - Second Batch Petch. No Bun- Decay Heat Cumm Heat Cumm M Bun Ts
-1 104 .11113 .11113 104 19.00'
'2' 108 .11267 .22380 212 20.00 3 120 .12223 .34604 332 21.00
!4- 80 .07956 42560 412 22.00 5 96 .09322 .51882 508 23.00
'6 96 .09102 .60984 604 24.00 7- 84 .07776 .68759- 688 25.00 8 136 .12292 .61051 824 26.00 9 328 .28944 1.09995 1152 27.00 ,
10 30 .02DED 1.12580 1182 28.00 f 1
Assumptions:
. ~ _ _ _ _ -
' Power per. bun 41e u 1G.44197 MBTU/Hr 1 r redi'eti on Time u 35040. Hours React or Fower e 1665. MWth Time After Shutdown =166440. Hour s or 6935.00 Days Reforence NUREG-0000, Etandard Review Plan .
e (1) Section 9.1.3 Spent. FuH . Pcol Cooling and Clean-up Eyctemk, Rev. 2 (2) Eiranch Technical Position ASB 9-2 Retidual Decay. E.ner gy f or Light-Water Reactors for Long Term Cooling, Pev. 2 i
Spent Fuel Pool Cooling System 5er vice Water System
.t Dut i et : Ternper atur er iF) 92.59 91.43 !
Inlet Temperaturrm (F) 97.59 (Pool Temp) 85.00' Wat er Flow Rates (M1b/Hr) .225 .175 Heat Exchanger Surface Area (SqFt) 500.0 Heat Ex c hanger - Over el l Conductance 375. O BTU /SqFt-Hr -F The two pass correction factor, F= .873 1.
l l
, we *
.;mc. .c r --- - .
~~a. . . . - . . , . . .~~
Ver mont Yeokee Part I - Second Betch huf i;no Dece heat Pool Temperature F fcctor (Hour r (Da p) (METU/Hr) (degr een F )
+++n+ m ++* 1.12580 97.59418 .8734 o*++* *+++++ 1.12572 97.58707 .8731
+n++# *n+n 1.12560 97.58837 .8732 oo+a ****** 1.12558 97.58683 .8732
++*++4 *+n++ 1.12550 C7.58625 .8732 I
n+n+ + + o +4 1.12543 97.58534 .8732
+ u +++- o+n* 1.12535 97.58454 .E732
~~w++* * * " + + 1.12520 97.58371 .8732~~
~~+++++* +-+>+*+- 1.12521 97.38789 .8732~~
*o++ j.12513 97.58206 .8732 u++++ +*+W++ 1.12506 97.58124 .6732
* * * * *+**++ 1.12499 97.58041 .8732
~~+++n+ + n*++ 1.12491 97.57959 .8732~~
~~+++++ *+*-++ 1.12484 97.57877 .E732~~
%++++ -
+***+. 1.12476 97.57794 .8722 u a * *+ w+o++ 1.12469 97.57712 .8732 o+n+ + wu u 1.1:.462 97.57629 , 873'?
~~++ m * * , + . 1.12454 97.57548 .8732~~
~~++++* +*+e 1.12447 97.57t65 .8732~~
~~++* *+++++ 1.12440 97.57583 .8732~~
++++++ n**++~ 1.12432 97.57300 .877?
+++++* ++**+* 1.12425 97.57217 .8732
., + + + o ++-+w 1.12418 97.57136 . 87 3:.
~~+++** +++++- 1.12410 97.57053 .8732~~
+nm **HH 1.12403 97.56971 .8732 m * , ,- ++***+ 1.12395 97.56838 . C 7.32
++++n 4- u + + + 1.1236S 97.56005 .8732
"++** *om 1.12381 97.56723 .8732
++++++ "+a+ 1.12373 97.56641 .8732 n++++
~~o .+4 1.12366 97.56558 .8732~~
~~++*H + +++*#+ 1.12359 97.56477 .8732~~
~~+++*+ ++,.-*++ 1,12 51 97.56394 .8732~~
~~+o *o+*& 1.12344 97.56312 .8732~~
~~++ * +o++* 1.12336 97.56229 .3732~~
+ o+** H++*+ 1.12329 97.56147 .8732
+w.*+ ++++++ 1.12322 97.'56065 .87~7
++nn a++*+ 1.12314 97.55982 .8732 n +*** +++u- 1.12307 97.55900 .8732
+***** ++++++- 1.12300 97.55817 .8732
~~+- *++u+ 1.12292 97.55735 .8732~~
+o*o +.++H 1.1228". 97.55653 .8732
+++**+ ****** 1.12278 97.55571 .8732 oen+ -++++++ 1.12270 97.55489 .8732
****++ 1.12263 97.55405 .8732
~~+***++ ***u* 1.12256 97.55325 .8732~~
*****+ 1.12248 97.55242 .8732
~~+++* **+n+ 1.12241 97.55160 .8732~~
,***** ****++ 1.12233 97.55077 .8732 u ++* +++.++ 1.12226 97.54995 .8732
++++*+ n*+++ 1.12219 97.54913 .8732 H u ++ ***+++ 1.12211 97.54830 . 832
****++ 1.12204 97.54749 .8732
~~++++n *no* 1.12197 97.54666 .8732~~
~~+++*+ +++++* 1.12189 97.54584 .5732~~
u*+ nuo 1.12182 97.54501 .8732 n*** ****++ 1.12175 97.54420 .8732
+u*w m++- 1.12167 97.54337 .8732
... ._........_ _ _..-. , ._..-.m._._..
. D ' *, + a . .* .*u*** 1. 121eiv v / . b4.RZi
'. 0 / 52 I
,,, + ny. 4+ *n*** 1.12153 97.54173 .8732
~~++++,* 1.12145- 97.54091 .8732 1~~
~~,nsa* **+++* 1.12138 97.54008 .873'. f '~~
~~o4* *+**** 1.12131 97.53926 .8732~~
.*u++ ++**** 1.12123 97.53844 .0732
+-+.+*u. ****+* l'.12116 97 53762 .8732
++n*+ +++++* 1.12109 97.536E0 .8732 .
.n+<,*+ u * ,- +
~~1.12101 97.55598 .8732 l~~
~~++ *r n+** 1.12094 97.53516 .8732 l~~
=+,*W **n** 1.12006 97.53433 .8732
++++*+ *n ++:* '1.12079 97.53351 .8732
+****. ****** 1.12072 97.53269 .8732
+*nu *u*** 1.12064 97.53187 .6732
. o . *+ * *s(, 1.12057 97.53104 .8732
+++++* *+*n+ 1.120D0 97.53023 .8732 i
~~+-** u**** 1.12042 97.52940 .8732~~
~~+**u +++++v 1.12035 97.52858 .5732~~
~~.***- *n*u 1.12028 97.52776 .8732~~
~~++x+++- o++++ 1.12020 97.52694 .0732~~
~~+,** *ne 1.12013 97.52612 .8732~~
, + * , - * +++.-*4 1.12006 97.52530 .8/32 o*n- H+*++ 1.11998 97.52448 .0732 f
I m______._________ _ _ _ _ _ _
~~.-: n : : . .. -- : -~~
n .. . ..
. Vermont Yankee End Run !! 18 Month Betch No Bun . Decay Heat Cumm Heat Cumm # Bun .Te 1
i cl. 132 1.77895 1.77895 132 .27 2' 128 .40097 2.17993 260 1.77
, ~5 132 .26218 2.44211 392 3.27
/I ' -128 .21385 2.65596 520 4.77 S- 132 .20478 2.86074 652 6.27 6; ,128 .18967 3.05041- 780 7.77
'/ 132 .10S21 3.23862- 912 9.27.
C. 129 .17D96 3.41459 1040 10.77 9 132 .17504 3.58962 1172 12.27 10 128 .16374 3.75336 1300 13.77 11 132 .16291 3.91627 1432 1D.27 iT 136 ,16193 4 . O'7 E 2 0 - 1G69 26.77 13 120 .13784 4.21604 1688 10.27
. Astepti vou:
Power per Bun d l 6. -
- 15.44197 MBTU/Hr 1rradiction.Timo a 39420 Hours Reactor Power " 1660. MWLh l Ti me Af ter Shuttiawn u 2400. . Hours or '100.00 Deys .
Ficf er once t 'trIREF -oil 00. 5 tar.dard Pevies P1an
-(1) Section 9.1.3 Spunt Fuel f- col Cool i r.g and Cl ean-up E) si ems , Rev. 2 (2) Branch Techni cal Posi ti on ASB 9-2 F esi dual Decoy _Enorgy 4 or i_i g b t-W e t er Reett t ur e for ot ig .l er m Cooli r,g , Rev. 2 Spent Fuel Fool Cooling Syst em Eer v2 ce Water Syst en Outlet 1 emptr et ur ets (F> 113,.43 109.09 Ird et Te mp eat uren F) 132.16 (Pool Temp) e5.00 Water Flow h i.es (N1 ts /Hr ) .225 .175 Heat Ehchanger Surfece Area (SqFt) 500.0 1 Heat Exchanger Over al l Conductance 375.0 BTU /SqFt-Hr-F 1he two pass correction factor, F= .873 3 1
l 9
' .. + .
-e +.**..u.W ate. .a 5 . , .e-+-..
[, +
%.r munt Yuikce Ehd Run ! !: 1C Month L'
Dcu y Tido _Decey Heat -
Pool Temp er dur o F' factor (Hour t.) - (Layt) (MBTU/Hr ) (degrees F)
__a_....._-______.. ________- _ _ _ _ _ _ _ _ _ . _ _ _ __ _ __
2400. 100.00 4.21604 132.16440 . 8734 2424.- 101. 00 4.20513 132.01890 . 8~731
~2448. '102.00 4.19434 131.90620 . 8732 2472. -103.00- 4.18369 131.78440 . S732 h
'2496. 104.00- 4'.17315 131.66750 . E'732 2320. '105.00 4.16274 131.55070 . 6732 2544, 106.00 -4.15244 131.43570 . F732 2563. 107.00 '4.14226 131.32170 . 8732 2092. 10E.00 4.1321P 131.20910 . 8732 l 2616. 169.00 4.12222
.131.09770 . 8'732, 2640. '10.00
) 4.I1236 130.987t.O , 8732-2664. 111.00 4.10260 130.87830 . 8732 26 b(J. 112.00 4.09294 130,,77030 . 8732 h 2712. 113.00 4.08338 130,66340- . 8732 2736. 1.14.00 4.07392 130.DD760' . 8732 2760. 115.00 4.06455 130.45290 . 8732 27E4. 116.'00 4.0D528 130.34910 . 9732 2603, 117.00 4.04609 130.24640 . 8732 2632. I19.00 A.03699 130.14460 . 873;'
' --26 5 6 , 1 l 'i. 6 0 4.02799 130.04380 . 8'/32 2000 120.00 4.OJ405 129.04400 , 873T 2904,,- 111.00 4.01020 129.S4500 . 8731 Fe/28, 12 .00 4.00144 12R.74700 . C732 2952. 173.00 3.99275 129,64990 . 8732 2976. 124.DO 3.98414 129.55360 . 8732
-3000. 175.00 3.97Db1 129.40820 . 3732 3024. :2 .00 3.96710 129.36360 . E732
~3v4 6. 127.00 3.95877 129.26990 . 8732 3 0'< 2. RE. W 3. . %O*; o 129.17690 . 8732 3096. 1:"9.00 3.94222 129.08430 . 8732 3120, 1 ~30. W 3. t?3405 128,99340 . G732 3144. 't 31, 00 3.92595 123,90280 . 8732
'3168. 132.00 3.91771 128.81300 *
. 8732 3197. 133.00. '3.90994 128,72090 . 37:2 3216u 134.O'; 3.90204 128.63550 . 6732 3240. J '.d . 00 3.80420 128.D4780 . E7~.2 3264. I 'l. b . 60 3.68643 128.46090 . 8732 3183. 1 ~. 700 3.87E71 128.37460 . 673*'.
3312. 13E. t U 3.67106 128.28900 . 8732 3336. 17".00 3,86347 128.20420 . 8732 3340, 1av. M O.85D94 128.11990 . 8732
.3386. 141.00 3.84846 128.03640 . 8732 3408. -142.00 3.84105 127.95340' . 8732 3432. 143.00 3.83369 127.8711O , 8732 3456. 144.00 3.8260Y 127.78950 . 8732 3480. 145.00 3.O1914 127.70950 . 8732 OLO4 146.00 3.811 % 127.62800 . E732 3528. 147.00 3.80481 127.54820 . 8732 3D52. 148.00 3.79772 127.46890 . 8732 35'76. 149.00 3.790e9 127.39030 . 8732 i
i 4
+ 4 . - < -
_.m_____-m.__ .m.--._____-___.-__.______m____ .o._ ___.____-mmhh-
y j g. pp ,_ ._.. .y. r_;.y.y . , - ._, .. , : , % ; m
6 Vermont. Yarlkee End Run !!~12 Month l Batch uo Don Decay Heat Cumm Heat Cumm # Bun Tc
-1 104' .11040' .11040 104 19.27 )
p '108 .11194 .22234 212 20.27 3 120 .12144 .34378 332 21.27 4 80 .07904 42262 412 22.27 5 96 .09261 .51543 508 23.27 6 96 .04042 .60535 604 24.27 7 54 .07725- .68310 688 25.27 G 136 .12211 .00522 824 26,27
~~9 328. .28755 1.09276 1152 27.27 !~~
10 30 .O?568 1.11844 1162 28.27
+
Assumpt i on s, t _
r . Power per B,.to r_il e ' <- 15.44197 MDTU/Hr
' Irradiation Time = 50040. Hours l: Recctor F w er =. 1660. MWt h Tine After Shutdown 5168540. Hour s or 7035.00 Days Re f er ente t Nt> REG-0800, Standerd Review Plan
.(1) Section 9.1.3 Epent Fwe). Foul Cool i ne cod C1 can-up 5 yrt emy 3 Rev. 2
-( 2 ) Branch Technical Position ASB 9-2 Renidud Detey Ener gy f or Li ght -Wat er- Reactors for Long Term Cooling, Flev 2 Spent Fuel ' Pool Luuling System Sc r vi ce Water Eye, tem Dat l e t' Tenper m. ur uts (F ) 92.5" C1.39 Inlet Temperatures (F) 97.51 (Pool Temp) 85.00 Weter Flow Hates (M) b/Hr ) .225 .175 Hect. Exchanger Surfece Area (SqFt) 500.O Heat Eucheoger Overal1 Conduct ance 375.O BTU /SqFt-H -F The two pass correction f act or , F= .573
u- .. .. . . . . . . . . .4 . . . . . . , . .
% r rnont Yenkece End Pun !! 12 Month Lc c c.y 'T i m ' Decm, Heet. Pool Tempcr etur u r .f er t or (Hour c ) ( D ey s.. i (t iB1 U / Hr ) ( deg r ams. F)
++*.*+ **+o+ too. 1.11844 97.51192 .0734
~~+ + * " *++**+ 1.11877 97.50486 .8731~~
~~++++ **++*+ 1.11830 97.50614 .8732~~
~~++++* ****** 1.11822 97.50462 .8732~~
~~+"*+* +"h" 1.11815 97.50404 .8732~~
~~v*+*n * + + + + , 1.11508 c;7. 50313 .8732~~
+o+r* .*"+** 1.118J0 97.50235 .E732
+ *+** +*,.. 1.11793 97.50151 .8732
+*++*t ++ "+ 1.11796 97.DOO71 . (;7 7?
n++++ +**+++ 1.11770 97.49987 .6732
~~+++y+ ++,+>+ 1.11771 97.49907 ,8732~~
++++++ ****+- 1.11764 97.49825 .C732
++a ,+ ++++++ 1.11756 97.49742 .8732
+++oy ++++e ;.11747 97.49660 .8732
++**++ *++w+ 1.11742 97.49579 .8732
~~++++ +,..+(< 1,11734 97.49497 .9732~~
~~.* + v + c + +*v*+* 1.j1727 97.49414 .873;~~
~~+ + .> u + + + , - - +~~
1.11720 97.49333 ,e737
+,++n o++++ 1.11712 57,492'1 .n732 l
~~++ +***** 1.I1705 97.49169 .0732~~
~~++**u +"+++ 1.11698 97.49068 .E735~~
~~++*** *****+ 1.I1691 97.49006 .8732~~
++v+++ +++v-- 1.11683 97.40Y:3 . E7L
>**+=a - " * *
~~1.11676 97.48842 .8732~~
~~H*+ +++++* 1.11669 97.48759 .8732~~
>*+*h ***++* 1.1 1 ci61 97.45678 .8732
++.*.+.<- H++++ 1 iI654 97.48596 .6732
**+++> 1.11647 97.48514 .8732 !
~~++k*++ ++++++ 1.11639 97.4E'433 .873:~~
~~++***+ 1.11632 97.40351 .R732~~
++++*+ . + o- *f70 1.1162D 97.48266 .873:
++.**+ o*"+ 1.11617 97.49187 .9732
++++++ ++**++ 1.11610 97.4810e . .E732 I i
-*+++*+ +wx*+ 1.11607 97.48024 .8732
++++++ +++++* 1.11595 97.47942 .[1732
++ +n* +. +++ 1.1150s 97,47Ee0 .87 7
+++*++ +*+++* 1.11581 97.47778 ,8/32 un+* +*++** 1.11573 97.47697 .E732
++++++ ++++++ 1.115e6 97.47614 .6732
+****- ono 1.11559 97.47533 .8732 H+++* ++++++ 1.11 % 2 97.47451 .E73; h***. +++*** 1.11544 97.47369 .2732
+*o++ ***+w 1.11537 97.472e8 ,8732
~~++++ +-***++ 1.11530 97.47205 .E 32~~
~~i. +**+*+ *****+ 1.11522 97,47124 .0732~~
.*****+ ****n/ N 1.1151b 97.47042 .8732 l **++"- *+++++ 1.11508 97.469o0 .8732 l
l 1
+++++* ***+*+ 1.11500 97.46879 .6732
++"++ ++*+++ 2.11493 97.46798 .E732
~~u+* ++++++ 1.11486 97.46715 .8732 i~~
~~i l~~
~~t. _ ______ ____________-__ _-___ -~~
- . - , . . _ ~ . . ,
ala cy&
Ver niorit Yeni.ee SFF Terr.per et ur c s 5 pent Fuel Pool Cooling System Service Water System t _m t l e t T emperatur es (F) 153.92 143.41 1 ri; ei Tempereitutcs (F) 199.35 (Poul Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Hect L o t.d ( MB'I U / Hr ) 10.222 Hee: Ecchenger Surface Area (3qFt) 500.0 Heat F.:t heno a D.er al l C on chtc t an t e 375.O Dtu/5qFt-H-The two pass cor rec ti on factor, F a .873 Ver mc.nt Yc il et BFP Temper at ur en r
Spent Fuel Pool Cooling System 5er vice Water System Outlet Tempc r rtures W) 151.12 141.09 lolet T c n.per 4.t ut en (F 294.7A (Pool Temp) E5.00 Water Flow Retes (hlb /Hr) .225 .175 Decey Hrut Lo.:- ' 91;T U/ hr ) 9.815 Heat D:chenger Eurface Area (SqFt) 500.0 liest En c h ar ego De r-r a l l Conduct once 375.O Ptu/EqFt-Hr-F Tlie two paws correction factor, F= .873 Ver murit Yunl.eu SF P Temper atur et Spent Fuel Fool Cooling System Service Water System Outlet lamperaturoc (F ) 148.95 139.25 In) et Temper at ur n (F) 191.14 (Fool T erop ) 85 00 Wster F1 4u lates (Mllt/Hr) .225 .175 Deccy Hent Load (MBTU/Hr) 9.493 Heet E.n cl un g er Surface Area (SqFt) 500.0 Heat E x c h en tj es Ove-.r al l Conductance 375.O Btu /5qFt-Hr-F The to pe w correc Li on factor, F= .875 Ver mont Yankee EFP temperatures Spent Fuel Pool Cooling System Service Weter System Outlet Temperaturera (F) 147.19 137.75 ]
Inlet Temperatures. (F) 189.21 (Pool Temp) 85.00 I
Water Flow Rates (M1b/Hr) .225 .175 j Decay Heat Loed (11bT U/Hr ) 9.231 J Heet E.Lchanger Surfece Area (SqFt) 500.0 l Heat Exchanger Overall Conduct ance 375. 0 bt u /SqF t-Hr -F j The two pass c or r et t i on factor, F= .873 l l
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1' Outlet ~lemperaturei, (F). it5.71 . 136.50 Inl et Temper atur ec - (F ). 185.77 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .275 Decay Heat Load -(MBTU/Hr) 9.012 Heet Enchanger Sur f ach Aree (SqFt) 500.0 Heat -. Ex c h ang er Over al 1 Cce,ductance 375.O' Btu /SqFt-He-F
, The.two pas.,s correction factor, F= .873 Wr esont Yankee SFP Temperatur ec Epent Fuel Fool Cooling System Servico Water Syrtem Outlet remporuturoc (Fi 144.45 133.43 Inl e t- Tekpr r a L ur ec (F) 183.68 (Pool Temp) 85.00 Water F)oH RJtes (M1b/Hr) .225 .175 i Detny Heal Lord (MBTU/Hr) 8.825 Heet Ei: c h e n g:.r Surface Area (SqFt) 500.0
. Heat Enthanper Overall Conduct ance 375. 0 Bt u/SqFt-He-F The two pass correction factor, F= .873 W' r mont Vent:ec EFP Temperature Spent Fuel Pool Cooling System Service Water System Outlet i trmper u t ur es (F) 143.34 134.49
. J nl et .Tempcr atur er, (F) 181.83 (Pool Temp) 85.00 i
Wat a Flow Pa. ten (M1 b /Hr-) .225 .175 Dec.a y Heat Load (MBTU/Hr ) 8.660 Hea1. C>.chenger Surface Area (SqFt) 5 ~'O. O He. at Enchanger Overall Conductance 375.0 Btu /SqFt-Hr-F
'The two pn , ccrrecti on f actor, F= .873 Vermont Yankee SFP Temperatur es i
Spent Fuel Pool Cooling System Service Water System )
1
' Outlet Temperatures (F) 142.34 133.64 I Inlet Temperatures (F) 180.18 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 !
Decay He+t Load (MBTU/Hr) 8.512
" Heat Exchanger Surface Area (SqFt) 500.0 Heat-Exchenper Over all Conduct ante 375.O Btu /SqFt-Hr-F The t wo paun correction factor, F= ,873
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Outlet Temperatures-(F) 141.44 132.87 Irij et Temperatures ( f' ) 178.68 (Poul Temp) 65.00 Water Flow Rates (M1b/Hr) . 225 .175 Decay Heat Load (MBTU/Hr) 8.378 He eL E>Le hanq6r Surface Area (SqFt) 500.O Heat Enchanger . Overs:11 Conductance 37U.O Btu /SqFt-Hr-F The two past cor rection f actor , F= .873 Vermont Yankee SFP Temperciures Spent Fuel Pool Cooling System Service Water Systcm
- Ou ll e. t Temp er at ur es, iF) 140.61 132.17 Irilet i temper at ur ec (F) 177.30 (Pool Temp) 85.00 Wetor Flow Eb lu ( Mi ti/Hr ) .225 .175 Du ay Heal Loco (ME'TU /Hr ) 8.25D Heat Exchencor Surface Area'(SqFt) 500.O Heet Exchangcr Omr all Cc,n d u c t an c e 370.0 Btu /SqFt-Hr-F Tht: two pas.s cor rection factor, F= .873 Vermoni Yunfec SFP Tempcraturec Spent Fuel Pool Cooling System Servico Water Syctem t
Out1et 1 empe r et ur et. (F> 139.84 131.52 Inlet Tetoper t ur es (F) 176.03 (Pool Ten.p ) 55.00 Water Flow RA ec (M1b/Hr) .225 .175 Decey Heat. Loed (MBTU/Hr ) 8.141 Heat. E n c l .an g er Eurface Area (SqFt) 500.0 Hent Enchanger O ver al l Conductance 375.O Bt.u/SqFt-Hr-F The two pees c or r t. c t i on f actor , F= .873 Ver mont. Yankee SFP Temper atures Spent Fuel Pool Cooling System Service Water System i'
Outlet Temperatures (F) 139.12 130.91 Inl et Temperatures (F) 174.83 (Pool Ten'p ) E5.00 Water Flow Rates (M1b/Hr) .225 .175 i i
Decay Heat Load (MBTU/Hr) 0.004 l Heat Enchanger Surface Area (SqFt) 500.O Heat Exchanger Overall Conductance 375. O Btu /SqFt-Hr -F The two pass correction factor, F= .873
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Vt.r mant veni:ce SF P 7 ernper at ur es Spent Fuel Pool Cooling System Servi c e Water System Out a et. Temperature (F) 138.4S 130.34 I nl et Temper atur es (F) 173.71 (Pool Temp) 85.00 Water Flow Rates (M1b/He) .225 .175 bec_ay Heat Lord (MBTU/He) 7.934 Heat E r. c h e r . g nr Surfeco Aree 'SqFt> 500.0 Heat Exchangur O ver ril l Conductance 375.0 Btu /SqFt-Hr-F The- two pass correction factor, Fn .873 Ver n ont Yantee SFP Temperatures Spent Fuel Pool Cooling System Servi c e Water System butlel Temper ature . (F) 137.81 129.79 I nl i 1. Tempe- stiir es (F) 172.65 (Pool T en;p ) 85.00 Weter Flow Ke t et (M1b/Hr) .225 .175 Decey H:ci Loed (t IBTU /Hr ) 7.EST Heat Exchenger Surface Area (EqFt) 500.O Heet Euchenrgr Ove-oll Conductance 375.0 Btu /SqFt-Hr-F The two pasis corr ecti on iactor, F = .673 Ver mor.t Yeniee SFP Temperature Spent Fuel Pool Cooling System Service Water System Outlet Temperatures (F) 137.20 129.27 Inle' Temper atu-t w (F) 171.63 (Pool T enip ) 85.00 W e.t er F1ow Rat es (H b/Hr) .225 .175 De c e.;y Heat Load (MBTU/Hr) 7.748 Heel Enchanger Surface Area (SqFt) 500.0 Heat Enchanger Over al l Conductance 375.0 Btu / SOFT-Hr-F The t wo pas s t.or r ec t i on factor, F = .873 Ver mont Yenkee SFP Temper aturcs Spent Fuel Pool Cooling System Service Water System Outlet Temperatures (F) 136.62 128.78 Inl et Temperatures (F) 170.67 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (MBTU/Hr) 7.662 Heat Exchanger Surface Area (SqFt) 500.0 Heat Enthanger Deeval) Conductance 375.0 Btu /SqFt-Hr-F The two pass correction factor, F = .873
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Ver mont Yonkee SFP Temperatur es Spent Fuel Pool Cooling System ser vice Water System Dutic' Temperatures (F) 136.07 128.31 Initt Temperatures (F) 169.75 (Pool Temp) 55.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Hat Load ( MBT U / Hr ) 7.580 Host Eschar.ger Surfeco Ar ea (5qFt) 500.O Hef t Enchunger OveralJ Conductance 375.O Btu /SqFt-Hr-F The two pass correction f sctor , F= .873 Ver mont Yankee EFP Temper atures 5 pent Fuel Pool Cooling System Service Water System Outlet l e,nper atures (F) 135.54 127.87 Inlet Tempo r at ur ew (F) 168.88 (Pool Temp) 85.00 Water F) ow Rates (MIb/Hr) .225 .175 Delay Hvet Lend (MblU/Hr) 7.502 Heat Ex c hanger Eurface Area (SqFt) 500.0 Heat E u c h anger Over all Coiiductence 375.0 Dtu/5qFt-Hr-F The two pass correction factor, F u. .873 Vermo'it Yeidee SFP Temperatures Spent Fuel Pool Cooling System Service Water System Outlet temperatures (F) 135.03 127.44 Inlet Temper at ur es (F ) 169.04 (Pool Temp) 85.00 Wat er - Flow Patet (Mib/Hr) .225 .175 De c c<y Heat Lued ( MBT U / Hr ) 7.427 Hoet Ench< enc;er . Sur f ece Area (SqFt) 500.O He c< t Enchanger Ove r al l Conductance 375.0 Blu/SqFt-Hr-F The t w:2 p m correction factor, F = .873 Ver n ont Yankee SFP Temperatures Spent Fuel Pool Cooling System Service Water System l
l Outlet Temperatur es (F) 134.55 127.03 l Inlet Temperatures (F) 167.24 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Herat Load (MbTU/Hr ) 7.355 Heat Exchanger Surface Area (SqFt) 500.0 Heat Exchanger Over all Conductance 375.0 Otu/5qFt-Hr-F The two pass correcti on f actor , F = .873
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.Ver toont. Yankee EF F leanper at ur es Spent Fuel' Pool Cooling System Seevice Water System Ou t. i cu t. Temperatures (F) 134.08 126.63 I nl et Temper at ur et. . ( F '> 166.47 (Pool Temp) 85.00 Water Flow f%tes (M1b/Hr) .22D .175 Decay Hect Load (MBTU/Hr ) 7.286
. Heat Exch nger L Surf eco Area (S;,Ft ) 500.O Heat Es t harigt:r Over al1. Coot:uct ente 37D. O Bt u/SqFt-Hr -F The- tWu-para cor e ect i on factor, F e .873 Ver munt Y4aikeo SWP Temper aturer Spent Fuel Pool Cooling System Service Water Syctem Dutlet Temper aturer (F) 133.64 126.26 In1et it n.pe r at ur ee iF)-
165.73 (Paul Temp) 85.00 Water Flow Raten (M1b/Hr) .225 .175 De-cov HetL Lued (MPTU/Hr > 7.220 Heal Exchanger Surface Aree (SqFt) 500.O H e r..t Exchcnget Dwr all Conductance 375. O Btu /SqFt -Hr r The'two pass correction factor, F= .873 Ver moni Vcnkee SFP 1emper atoreu' Spent Fuel Pool Cooling System Service Water System Jutlet Teiapor a tur es (F) 133.21 125.39 lril et Temperature (F) '16D.01 (Pool l erup ) 05.00 Watar F] cw R; ten < M113 /t 'r.) .225 .175 !
Deccy Heat Load (Mitt U/Hr .) 7.156 Heat Exchenger Surface Aree (SqFt> 500.0 Heat Exchanger Dverall Cariductance 375.0 Btu /SqFt-Hr-F The two parw corr ec tion f actor , F= .873 Ver mont Yankee SFr Temperatures Spent Fuel Pool Cooling System Ser vice Water System Outlet Temperatures. (F) 132.80 125.54 Inlet Temperatures (F) 164.33 (Pool Temp) 85.00 Water Flow Rates (M]b/Hr) .225 .175 Decay Heat Loed (MBTU/Hr ) 7.095 Heat Exchanger Surface Area (SqFt) 500.0 Heat Exchanger G eral1 Conductance 375. O Btu /SqFt-Hr -F The two pass correction factor, F= .873 i
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Out. l et temperatures (F) 132.40 125.21 I n) ct T eniper at ur es (F) 163.67 (Pooi T erop ) 85.00 i
Weter Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (t1BT U / Hr ) 7. O'f.6 i
Heat C x c.h e ng er Gutface Area ( Sq F t. ) 500.0 Heat En t hanger Overall Conductance 370. O Btu /Sqct-Hr -F Ti> e two piss cor r ec t i on factor, F = .873 V(-r hont YonEce FFP Temperat uree Spent Fuel Poul Cooling Systen, Ser vice Water System Ou tl ei temperatures (F) 132.02 124.EG Inl ei T evoptr ct or .ct (F) 163.03 (Pool Terop ) EE.OD Wa t err Flow Faten (M1b/Hr) .225 .175 Dicey He~I t oe.G (NB1U/Pr) 6. 97 r, Heat Enchang-r Sur f ece Area (SqFt) 500.0 He e4 t E :c horep t D w all Conductance 37D.O Btu /9qFt-Hr-F ,
The two pass t arretti on fector, Fa .873 Ver nan it. Yonlee SFP 1 erger at ur eu Spent Fuel Poul Cooling System Service Water System Outlet 1 empor atur es, (F> 131.65 124.57 In]ct T eruper a t ur ee (F) 162.43 (Fool T enip ) 85.00 Water Floo Retes (htb/Hr) .225 .175 Decay Heat L.oed (t1BT U/Hr ) 6.925 Huat f ,,clianger Surface Ar ea (SqFt) 500.0 Heat Exchanger Over al l Conductance 375. 0 Btu /SqFt-Hr -F Tlio iwo peus toire tion f sc tor , F = .873 Ver niont Yankee SFP Temperatures Spent Fuel Pool Cooling System Service Water System Outlet Temperatures (F) 131.30 124.27 I nl et Temperatures (F) 161.84 (Poel Ternp ) E5.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Loed (MBTU/Hr ) 6.872 Hewt Exchanger Surface Area (SqFt) 500.0 Heat En c h anger Over al l Conductance 375.O Btu /SqFt-H"-F The two pass tor rt*cti on factor, F= .873

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Ver n<ont Yankee BFP Teraper at ur ec Spent Fuel Pool Cooling System Service Water System Ou t l e t. Temperatures (F) 150.96 123.98 Inl et Ternperatur es (F ) 161.28 (Pool Teop) 80.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heet .Lcad (MD1U/Hr) 6.822 Heat. Exchanger Surf ac e - Area ' (SqFt) 500.O Heat Exchanger Dvet id 1 Conductance 375. 0 Bl u/SqFt-Hr -F The two pens correction factor, F= .873 Ver mon 0 Yankee GFP Temprr etures Spent Fuel Pool Cooling System Service Water System Outlet Temperatures (F)- 130.62 123.70 I nl et 7 emper atur es ' (F )~
2 360.72 (Pool Temp) 80.00 l
l -Water. Flow Retes (fil b /Hr ) .225 .175
[, Dec t y ' Hs.M L oaci _. (t1DT t1/H) > 6.772 Hee L Enchengm Surface Aree (SqFt) 500.O Heat Fuchaliger Dvere?1 Coriduct arece 375.O Btu /SqFt-He-F The two pisss correction factor, F ra .873 Ver cion t Yankee FFP Temper at ur cc Spent Fuel Pool Cooling System Service Water System
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Uutlet E mperat.oren (F) 130,31 123.43 Inict Ternper atur eu - (F ) 160.19 (Pool Temp) BD.00 Weter Flow-Ratet (M1b/Hr) .225 .175 Decay Heat Load (MBTU/Hr ) 6.725 ,
Heat Enchanger Sur f ace Area (SqFt) 500.0 Heat Exchanger over all Conductance 370.0 Btu /SqFt-Hr-F The tua p nv c or rect i on factor, F= .873 Ver inan t Vatikee SFP Temper atur ew Spent Fuel Pool Cooling System Ser vice Water Sytst em Outlet temperatures (F) 130.00 123.17 Inl et Temperatures (F) 159.69 (Pool Temp) 85.00 Water F1ow Rates (M1b/Hr) .225 .175 Decay Heet Lead (MDTU/Hr) 6.6EO Heat Exchanger Surface Area (SqFt) 500.0 Heat Exchanger Over all Conductance 375. 0 Dtu/SqFt-Hr -F ,
The two pass correction factor, F= .573 4
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, Spent. Fuel Pool _ Cooling Systern Service Water System Dutini Temperatures (F) 129.71 . 122.92 Inlet Temperatur t c (F) 159.20 (Pool Ternp ) 8D.00 Water Flow Rates ( Mi t; /Hr ) - .225 .175 Decay Heat Loed (MBTU/Hr) 6.636 Heat Exchanger Surface Aree-(SqFt) 300.O Heat Euc hro ege Overall Cencluc t ar ec e 375. 0 Btu /SqFt-Hr -F 7
-The two pase.correctionLfactor, F= .873 Vernic nt Yani-ee EFP Temper atures Spent Fuel Pool Cooling Sye: tem Service Water System Dutleti 'I woperaturer. (F) 129.42 122.67 Inlet Ten.pe r at ur co (F) 1DE.72 (Pool Temp) 80.00 Water Flow Rate. (hlb /Hr) .225 .175 Det oy Hvet Luod (tmTU/Hr ) 6.D93 Heat. Exchanger Surface Area (SqFL) 500.0 Hea t. , Ex c hanger Dveral) Conditct a.nce 373.0 Btu /SqFt-Hr-F ,
'The two pass correction f ac t or' , F= .G73 Ver rnc'n l Yarikee SFP Teraper atur ec
' Spent Fuel Pool Cooling System Service Weter System Out) et lemperaLurea (F) 129.14 122.44 Inlet Ternper atur er (F) 158.26 (Pool Temp) BD.00 Water Fi os Re.tes (M]b/Ur) .225 .175 Derey H>at Loat: (MBTU/Hr) 6.552 .'
-Heat EE.chinger Surface Area (SqFt) 500.0 Heat Exchanger Over all Conductance 375. 0 Btu /Sc;Ft-Hr-F The two pasu <orrectica fector, Fr .873 Ver toont Yankee SFP Temper atures Spent _ Fuel Fool. Cooling System Service Water Syst ern Ou tl e t Temperatures (F) 128.88 122.22 I nl et Temperatures (F) 157.82 (Pool Temp) 85.00 Water Flow Rates (M)b/Hr) .225 .175 Decay. Heat Load (MBTU/Hr ) 6.513 Heat Exchanger Surb.ce Area (SqFt) 500.0
' Heat Exchanger Overall Concluct ance 375. 0 Btu /SqF t-Hr--F ,
The two pass correction factor, F= .873 l l
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Vertnont Yahkeu SF'P Temper aturec Spent Fuel Fool Cooling System Se-vice Water System l
.. l Outlet Temperatures (F) 128.61 121.99 ,j In)et Temperatures (F) 107.39 (Pool Temp) 85.00 ]
Water Flow Rates (M1b/Hr) .225 .175-Det.ay Heat Load ( MBT U / Hr- ) 6.474 Heat Exchanger Surface Area (SqFt) 500.0 Fleat Exchanger Over al 1 Conducience 375. O Dtu/SqFt-Hr -F The two paos correction-factor, F= .873 Ver n,ont Yankee SFP Temperatures Spent Fuel Fool Cooling System Service Water Systee,
__.________________ 8 Outlet Teinpera tur es ' (F) 120.37 121.78 Ird el Temper i t or et (F) 156.97 (Poul Temp) 85.00 Water Flow Flat es (Mlb/Hr) .225 .175 Decoy Heat Loed (MDTU/Hr ) 6.437 HcaL Enchanger Surface Area (SqFt) 500.O Heat - Ex c hangrr D arn11 Conciuttance 375. 0 Bt u/SqFt-H' -F The two paus correction factor, F= .873 Ver omnt Yankee SFP Temper atur et Spent Fuel Pool Cooling System Service WP.ter. System Outlet Temperature (F) 120.12 121.58 Inlet Teniper at ar en (F) 156.57 (Pool Temp) 85.00 Water - Flow Retes bD b /Kr ) .225 .175 becay Hcat Load (MBTU/Hr ) 6.401 Heal thchanger Sur f ace Area (SqFt) 500.0 Heat Exchanger Over all Conduct ante 375.O Btu /SqFt-Hr-F Tlie twa pese con ection factor, F= .873' Ver mont Yanhee SFP Temperatur en Spent Fuel Pool Cooling System Service Water System autlet Temperatures (F) 127.89 121.38 Inlet Temperatures (F) 156.18 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Dec.ey Heat L. cad (MBTU/Hr) 6.366 Heat Exchanger Surfste Area (SqFt) 500.0 Heat Exchanger Over all Conductance 375. 0 Btu /SqFt-Hr -F The two pass, correction factor, F= .873 l
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' Outlet Temperature (F) 127.66 121.18 Inlet Teoperatur es (F') 105.80 (Pool Temp) 85.00 l' Wate:r - Fl ow Rates ~(M1b/Hr) .225 .175 Dtcuy Heat Load (MBl u/Hr )- 6.332 Heat Exchanger Surface Area-(SqFt) 500.0
' Heat Exchanger C'verall. Conductance 375. O Btu /5qFt-Hr -F The two pass correction factor, F= .873 Vermont Yankee SFP Temperatures Spent Fuel Pool Cooling System Service Water S ys tern Outl et Temperature (F) 127.44 121.00 Inl et Ternper aturet;- (F)- 155.44 (Pool Temp) 85.00 LJater Flow Rates (M1b/Hr) .225 .175 Decey Hee.,t Lued (MB7 U/Hr ; .
6.300 Heat Exchanger Surface Area (SqFt) 500.O Heat Exc hange r Dver ell Conciuc t aric e 375.0 Btu /5qFt-Hr-F ,
. The two pass cor rection ' f actor, F= .873 Vermont Yankee EFP Ternper atur es Spent Fuel Pool Cooling System Service Water System Ostlet Tempc-ratures (F) 127.23 120.82 I tal et Ten.per ct ur et. (F) 155.08 (Pool Temp) 85.00 Water Flon Rates (M1b/Hr) .225 .175 Decay Hebt L;oad (MBTU/Hr) 6.268 ,
Heat Exchanger. Surface Area (SqFt) 500.0 Heat Exchanger Overall Conductance 375.0 Btu /SqFt-Hr-F The twa pand correction factor, F= .873 Ver;nont Yankee SFP Temperature
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Spent Fuel Pool Cooling System Service Water System i
1 Outlet Temperatures (F) 127.02 120.64 !
Inlet Ternper at ur es (F) 154.74 (Pool Temp) 85.00 l Water Flow Rates (Mib/Hr) .225 .175 Decay Heat Loact (MBTU/Hr> 6.237 Heat Exchanger Surface Area (SqFt) 500.0 Heat Exchanger Over all Conductance 375.O Btu /5qFt-Hr-F The two pasu correction factor,-F = .873 I
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Outlet ~ Temperatures (F) 126.82 120.47 Inlet T emp er atur es, (F) 154.40 (Pool Temp) 85.00 Water' Flow. Rates ( Mi ts / Hr ) .225 . 175 Decay Heat Load (MB1U/Hr) .
6,207 Heat. Exchanger Surface Arca (SqFt' 500.O Heat Enthanger Ch ercd 1 Conductance 375. O btu /SqFt-Hr -F T h e+ two pass cor r ection f actor , F= .873 Ver niont Yonkee SFP Turnperat or er Spent Fuel Pool Cooling System Sersice Water S y s t.um Outlet Temperatures (F) .130.02 123.18 ,
In)et l en.per et or et- (F > 159.71 (F col Tertip ) 65. DO Water Flaw Ratub ( Ml L, / Hr ) .225 . 175 Deter Het,t L ond (tlBT U/ Hr > 6.692 Hea.t Exchanger Surface Aree (SqFt) 500.0 Heat Exchanger Ov er ed l Conduttonce 375. 0 Btu /SqFt-Hr -F
.The tuo pass corr ecti on factor, F = .873 Ver moni Yankee EFP Tenger at urec Spent Fuel Fool Cooling System Service Water Systern i I
Dutlu4-Tempetetures (F) 126.43 120.14 Inlet Temperature (F) 153.77 (Pool Temp) 85.00 .1 Water Flou Ratoa ( M1 b / Hr -) .225 . 175 Decay Heat Load (MUT U/Hr ) 6.150 ,
i Heat E::chan ger Suifece Area (Sa rt) 500.0 Heat En c h an g e- Overt 11 Conductance 375.0 Btu /SqFt-Hr-F The two pas *. co"rection factor, F= .875 Ver nion t. Yankee SFP Temper atur es, 1
Spent Fuel Fool Cooling Systern Sur vico Water Systen Ou t l' et Temperatures (F) 126.25 119.99 Inl et Temper stur es (F) 153.46 (Pool Terup ) 85.00 Water F1ow Rates (Mib/Hr) .225 . 175 Decay HeatLoed (MBT U/Hr ) 6.123 Heat Exchanger Surface Area (SqFt) 500.0 Heat Exchanger Overo11 Conductance 375.0 Btu /SqFt-Hr-F The two pess correction factor, F= .873
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%-munt Yeniee SFP Terape r atur es Spent Fuel Pool' Cooling. System servic.e Water Syst<a
. - - . . - - - - - - . . - . . . . - - . . . . ~ - - . . , . - . , . - - - . . . . - - - . . . . - . . . .
Outlut Temperature (F). 126.07 119.83 sin 1et. Temperatures (F) 153.16 (Pool Temp) 85.00
. Wate: Flow Rates (M1b/Hr) .225 .175 Decay Heat Lued - (MDTU/Hr ) 6.096 Huat E::: hanger . Sur f aco Area (SqFt ) 500.O Heat Enc.i mnyer Oserall Conductarice 375.O Btu /SqFt.Hr-F The two pass correction facter, F == .873 Vor mont Yankee SFP Temperatur en Spent Fuul Pool Cooling System Sc.rvi ce Wat er Syst en Dutlet Ten persturen (F) 125.89 119.69 In16t Temper alt..r es !F) 152.87 (Pool' Terop ) 80.00 W a t en - Flow Ratos (M1b/Hr) .225 .175
.DecLy H. cat Lo+se (t1 BTU /H- ) 6.070 Heat E::changur Surfece Area -( Sq F t ) 500.O Hoed Euc huhtie Over 01) Conductance 375.O Btu /SqFt-Hr.-F The two pc.es correction factor, F= .873 Ver ment Yer ti:ee SFP Ternper at ur et Spent Fuel Pool Cooling System Service Water System Outlet Teinper aturus (F> 125.72 119.54 Inl et . Temper at ur ov.- (F) 152.59 (Pool Temp) 85.00 Water-Flow Ratec'(hlb /Hr) .225 .175 l Dotay Heat Load (MBTU/Hr> 6.045 '
Heat E,vhenger Suriece Ar ea (SqFt) 500.O Heat Ex c henger Overall Conduct +<nce 375. 0 Bt u / SqF t-Hr -F The two pese correction factor, F= .873 Vermont Yankee SFP Temperatures Service Water System
'j Spent Fuel Pool Cooling System ;
Outl et. Temperatures (F) 125.56 119.40 Inlet Temperatures (F) 152.31 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (Ml;T U/ Hr ) 6.020 Heat'Enchanger Surface Area (SqFt) 500.0 Hest Exchanger Over all Conductance 375. 0 Bt u/SqFt-Hr -F The two pass corr ection factor, F= .873 j l
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-Outlet Temperatures.(F) 125.40 119.27 Inlei Temper atur es (F) 152.05-(Pool Temp) 85.00 1 i
Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Lcad . (MBTU/Hr ) 5.997 ;
Heat Enchangur Surface Area .(SqFt) 500.0 j Heat Enchanger Overall Conductance 375.0 Btu /SqFt-Hr-F The two pass correction factor,'F u .873 Vermont Yant:ee SFP Temperat ur oc Spent Fuel Fool Cooling System Eorvice Watter System Outlet Temperatures (F) 120.24 119.13 I Intet Temper atur es (F> 151.79 (Peol Temp) 85.00 Water F1ow Rates (M1b/Hr) .225 .175 Deca) . Heat L ( aad (MPTU/Hr) 5.973 Heat En hanger Surface Aree (SqFt) 500.O
~Hent Enthangtr Overall Conduct noce - 375. O Btu /SqFt --Hr -F The two pass correction factor, F= .873 Ver mont Yankee EFP Temper atures Spent Fuel Pool Cooling System Service Water System Outlet T empc r r,t ur e b (F) 125.09 119.01 I nl et i emper at ur t. t (F) 101.54 (Peal Terrp ) 85.00 Wat er Flow , Rates 011 t. / Hr ) .225 .175 :
Docay. Heat Load (MBTU/Hr> D.951 ,
Heat Exchengiar Surface Ares (SqFt) 500.0 Heat Exchangar Overall Conduct ance 375.0 Btu /SqFt-Hr-F The twc pecs corr ect2 cn factor, F= .873 Wer mont Yankee SFP Temper atures i
Spent Fuel Fool Cooling System Service Water System Dotlet Temper aturec (F) 124.94 118.88 iI nl et Temperatures (F)' 151.29 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (MBTU/Hr) 5.929 Heet Exchanger Surface Area (SqFt) 500.O Heat Exchanger Overall Condut.tance 375. 0 Dtu/SqFt-Hr -F The two pass correction factor, F= .873 l
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.Ver mont Yankee SFP Tc mper atur es X Spont' Fuel Poc1 Cooling Sybtem Service Water System OuiTet lemperaturvs (F) 124.79 118.75 Jnitt Temperatures (F) 151.05 (Pool Temp) 85.00 Whter Flow _ Rates '(M1b/Hr)
.225 .175 Decay Heat Load (MBT U/Hr ) _
5.907 heat Cachanger Surface' Area (SqFt) 500.0
, Heat ' Eschanger Over all Conduetarice 375. O Otu/SqFt-Hr -F This two pees cutrectiori factor, F= .873
.W:r roont Yrinkee SFP Tempor et urce Spont Fuel Pool Cooling System Scervi ce Water System i
.r Outlet Temperaturou (F) 124.65 118.63 Inlut.Temptu aLures (F) 150.01: (Poul Tempi 85.00 3 1
1
-Water Fluw Pates (M1b/Hr) .225 .175 l Dec n Hect Lord WOTU/Hre 5.886
. Heat Exchanger Surface Area (SqFt) 500.0 heb.t Enchanpr Deerall Con duc t are ce 375. 0 Btu /SqFt-Hr -F The two pass corr ection factor, F= .873 Ver toont Yarskev SFP . Teuper at ureb Spent Fuel Pool Cooling System Service Water System l Out l es Temperatures (F) 124.51 118.51 I r.] c't Temperature (F) 150.59 (Pool Temp) D5.00' Water Flow Retes (M1b/Hr) .225 .175 i Dc. cay He sit Load (MBT U/Hr ) 5.065 ,
Heal. - En ch ang er Surface f<rea (5qFL) 500.0 Heat Ex c h angur Ove all Conductance 375.0 Btu /SqFt-Hr-F Tiin tuo pass correction facto , F= 373
' Vermont Yankee SFP Temperatures i
i Spent Fuel Pool Cooling System Service Water System Outlet. Temperatures (F) 124.38 118.40 Irelet Temperature (F) 150.35 (Pool Temp) 85.00 ,
1 Weter Flow Rates (M1b/Hr) .225 .175 Decay. Heat Load.(MB7U/H-) 5.845 !
Heat Exchanger Surface Area (SqFt) 500.O Heat Exchanger Over all Conductance 375.0 Btu /5qFt-Hr-F The two pass correction factor, F = .873
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'Ver munL Yankeu. SFP 'Iemper at uren Spent Fuel Pool' Cooling System - Service Water System Out1 tt Temperatures (F), 124.24 118.29 E 'Iril et Temperatur es (F) ,150.13 ( F r. s remp) 85.00 W'ater Flow Rates (M1b/Hr)- .225 .175
- Lec ar Heat Load (MBTU/Hr ) - D.825
- Hu dt Exchangor Surface Area.(SqFt) 500.0 l
Hr.at En c hangccr ' O ver al l Coriductance 375. 0 Utu/SqFt-Hr- F .
l The two. pass correction factor, F= .873
' Ver rm ent Yankee SFP Temperaturett, Spent. Fuel Pool Cooling Syntem Service Water system;
. - - - - - - - - - - _ -.-..-.-----_---- ~,
4 Outletc Tanper ;.turec (F) 124.11 .
11E.18 Ini c t Ten pe r t4u eu - (F) '149 92 (Pool Ten p ) 85.00
. Water Flow Retes (M1b/HtU) .225 .175 I n ..< y He a t L uw.i (tIDT U/ Hr ) 5.PO6 Heal. Exchange. Surface Area (SqFL) 500.0 Heat ' Enchanger Cvc+r all- Conductance 375. O Ptu/SqFt -Hr -F
~~The two pasz dorrection factor, F= .873 Vu mont Yankee SFP Temperat ures.~~
Spent Fuel Pool Cooling System Servico Water System 10utlet temperatures :F) 123.99 110.07 Inlet Temperatur eb (F ) - 149.71 (Pool T ernp ) 85.00 Watsr F10o F a l.e s (il[b /Hr ) .225 .175 Decay Heat Load (MBTU/Hr) 5.787
Heat ' Exchangur Surface Area.(SqFt) 500.O Heat Epchanger Dverall' Conductance 375.0 Btu /SqFt-Hr-F Tne tua p m t. corroction factor, F= .873 Ver mont Yankee SFP Temperatur es Spent Fuel Pool Cooling System Service Water System l Outlet Temperature (F) 123.87 117.97
' Inl et Temperatures. (F) 149.51 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (MBT U/Hr ) 5.769 Heat Exchanger Surface Aree. (SqFt) 500.0 i 37D.0 Btu /SqFt-Hr-F l He..it Exthanger Overall Conductance The two pass currection f actor , F= .873
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I Outlet temperatures'IF) 123.74 117.86 lolet . T6mperatur et, (F) 149.30 (Pool Temp) 85.00
' Water - Flow Rates (M1b/Hr> .225 .175 Decay Heat Load-(MBTU/Hr) 5.701
- Heat E"3 changer Sur f ace Area (SqFt) 500.0 Hunt Ex ch an g er. Over cil- Conductance 375. O Btu /5qFt-Hr -F The two pess correction factor, F= .873 Vermont Yentec 5FP Temperat ur es Spent Fuel Pool Cooling System Ser vice Water System Outlet Teniperatures (F) 123.62 117.76 Inlet Temperature (F). 149.10 (Poul Temp) 05.00 Water - Fl ow Raton (M1b/Hr) .225 .175 Dechy~Hect Lui d W.DTU/Hr) 0.733 Heat'Exchatyger Surfsee Area (5qFt) 500.0 Heat Enthanger Ove-all Conductance 370.O Blu/SqFt-Hr-F The two pass correction factor, F= .873 .;
Vertoont Yantee SFP Temper atures spent Fuel Fool' Cooling System Service: Water System Outl et . Temp-ar e tures ' (F) 123,51 117.66 Inlet Temper atur ea (F) 148.91 (Pool Temp) 85.00 Wat er F1ow Rates (M1b/Hr] .225 .173
. Deca >LHeat Load '(Mb7U/Hr') 5.716 ,
Heat.ENchanger.Gurface. Area (SqFt)~ 500.0 Heat Enthanger Overall Conductance 375.O Btu /SqFt-Hr-F The two peso correction fector, F= .873 Ver mon L Yankee SFP Tempertturer.
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Spent Fuel Pool Cooling System Service Water system Outl et Temperatures (F) 123.39 117.57 j Inlet Temper atures (F) 148.72 (Pool Temp) 85.00 LWater Flow Rates (M1b/Hr) .225 .175 Decay Heat Load-(MBTU/Hr) 5.699 Heat Exchanger Surfece Area (Sq F 't ) 500.0 Heat Exchanger Overall Conductance 37D.0 Btu /SqFt-Hr-F ;
The two pass correction fcctor, F= .873 i
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4 Ver trerit Yani.ee SFP Tempereturen Spent Fuel Pool Cooling System Gervice Water System Go L 1 et Temperatures, (F) 122.86 117.11 Inlet Ter.aperatur cn (F) 147.84 (Pool Temp) 85.00 Weit er F1ow Rates (M1b/Hr) .225 .175 Decay Heat L oeci (MDTU/Ht ) 5.620 Huat Enchanger Surf ac e Ar N (SqFt) 500.O Heet Ex c h an ger Overall' Conductance 375. 0 Dtu/EqFt -Hr -F The two pess corruction factor, F= .873 W r mont Yonkee SFP Temper etur en Spont Fuel Pool Cooling System Service Wat.or System j
- - - - - - _ - . - - - - _ _ . _ i i
Outlet Teraperatur en (F ) 123.17 117.08 16 0 et 1 etapt r at iir et . (F> 14B.3D (Pool Temp) 85.OC We.ter Flow Rotat ( M I L, / Hr '/ .225 .175 Deay hem L c, a d (ME:T U/Hr ) 5.666 Hut Ex c hasiger Suriace Atee (SqFt) 500.O Heitt F.a c hang er Over al 1 Contiutt ente 7,75. O Btu /SqFt -Hr -F The t uo p e ., correction fattor, F= .873 V4 r niont Yankee FFP Temper at ur en Spent Fuel Pool Cooling System Service Water System OutleL Temp aratur ee (F) 123.06 117.2(?
Inlet Temper at ur vs. (F) 149.17 (Pool Temp) EiD, 00 W.a t es - F l c.w P a t e s, (Mlb/Hr) .225 .175 Decey Heat Loed (NUTU/Hr ) 5.650 Heet D:thanger Eurface Area (SqFt) 500.0 Heat Et;thnnger Over all Conductance 375. 0 Bt u/SqFt-Hr -F The t w o p>3s.n :.or r e n t i on factor, F= .873 Vet inunt Yani:ee SFP Temperatur et Spent Fuel Pool Cooling System Service Water System Outlet Te.nparatures (F) 122.96 117.19 Ird et Temperatures (F) 148.00 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (MBT U/Hr ) 5.634 Heat Exchanger Surface Area (SqFt) 500.0 Heat Enthanger Over all Conductance 375. 0 Ltu/SqFt-Hr -F The two pass correction factor, F= .873 4 l
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Outlet Temperatures _-(F) 122.85 117.11
-Inlet Temper atur eb (F) 147.83 (Poul Temp) 85.00 Waticr Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (Mi;T U/ Hr ) 5.619 Heat:Euchanger Surfaco. Area (SqFL) 500.0 Heat. En thanger Over all Conductance 375.O Blu/SqFt-Hr-F The two pese correction factor, F= .873 Wrmont Yankee SFP T emper atur es Spent Fuel Fool Cooling System Service Water System
' Outlet temperatures (F) 122.75 117.02 Inlet Temper atur n (F ) 147.66 (Pool Temp) BD.00 Water Fl ow Retes (M1b/Hr) .225 .175 Decay Heat Loed (MDTU/Hr) '5.604 Heat Exchanger Surface Area (SqFt) 500.0 Heat Enchange Omr c1 ? Conduct ante 375, O Bt u/SqFt-Hr-F The -two pat s corr ection factor, F= .873 Wr ten L Vesol et. SFP Temperctores Spent Fuel Poal Cooling System Service Water System
_ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ r Outlet Temperatut es (F) 122.65 116.94 Inlet Temper atur eu (F) 147.49 (Pool Terop) 85.00 i
Water. Flow Ra t e.a (hlb /Hr) .175 i
.225 Decay Heat Load (MBTU/Hr) 5.589 Heet Enchanger Susaco Area (SqFt) 500.0 Heat Exchanger Over all Conductance 375. 0 Btu /SqFt -Hr -F l The two pass cortection factor, F= .873 l Ver mont Yankee SFP Temperatur es, Spent Fuel Pool Cooling System Service Water System Outlet. Temperatures (F) 122.55 116.85 Inlet' Temperatures (F) 147.32 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (MBTU/Hr> 5.574 !
Heat Exchanger Surface Area (SqFt) 500.0
. Heat Exchanger Overall Conductance 375.0 Btu /SqFt-Hr-F The two pass correction factor, F= .873 4
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Spent Fuel Pool Ccioling System Service Water System j Uut 1 rit lemperaturec (F) 122.46 116.77 2 ni nt Temper atures (F) 147.17 (Puol Temp) 85.00 Water F' low Rates (M1b/He) .225 .175 l Du.ey Heat Load (MDTU/Hr) D.560 Hcal Exchanger Surface Area (SqFt) 500.0 i Heat Enthanger Over al l Conductencu 775.0 Etu/SqFt-Hr-F The two pass correction. factor, F= .873.
Wr na ent Yankee SFP Temper at ur et
$pe:it Fuel Fool Cooling E'yttem Ser vice Water Syntem Outlet T emp er a t ures. - (F ) 122.36 316.62 !
' Inlet Temper ut ures, (F ) 147.01 (Pool Temp) GD.00 f 1
Water F1aw RStas ( feil b / Hr ) .225 .175 1 Tiet w HWt Lund ( MDi !/ H* ) 5.546 )
500.0 Btu /SqFt-Hr-F i Heat Exchanger Owd elb DonJucLqhto The two pacu Jorrection. factor, F - .873 Wr vnunt Yankee SFP Ternper aturen Spent Fuel Pool Cooling System Eervice Water System -
1 Outlet "l etop er a ,.ur es (F) 122.27 116.61 l Inlet T en.pt r atur eo (F) 146.86 (Fool Temp' 0D.00 Water Flow Rates (M1b/Hr ) .225 .175 Decoy Hect Lept: (MBT ti/ Hr :- 5.532 Heet Exchanger Surf ace Arua (SqFt) 500.0 l Heat Exchanger. O mrall Conduc t ance 37D. 0 Blu/SqFt-Hr W i The two paru correction factor, F= .873 L Ver rnunt YonPre SFP Temper aturen l
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Outlet Temperatures, (F) 122.17 116.53 l Inl et Temperatures (F) 146.70 (Poul Temp) 85.00 Water Flow Raton (Mib/Hr) .225 .175
' Decay Heat Load (MDTU/Hr) 5.518 Heat Exchanger Surface Area (SqFt) 500.O Heat. Exchanger Over al l Conductance 375. O Btu /5qFt-Hr -F The two pess correction factor, F= .873 l
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Verniont Yankee End Run Spent Fuel Pool Cooling System Service Water Sys,t ern j 1
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l Out let' Temperatures (F) 120.96 115.48 j Inlet Temperatures-(F) 144.67 (Pool Temp) 85.00 {
Wat er: Fl o'w Rat es . (M1 b /Hr ) .225 .275 Lecoy Heat Leed.(MBTU/Hr) 5.334 Heat Exchanger Eurf ace Area (SqFt) 500.0 Heet Exchanger Overall Conductance 375.0 Blu/SqFt-Hr-F Tho two penc cor rection f actor , F= .873 Ver<ront Yeulme Fnd Run r
Spent Fuc1 Pool Couli ng Systern Eervice Wate- S y s t ern Outlet l emper at ur e s' (F) 119.91 114.61 Inlet Temper stur es (F) 142.94 (Pool Temp) 85.00
. Water Flow Rstes (M1b/Hr) .225 .175 Dticay H%i Loed (MG'iU/Hr ) 5.182 Heet Exchanger Sur-(nce Area.(SqFt) 50 '.). O !
Hea.t Exchergtr Overa;l Con d u.:t e n cia 375. C Btu / 5qF t-Hr -F , l The two-pass corr et ti ore facto ,'F = .8'70 Vermont nhee End Run Spent Fuel Pool Cooling System Service Water System Outl et -l tat per at ur es (F) 119.02 113.86 Inlet Te;mperaturen (F) 141.47 (Pool Temp) E5.00 Wet.er Fl ow hetes (til b /Hr ) .22D 175 Decay Hoat Load (MBTU/Hr)' 5.050 .
Heat Exchanger Euri at.e Ar ea- (SqF t ) 500.0 Heat Exchanger Overe11 Conductance 375.0 Btu /SqFt-Hr-F The t wo p o t.4, cor r et t i on f ac.t or , Fn .573 Vermont Yankee End Run Spent F uel Pool Cooling Systern Eer vice Water System
- - - - - - . - - - - - _ . - = = - - - - - - - - - - - ---- . - -------- --
Uutlet l ernper atur et. (F) 119.24 113,19 Inlet Temperatures (F) 140.1'7 (Pool Temp) 85.00 Water Flow Rates (M1b/Hr) .225 .175 Decay Heat Load (MBTLI/Hr ) 4.934 Heat Euchanger E,urface Area (SqFt) ,500.0 Hwat Er. changer Overall Conductance 375.0 Btu /SqFt-Hr-F T he t wo pass cor r ec tion f act or , F= .873
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l Wer marri. Yankee End Run Spent. Fue] Fool Cool i ng Syst.ern 5ervice Water S ye.t ein Out l et Temperat ur er. (F) 118.08 113.03 I n l e i. Temperature (F) 139.88 (Pool Ten.p ) 85.00 We t er Flow Rates (Mi t) / Hr ) .225 .175 Decey Heat Lead (MBTU/Hr) 4.906 Heat Ex c h e.t n g er Eur iete Aree ISqFt) 500.O Heat Exchanger Cverall Conductance 375.0 Btu /SqFt-Hr-F The t wo p a s t+ correct i on f att or , Fu .873 I
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ENCLOSURE 3 MISCELLANE0US DOCUMENTS A e M
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SEP 3 01986 MEMORANDUM FOP: Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing FROM: L. G. Hulman, Chief -
Plant Systems Branch Division of BWR Licensing A
~~SUBJECT:~~
VERMONT YANKEE, RERACKING OF SPENT FUEL POOL, REVIEW AND REQUEST FOR ADDITIONAL INFORMATION (TAC #61351)
The Plant Systems Branch has conducted a review of the proposed spent fuel pool reracking for Vermont Yankee. The proposed reracking is described in the Vermont Yankee Nuclear Power Corporation submittal of April 25, 1986.
Attached is a request for additional infonnation which is needed to complete i the review. We are prepared to discuss these issues with the licensee in order te expedite our review.
W
.G Hulman, Chief k Plan Systems Branch Division of BWR Licensinc
~~Enclosure:~~
As statec' cc w/ enclosures:
W. Houston G. Lainas J. M11hoen J. Calvo R. Bosnak V. Rooney Contact M. Lamastra (X28507)
J. Ridgely (X29443)
DISTRIBUTION CENTRAL FILE PSB R/F ,
MLamastra KCampe ..
~~See Previous Concurrence -~~
5520 Document Name: VERMONT YANKEE REVIEW / REQUEST
)FC :PSB: DBL * :FSB: DBL * : DBL : : :
iAME :MLamastra/c . 1P.idgely Campe Hulman : : : : :
MTE :9/23/86 :9/23/86 :9/9/86 :9/'ky86 : : :
y -
0FFICIAL RECORD COPY
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PLANT SYSTEMS BRANCH REQUEST FOR ADDITIONAL INFORMATION REGAPDING THE VERMONT YANKEE SPENT FUEL POOL RERACK MODIFICATION.
1. Describe, in some detail, the procedure that will be used for (1) removal of the fuel assemblies from the present racks. (2) removal and disposal of the racks themselves, (i.e., crating intact, or cutting and drumming), (3) installation of the new high density racks and (4) Ioading the new racks with the presently stored spent fuel assemblies. The description should include, step by step, the number of people involved in each step of the procedure, including divers if they are found to be necessary, the dose rate they will be exposed to, the times spent in the radiation fields, and the estimated man-rem associated with each step of the operation.
~~2. Demo.. strate that the method used for removal and disposal of the old racks will result in ALARA exposures.~~
~~3. Discuss the build-up of crud along the sides of the pool and the removal methods that will be used to reduce radiation levels at the edge of the pool to ALARA levels.~~
l. Identify the raciation monitoring systems that will be used, and, indicate their locations in the spent fuel pool area. The monitoring systems in question are those which are to provide a warning to the personnel when-ever the area radiation levels ir. crease inadvertently to preset alarm trigoer levels.
5. Specify the present dese rate in occupied areas outside the spent fuel pool concrete shield wall. Provide an estimate of the potential increase of this dose rate if the space between the spent fuel and inside concrete shield wall is reduced due to the proposed modification.
~~6. Provide a sumary estimate of the projected changes in environmental doses ,~~
resulting from the spent fuel pool modification. The estimates should j include annual, as well as, total plant life doses. This also should 1 include an estimate of the potential amount of K-85 and H-3 released due {
to additional fuel being stored in the spent fuel pool.
~~7. Describe the estimated changes in the spent fuel pool filter dose rates, if any.~~
~~8. Indicate the depth of spent fuel pool water which nonnally will lie over stored-in-place fuel elements, and provide the resulting pool surface dose rates for this condition.~~
9. Relative to the proposed rerack modification, describe how an ALARA design reviest was conducted and documented. Indicate how the guidelines of Regulatory Guide 8.8 were met. The description should include available examples of how the experience of other facilities was utilized to help assure that post-modification operation and maintenance doses would be ALARA. Include a general description of the Vermon-Yankee ALARA program
- that will be applied during spent. fuel pool operation.
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10. Provide the number of fuel bundles and cycle length (annual or 18 month) for each reload from the first operating cycle until the spent fuel pool -
is fu11 (2,870 fuel bundles). ]
1 j
11. The submittal identifies compliance with the guidelines of Branch Techni- i cal Position ASB 9-2, but does not identify compliance with Standard l
~~Review Plan Section 9.1.3. Verify compliance with the guidelines of Standard Review Plan Section 9.1.3 and in particular with the specified uncertainty factor.~~
12. In the spent fuel cooling analysis, the submittal specifies the assumed operating time for the fuel is 16,000 hours0 days 0 hours 0 weeks 0 months . This is not correct. For the annual refueling case the power operating time should be 35,040 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months and for the 18 month fuel cycle should be 39.420 hours0.00486 days 0.117 hours 6.944444e-4 weeks 1.5981e-4 months . Provide the results of a revised analysis using the correct power operating times. 1 (Note: a capacity factor maybe applied if adequate justification for the I selected value is provided.)
13. Provide a discussion of the capability of the service water system to j remove the increased heat load associated with the increased storage j capability vithout raising the RBCCWS water temperature above 85*F for the worst heat load conditions.
~~14. Provide the following information for each spent fuel pool cooling system heat exchanger.~~
~~c1 Tubesurfacearea(squarefeet) b) Overall conductance (BTU /sq-ft-Hr 'f) c) Design flow rates for~~
~~1) Spent fuel pool water (Lbs/ hour)~~
~~2) RBCCWS water (Lbs/ hour) i j~~
~~15. Provide a discussion of when and how the numbers provided in response to '~~
~~question 14 were verified by actual performance testing.~~
~~16. Provide the design details of the spent fuel pool cooling system heat exchangers.~~
17. Provide a discussion as to the design features of the spent fuel pool cooling system related to meeting the single failure criterion with the spent fuel storage fecility filled with normal refueling and maintaining the pool vater temperature at less than 140*F. 3 18.. Provide the prcposed changes to the Technical Specifications. Based on the submittal indicating that the spent fuel pool cooling system is under sized, provide a commitment in the proposed Technical Specification modification to require the facility to be in cold shutdown conditions prior to aligning a loop of RHR for cooling the spent fuel pool.
x.
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~~19. Provide the spent fuel cooling capability model which was used to verify adequate natural circulation along with the detailed assumptions and ..~~
inputs. For each input / assumption which is not identical to the General Electric Company's fuel analysis for licensing the fuel, provide the General Electric Company's input / assumption and a discussion as to why the input / assumption used is conservative.
~~20. Provide design drawinns of the rack modules.~~
~~21. Provide the fuel loading pattern in the spent fuel pool showing each reload with each bundle's equivalent full power operating time in hours.~~
22. Provide the results cf a vector velocity analysis which demonstrates that each bundle will receive adequate flow for the fuel loading pattern specified in response to question 21 with respect to the spent fuel pool water return lines. - Alternatively, show by analysis that each bundle will receive adequate flow independently of the fuel loading pattern.
~~23. Provide drawings showing the proposed exact routing and water distribution system within the spent fuel storage facility.~~
24 Provide a discussion as to why ignoring the downcomer effects in.the
' . ant fuel storage locations is conservative. If the fuel is stored in a
~~sck and white pattern, icnoring the vacant storage locations may not be conservative.~~
~~25. Verify that the reactor building bridge crane, the lifting and handling rigs, and all special handling teols are single failure proof.~~
~~26. Verify that the attachment to the new and existing racks are single failure proof.~~
~~27. For every item which has not been identified as single failure proof, provide the following information, a) Detailed drawings of the component and the method and location of attachment (s),~~
b) The design and the actual stress. factors to be applied during re-racking operations.
c) Adiscussionastothepurpose(s)anduse(s)ofthecomponent.
d) A discussion of the component testing and inspection and frequency of each."
e) A discussion of the results and effects of the failure of the component at the most adverse time and the protective actions which will be taken to prevent fuel damage or damage to safety-related structures, systems and components.
sw v .. 6 .. ~. ~ ,. m m a s ; w m m .-- _ w ei ~ u: a. _ _ , _ _ _
l 4
f) Provide a discussion of the proposed methods to verify that no' damage has resulted to the pool liner, the fuel storage racks, or other safety-related structures, systems, or components as the result of i the failure of this component.
. 28. Provide drawints which show:
a) The arrangement of the spent fuel in the pool when each existing rack l is being' removed and each new rack is being installed.
-b) The sequence of each rack removal and installation of each new rack, c) The path'which each rack will follow within the pool, within the reactor building, and within any other area until the rack is outside of all safety-related structures.
d) The maximur 'neight that each rack will be lifted above the surface below.
e) The relationship of each load path to other safety-related systems and components and to embedded structural beams within each surface over which the rack will pass.
29. For'the spent fuel pool, and any other body of water over which the racks may pass, provide the results of an analysis of dropping the rack at its maximum carrying height. The results should address the effects of splashing water out of the pool, impact effects en the pool liner at its weakest point (at the leakage detection traces),
~~and impact on other fuel storage racks.~~
~~30. Verify that if.the rack were to fall on its side that no fuc1 would be impacted.~~
~~31. Verify that all operator training, load handling procedures, and eye testing of operators will be in conformance with NUREG-0612, "' Control of -~~
Heavy >.oads at Nuclear Pover Plants."
4 4
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1 .
, , JAN 0 61987
, d Docket flo. 50 '271 DISTRIBUTION Docket File VRooney PPL/TOSB y NRC/ Local PDR OGC-Bethesda Rbernere MEMORANDUMIq0R: Daniel R. Muller, Director PD#2 Memo File EJordan ACRS(IO)
BWR Project Directorate #2 ORAS BGrimes ,0PA..
Division.of BWR Licensing Denton/Vollmer Glainas t'RC Participants FROM: Vernon L. Rooney, Project Manager BUR Project Directorate #2 Division of BWR Licensing
~~SUBJECT:~~
FORTHCOMING HEETING WITH VERMONT YANKEE NUCLEAR POWER CORPORATION DATf. t TIME: Thursday, January 15, 1987 1:00 p.m.
LOCATION: HanfordHouse(askreceptionistforroomlocation)
Richland, Vashington PURPOSE: fleeting with Vermont Yankee Nuclear Power Corporation (VYNPC) representatives to discuss thermal-hydraulic considerations related to proposed spent fuel pool expansion.
PARTICIPANTS *: NRC Licensee VRooney PBergeson
('JRidgely PNL(consultant) i Original algaad by Vernon L. Rooney, Project Manager BWR Project Directorate #2 Division of BWR Licensing cc: See next page
Meetings between NRC technical staff and applicants for licenses are open for interested members of the public, petitioners, interveners, or other parties to attend as abservers pursuant to Open Meeting Statement of NRC Staff ,
Policy," 43 Federal Register 28058,6/28/78.
' DBL:PD# DBL:PD#2 DBL:Spf2 S'rs VRooney:ptf DG4Ter 1/6/87 1/6./ p D C,/87 r
j
L
'E .. , .. .. .i.2; TJ,n . . .;a. . . .. . . :. ..
. a ::(nrv i
o ,' '
Mr.'R. W. Capstick- Vermont Yankee Nuclear Power
~~p. Vermont Yankee Nuclear Power Corporation Station r ,~~
cc:
Mr. J. G. - (eigand .
Mr. W. P. Murphy.. Vice President & .
President 8 Chief Executive Officer Manager of Operations Vermont Yankee Nuclear Power Corp. Vermont Yankee Nuclear Power Corp.i R. D. 5, Box:169 R. D. 5. Box 169 ,
Ferry Road . Ferry Road ~
Brcttleboro, Vermont 05301 ~ Brattleboro, Vermont 05301 ;
Mr. Donald Hunter, Vice President' . Gerald Tarrant, Consnissioner .
Vermont Yankee Nuclear Power Corp. Vermont Department of Public Service 1671 Worcester Road 120 State Street L framingham, Massachusetts 01701 Montpelier, Vermont 05602 New England Coalition on. Public Service Board Nuclear Pollution State of Vermont Hill'and Dele Fam 120 State Street R.'D. 2, Box 223 Montpelier, Vermont 05602 Putney, Vermont 05346 Vermont-Yankee Decommissioning Mr. Walter. Zaluzny Alliance Chaiman, Board of Selectman Box 53 Post Office Box 116 . Montpelier, Vermont 05602-0053 Vernon, Vermont 05345 Resident Inspector Mr. J. P. Pelletier, Plant Managt: U. S. Nuclear Regulatory Commission
- Vermont Yankee Nuclear Power Corp. Post Office Box 176 Post Office Box 157 Vernon, Vermont 05354 Vernon, Vermont 05354 (
Vermont Public Interest Mr. Raymond N. McCandless. Research Group, Inc.
Vermont Division of Occupational 43 State Street j
& Radiological Health Montpelier, Vermont ')5602 '
Administration Building
- 10 Baldwin Street Regional Administrator, Region I Montpelier, Vermont 05602 U. S. Nuclear. Regulatory Commission
.. 631 Park Avenue Honorable John'J. Easton King of Prussia, Pennsylvania 19406 Attorney General State of Vemont Hamon & Weiss 109 State Street 2001 S. Street, N. W.
Montpelier, Vermont 05602 Suite 430 Washington, D.C. 20009-1125 John A. Ritscher, Esquire Ropes & Gray 225 Franklin Street Boston, Massachusetts 02110 i
.A g . P dD" 7 [ ' *
. VERMONT YANKEE PROPOSED SFP RERACK DECAY HEAT LOADS & POOL TEMPERATURES
~~Background~~
~~At time of licensing Vermont Yankee, spent fuel reprocessing was anticipated~~
~~Anticipated heat-loads would be from 1.25 to 1.5 cores~~
~~The proposed rerack is the second rerack for Verront Yankee~~
~~. Basis~~
~~Decay heat loads are calculated using the guidance of Section 9.1.3 and the Branch Technical Position ASB 9.5-2 of the Standard Review Plan (NUREG-0800)~~
~~Assumes the SFP is filled with fuel from normal refuelings~~

~~Uses the actual past reload data and the licensee's projected refueling data.~~
Applies the single failure criterion.
" Consider only the spent fuel pool bulk temperature at the time when the reactor will be isolated from the spent fuel pool, as opposed to the heat load at 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months after the initial reactor shutdown going into the refueling outage C oncerns
~~Smallest SPF cooling system (pump flow and heat exchanger capacity) that we have seen todate~~
~~Licensee has requested continuance of the current 150*F Technical Specification limit for the SFP water - this is the design limit for the cooling system~~
~~Degradation of SFP Cleanup System when water temperature exceeds 140'F i~~
~~Potential Technical Solutions~~
~~Addition of third SFP Cooling train~~
~~Increase the pump flow rates and heat exchanger capacities for both trains Current Position~~
~~For each refueling, remain in cold shutdown for a minimum of 68 days (149 days if the Technical Specification temperature limit is 140*F)~~
)
I
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ML20236P098

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