ML20100M839

Responds to IE Bulletin 84-03 Re Refueling Cavity Water Seal.Probability of Seal Failure Significantly Lower than at Haddam Neck & Gross Failure Highly Unlikely to Occur

ML20100M839
Person / Time
Site:	Salem
Issue date:	11/21/1984
From:	Liden E Public Service Enterprise Group
To:	Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
IEB-84-03, IEB-84-3, NUDOCS 8412120473
Download: ML20100M839 (16)

Text

E d PSIEG Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department November 21, 1984 Dr. Thomas E. Murley, Administrator U.

S.

Nuclear Regulatory Commission Region 1 631 Park Avenue King of Prussia, Pa.

19406

Dear Dr. Murley:

IE BULLETIN 84-03 REFUELING CAVITY WATER SEAL SALEM GENERATING STATION UNITS NO. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 The purposes of IE Bulletin 84-03 are to:

(1) notify addressees of an incident in which the refueling cavity water seal failed and rapidly drained the refueling cavity, and (2) request certain actions to assure that fuel uncovery during refueling remains an unlikely event.

A summary of the applicable section and PSE&G's method of compliance is presented below:

ITEM 2 Evaluate the potential.for and consequences of a refueling cavity water seal failure and provide a summary report of these actions.

Such evaluations should include consideration of:

gross seal failure; maximum leak rate due to f ailure of active components such as inflated seals; makeup capacity; time to cladding damage without operator action; potential effect on stored fuel and fuel in transfer; and emergency operating procedures.

8412120473 e41121 PDR ADOCK 05000272 kl PDR G

The Energy People h 2 t t>8180M; I182

Dr.-Thomas E. Murley 11/21/84

RESPONSE

The attached evaluation addresses the potential for and consequences of a refueling cavity water seal failure as requested by IE Bulletin 84-03.

This evaluation concludes that there are a number of substantial differences between the inflatable seal design used at Connecticut Yankee's Haddam Neck and that used at the Salem Nuclear Generating Station.

The differences include. seal surface conditions, size of gap to

~

seal' dimensions, seal material, seating procedures and

. placement of brackets on top of the seal.

As a result of the numerous differences, it is concluded that the probability of seal failure at the Salem Station is significantly lower than at Haddam Neck and a gross seal failure is considered highly unlikely to occur.

Although precautions have been taken to assure the reliability of.the refueling cavity water seal at the Salem Generating Station, the consequences of a seal failure have been evaluated.- There are adequate means of detecting a seal failure and subsequently preventing fuel failure through existing signals, procedures and Technical Specifications.

Implementation of recommended emergency procedures will further increase the safety margin at the Salem Station.

Should you have any questions, please contact us.

Sincerely,

• A E..A.-Liden Manager - Nuclear Licensing and Regulation Attachment C

Mr. Donald C.

Fischer Licensing Project Manager Mr. James Linville Senior Resident Inspector U.S. Nuclear Regulatory Commission Document Control Desk Washington, D. C.

20555

D.

t STATE OF NEW JERSEY )

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ss:

COUNTY OP SALEM-COUNTY OF SALEM

)

RICHARD A. UDERITZ, being duly sworn according to law deposes ~and says:

I am a Vice President of Public Service Electric and Gas Company, and as such, I find the matters se't forth in~our response dated November 21, 1984, to IE BULLETIN 84-03

" Refueling Cavity Water Seal", are true to the best of my knowledge, information and belief.

RICHARD n. UDERIT.P Subscribed and sworn to before me this 2'ha day of Novenes R

_, 1984 lY

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fotarf Public of Nfw Jersey w~

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.v My Commission expires on to ca.x. w mi a m.; et w. ma k

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%d Date 11/13/84 Public Service Electnc and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department S-C-N300-MSE-285 TITLE: IE BULLETIN NO. 84-03:

REFUELING - CAVITY WATER SEAL

1.0 PURPOSE

The purpose of this Safety Evaluation is to evaluate the potential for and consequences of a refueling cavity water seal failure as requested by IE Bulletin No. 84-03.

2.0 SCOPE

This Safety Evaluation and its conclusions are applicable to both Units of the Salem Nuclear Generating Station i

during a refueling outage.

3.0 REFERENCES

3.1 IE Bulletin No. 84-03:

" Refueling Water Cavity Seal", August 24, 1984.

3.2 Operating Plant Experiences 8-27 OElll7 " Connecticut Yankee Leakage Past the Reactor Cavity Pool Seal".

3.3 Telecon From C.

R.

Gerstberger to G.

Dillion August 24, 1984 " Connecticut Yankee Sealing Ring Incident".

3.4 PSE&G Design Calculation, S-C-N300-MDC-079 " Effects of a Gross Seal Failure of Refueling Cavity Water Seal".

3.5 PSE&G Safety Evaluation SGS/M-SE-037,

• Inflatable Reactor Cavity Refueling Seal Restraints".

3.6 Sandia Laboratories Report:

" Spent Fuel Heatup Following Loss of Water During Storage", March 1979.

3.7 Maintenance Procedure, M8H, " Reactor Cavity Inflatable Seal Installation and Handling".

3.8 Maintenance Procedure, M8C, " Reactor Vessel Head and Internals Removal and Installation".

3.9 Operating Instructions II-8.3.8, " Emergency Filling of the Spent Fuel Pool from the RWST".

EDD-7 FORM 1 REV 0 10 SEPT 81 m :w:w w

1 e

Pago 2 of 10 B-C-N300-MSE-285 s

3.10 Operating Instructions II-8.3.1, " Filling the Spent Fuel Pit".

3.11 212358 A 8874, " Refueling Canal Inspection Plugs and_

Structural Concrete Forms".

3.12 205213 A 8760, " Demineralized Water Make Up".

3.'13 205229_A 8761, " Chemical and Volume Control Boric Acid Recovery".

3.14 205230 K 8761, " Chemical and Volume Control Primary Water Recovery".

3.15 20523'4 A 8760, " Safety Injection".

3.16 PSBP 112177, " Reactor Vessel Cavity Seal Assembly and Details".

3.17 PSBP 145161, " Fuel Assembly Outline and Reprocessing Drawing".

3.18 PSBP 148820, " Spent Fuel Module (9 x 10)".

3.19 Technical Specification 3.9.5, " Refueling Operations

- Communications",.

3.20 Technical Specification 3.9.8, " Refueling Operations

- Coolant Circulation".

3.21 PSE&G Alarm Book.

3.22 Letter to Mr. Theodore Hollander, Jr. from R. T.

Stanley dated November 6, 1984 entitled " Refueling Cavity Water Seal."

4.0 BACKGROUND

On August 21, 1984, the Connecticut-Yankee Haddem Neck plant experienced a f ailure of the refueling cavity water seal with the refueling cavity flooded.

The seal assembly consisted of an annular plate seal ring (approximately two feet across) with two Presray inflatable seals to fill two inch openings on either side of the seal ring (See Figure 1).

The outer seal was subject to a gross seal failure which allowed 1/4 of the seal to f all through the annulus.

- Contributing factors to the failure were the inflation pressure, use of lubricants, and the size and configuration of the gap to seal dimensions.

These conditions resulted in bowing of the top of the seal which allowed it to be pulled through the annulus.

EDD-7 FORM 1 REV 0 10 SEPT 81

c Page 3 of 10 S-C-N300-MSE-285 Date:

11/13/84 The seal. failure caused. the refueling water cavity to drain

.its entire vo'lume, approximately 200,000 gallons, in 22 Lminutes. ~No fuelchad been in transfer at;the time of the failure.-

If fuel had been in transfer, it could have been partially or completely uncovered.with possible high radiation-levels.

If the fuel was exposed for a s

significant amount of-time and allowed to increase in temperature, Lthe possibility of fuel cladding failure and release of radioactivity may exist.

Furthermore, if the.

fuel transfer tube had been open, the spent fuel pool could have drained to a level which may lead to-the uncovering-of the top of the fuel.

5.0

-DISCUSSION:.

5.1 DIFFERENCES The refueling cavity water seal used by the Salem Nuclear Generating Station is only.slightly similar in design to that used at Haddem-Neck.

However, there are great differences in the dime,nsions, material, and utilization of the. seal.

-The annulus surrounding The reactor at.the Salem Station is much smaller than that at Haddem Neck, two inches ~as opposed to two feet four inches, therefore no seal ring is'necessary.

Only one. Presray inflatable refueling seal is used to form a secure closure between the reactor vessel seal ledge and the cavity wall.

Prior to the. initial installation of the seal at the Salem Station,. the cavity wall ledge was beveled to a 20' angle, the same angle as the wedge portion of the seal.

This' produced a dependable cavity wall seal surface by providing an area contact as opposed to the line contact seen at Haddem Neck.

If the seal becomes dislodged and begins to slip, the beveled area also provides an increase in frictional contact.

This increased frictional contact will aid in retaining the proper placement of the. seal.- Many additional p,recautions were taken at the Salem Station prior to the initial use of the Presray seal.

Any -irregular or iinterupted seal surfaces were reconditioned and backfilled.

All local annulus surface conditions of weld splatter, grout, rough or sharp metal edges were. removed.

The cavity wall side was machined to smooth and contour the ' s'urf ace.

The reactor vessel seal ledge side surface was hand deburred and cleaned.

All this was completed to provide a smooth surface finish necessary for inflatable - seal support, protection and seal surface development.

~

E DD-74: FORM 1 REV 0 10 SEPT 81

+

Page 4 of-10

. 11/13/84

' Dates

[S-C-N300-MSE-285 The; inflatable portion of the seal at the Salem Etation.is exposed to a greater amount of surface contact area from the annulus walls, 2 1/4 inches on

.one side and~ full length contact on the other, side (See Figure-2).

Connecticut Yankee has an equal i

amount of surfaceLcontact on each side, l~5/8; inches.

Thus, the s'eal at the Salem Station will balloon out only on the reactor flan e side while the w

seal at Connecticut Yankee will~ experience-this on

.both-sides. ~ At the Salem Station less-directional force will be. exerted ~on the seal that tends to pull the seal downwerd.

Therefore, the annulus designoat the Salem Station leads to an increase in the. margin of safety.

In addition to the dimensional differences in the annulus at Haddem' Neck and the Salem Station, the seals themselves differ in size.

The seals used at

' the Salem Station are 4~ inches wide across the top

. edge portion, as opposed to 3 1/2 inches at Haddem w

Neck. - Both of these seals are used to secure a'two 1

inch area.

Therefore, the seal size will aid in prohibiting-the seal at the Salem Station from pulling through the annulus.

.The material difference bet'een the seals also w

increases the margin of safety at the Salem Station.

The Salem seal is 60 durometer, while the seal used at Haddem Neck is 40 durometer.

This increase in i.

hardness will assist in,the prevention of the seal failure.

The hardness will impede the. seal from bowing and bending and therefore hinder it from being pulled through the two inch annulus opening.

~

Prior to each installation of the seal at Haddem Neck, a lubricant'such as' silicone grease is applied to the annulus.

This is done in conjuction' with the air tight test-that is. performed to test the seal for proper seating.

This lubricant will actually aid.in the failure of the seal by reducing the frictional-resistance the seal'would experience from the annulus wall.- At the Salem Station no lubricant is used,

.thus reducing the chances of seal failure.

' 'To further increase the safety margin at Salem Stations, brackets are placed on top of the Presray seal (Reference 3.5).

No such brackets are used at O

Haddem Neck.

Haddem Neck does utilize a seal

'supporte but this is employed only during the initial

~

=

placement of the seal.

It does not aid in retaining proper positioning or support the seal during use.

1 EDD-7' FORM.lLREV 0-10 SEPT 81 4

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Pcgo 5 of 10 Date 11/13/84

-S-C-N300-MSE-285 7

. The brackets at the Salem Station are a minimum of

. three inches in diameter, therefore fully covering' the L reactor. cavity annulus-of two inches.

The additional coverage of-the brackets will reinforce

-the seal capabilities.

The brackets are.used to

' assure-that the inflatable refueling seal will~not

. become. dislodged-from-the reactor cavity seal. ledge.

-The'use cf brackets also aid in the prevention of bowing of the top of the seal.

Possible ~ failures of the Presray. seal used at the Salem Nuclear. Generating Station'have previously been reviewed' in a Safety Evaluation.

The'results-provided necessary assurance 'that the seal will function as ' required without the possibility of

~

-dislodgement from the reactor cavity seal ledge (Reference 3.5).

'. Maintenance procedure ~ inspection hold points will further assure the inflatable ' seal is ' in proper-position.-

The procedure for.the-reactor cavity seal installation-(Reference 3.7) contains Supervisor /

4 Witness inspection hold points and twice confirms

~

proper placement of the seal.

The seal is: first inflated to a pressure of 10 psig and~ inspected for

-positioning.

Iflthe seating is acceptable, the l

pressure in the seal will then be increased to 30 l'

psig and again reviewed for effective seating.

The l'

reactor cavity water level is raised with a

- Supervisor / Witness. present and the validity of the sealing is verified with the. Control Room assuring that there~is no abnormal running of the Reactor Sump Pump.

These added precautions are taken to further.

assure the reliability of the reactor cavity' seal.

L b

Because of the many dif ferences in dimension,

- material and utilization, and the numerous additional levels of safety at che Salem Station, we forsee no L

reason why the use of. the Presray seal will result in L

a gross seal failure.

l

- 5.2 GENERAL INTRODUCTION Although a gross seal failure during a refueling operation is highly unlikely'to occur, the n

L consequences of this event have been evaluated.

The

. flowrate of the liquid through the annulus would vary according to the height of the liquid.

In the case L

of the Salem Nuclear Generating Station, i f the entire seal were to fail the maximum flow rate would L

be-104,000 GPM (Reference 3.4).

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Pago 6 of 10 S-C-N300-MSE-285

'Date:

11/13/84 The time to drain the volume of liquid to the; level of the seal would depend on a number of items including the percentage of the seal which fails, if the fuel transfer tube were open to the Transfer Pool, and if the canal gates were open to the' Spent Fuel Storage Pool.

Assuming the entire seal failed, it would take 4 minutes, 47 seconds to drain the Refueling Area alone (248,000 gallons).

In order to drain ~ both the Transfer Pool and the Refueling Area (332,000 gallons total), the fuel transfer tube must be open and 6 minutes, 23 seconds must pass.

If both the fuel transfer tube and the canal gates were open,.

the time to drain the Spent Fuel Storage Pool, Transfer Pool and the Refueling Area (525,000 gitllons total) is slightly over ten minutes (Reference 3.4).

5.3 FUEL IN TRANSFER

-The worst case possible resulting from this failure situation ~for fuel in transfer would come about if four fuel assemblies were between the Reactor and the Transfer Pool:

two in the Rod Cluster Control carriage compartment, the third in the upender, and-1 the fourth fuel assembly in the manipulator crane.

If.an assembly were in the upender, it must -be layed down to prevent exposure.

Any fuel assembly that may be in transfer at the time of the seal f ailure must-either be' returned to the reactor or placed in the upender, if available, and set down.

If the assembly

+

were half-way through the transfer process,'it would take less than five minutes to move the assembly to g

either safe position.

The top of the assembly in the Rod Cluster Control carriage compartment would become exposed to the atmosphere.

l t

j With no operator action cladding damage may occur to the fuel assemblies in the manipulator crane and in the Rod Cluster Control carriage compartment.

An extremely conservative estimate for time to cladding damage would be two hours (Reference 3.6).

This

~

- estimate is based on an analysis done for a full core unloading in an emptied spent fuel pool.

As.a result

.of the differences in number of fuel assemblies involved, 'a maximum' of 3 in actuality as opposed to

'193 in the analysis, and the distance between assemblies, the two hour estimate is a worst case situation.

The actual time to possible cladding g

rupture would be increased.

Cladding damage to the

[

other assembly in the upender would not occur until l

24 days af ter initial drainage to the seal because of the large volume of-water surrounding it.

EDD-7 FORM 1 REV 0 10 SEPT 81 I

Pcgo 7 of 10 S-C-N300-MSE-285.

Date:

11/13/84 5.4 FUEL IN REACTOR

. If'the water 11n the Refueling Area-has drained to the

]:

level of the refueling seal, the water remaining in the reactor will begin to increase in temperature if there was no. circulation.

This will be relieved by

. the Residual Heat Removal System (RHRS) which is functioning.~during the refueling process accordin, to

, Technical Specifications (Reference 3.20).

The RHRS will remove the heat energy from the core and the Reactor Coolant System by recirculating a minimum of

- 3000 GPM through the system.

Therefore there is no possibility of cladding damage even if no operator action is taken because the RHR. System is functioning during any refueling procedure.

I The make-up capabilities to the reactor are supplied from two sources.

The first is the remaining water in the Refueling Water. Storage Tank.

This tank will contain over 100,000 gallons of water available for use.

An alternative source of make-up comes from the Reactor Sump.

Use of this - sump would recirculate the drained water-into the reactor, therefore achieving minimal ~ water losses due to the seal failure.

5.5 FUEL'IN SPENT FUEL POOL Once the liquid has drained to the level of the refueling cavity seal, another situation may arise.

The liquid in the Spent Fuel Pool will begin to increase in temperature and may begin to boil resulting in the possibility of ' exposing spent fuel.

The worst case considered is when a full core-load is removed from the reactor 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> after shutdown.

Although the. Technical Specifications allow for fuel removal after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, it is not expected.that any unloading will occur until at least 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> have-L passed.

This is a result of all other procedures which must take place prior to the actual unloading of fuel during a refueling outage.

It would take three hours twenty-three minutes for the water in the 3

(

Spent Fuel Pool to reach the boiling. temperature of 212

• F.

The water would then boil off at a rate of 52 GPM resulting in the water level to drop at a rate of 4 3/4 inches per hour. -Since the active portion l-of-the fuel assembly is only three inches below the L

level of the refueling seal,'the fuel wil1~be exposed L

approximately four hours after the initial drainage i'

occurs (Reference 3.4).

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11/13/84 S-C-N300-MSE-285 LIf no operator action was taken, the fuel rods would become exposed to the atmosphere.

If no credit for any cooling by. water or steam-is taken af ter the water. level drops to the active portion of the fuel (an extremely conservative assumption) there is a 3

possibility'of cladding f ailure two hours af ter the active fuel is first uncovered (Reference 3.6) or six hours after drainage to the seal level.

In actuality, it would take almost thirty hours to boil off the total volume of liquid.

The majority of heat generated from the fuel rods is produced in the central region,'which will remain covered with water for fifteen hours.

The boiling water in the Spent Fuel Pool can be replaced from the Demineralized Water System, Holdup Tanks, Primary Water Tanks and the Refueling Water Storage Tanks, as outlined in the Operating Instructions (Reference 3.9 and.3.10).

The Demineralized Water System contains two 500,000 gallon tanks with a pumping capability of 650 GPM to the Spent Fuel Pool.

There. are three 63,500 gallon hold-up tanks connected to a pump that supplies 500 GPM to the Spent Fuel Pool.

A third source of make-up water comes from the 250,000 gallon Primary Water Tank and pumps that. provide up to 200 GPM of water.

Any of these three sources can be made available within 30 minutes.

Water can also be taken from the 100,000 gallons remaining in the Refueling Water Storage Tank.

This can assure 100 GPM through the Refueling Water Purification Pump given a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> preparation period to properly align piping.

This total make-up water supply will more than replace any water lost from boil off in the Spent Fuel Pool.

5.6 RECOMMENDED EMERGENCY PROCEDURES p

L

(

Direct communications are maintained between control room and personnel at the refueling station as required by Technical Specification 3.9.5 during core

. alterations (Reference 3.19).

EXISTING SIGNALS IN THE CONTROL ROOM Reactor Sump Overflow ( ARP OHA D27) activated when running of sump pump occurs and i

Spent Fuel Pool Low Level (ARP OHA C28) activated if the water level in the pool is low (below 128 feet 2 inches) indicating loss of water in the spent fuel pool.

l l

EDD-7 FORM 1 REV 0 10 SEPT 81

n-

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^

.S-C-N300-MSE-285

'Date: 111/13/84

-RECOMMENDED. EMERGENCY ACTION Visually inspect. water level line in the 4-

, Refueling Water Cavity or the Spent Fuel Pool.

If the-level is:well;below the normal water level, begin emergency action.

'If'the water level-istcrogging, transfer the RHk

-pump suction lfrom.the. hot leg of the RHRS to the RWST..

If not, ensure the RHR System is running properly as'necessary during. refueling operations according to the Technical Specification 3.9.8 and-continue surveillance of water level, and Prepare for make-up to the Spent Fuel Pool

according to Operating. Instructions II

'8.3.8 and

' Begin make-up to. the Spent Fuel Pool to refill lost water and Move any fuel in the manipulator crane to the reactor.or upender, and Place any fuel. in the upender to the laying down position and Close the Fuel Transfer Tube isolation valve.

T6.0- l CONCLUSION /

SUMMARY

- There are a number of. substantial differences :between the refueling cavity. water seal design used at Connecticut

. Yankee's Haddem Neck and that used at the Salem Nuclear Generating Station.

At the Salem Station the cavity wall

-ledge is beveled to a 20' angle and has been machined and

' backfilled forming'a-smooth surface finish.

In addition, the; reactor vessel seal edge has been hand deburred:to

' produce a more effective seal suface. 'The inflatable seal used at the Salem Station is wider across'the wedge portion and is used to seal the same size area.

'The seal material is harder than that used in manufacturing the Haddem Neck seal and will aid in the prevention of bowing and bending.

Haddem Neck also utilizes a lubricant in seating the seal, which is not done at the Salem Station.

To' increase the safety margin at the Salem Station, brackets 'are placed on 1

top' of the Presray seal to further assure a secure: closure.

!p EDD-7 L FORM 1 RFV 0 10 SEPT 81 3

j._ _..

Page 10 of 10 Date:

11/13/84 S-C-N300-MSE-285 Maintenance procedures at the Salem Station further confirm proper placement and utilization of the seal..As a result.

of the numerous differences, the probability of seal

-failure at the Salem Station is considered significantly 1

' considered highly unlikely to. occur.

lower than at Haddem Neck and a gross seal failure is Although precautions have been taken to assure the reliability of the refueling cavity water seal at the Salem Nuclear Generating Station, the consequences of a seal failure have been evaluated.

There are adequate means of detecting a seal failure and subsequently preventing fuel failure through existing signals, procedures and Technical Specifications.

Implementation of recommended emergency procedures will further increase the safety margin at the.

Salem Station.

In addition, during extended periods of time where there are no core alterations or fuel transfer, the Fuel Transfer Tube isolation valve shall be closed.'

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