ML20207S780

Safety Evaluation Granting Relief from Tech Spec 4.7.A.3 on one-time Basis to Perform RHR Pump Wear Ring Replacement

ML20207S780
Person / Time
Site:	Vermont Yankee
Issue date:	03/12/1987
From:	Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20207S769	List: ML20207S768 ML20207S780
References
NUDOCS 8703200265
Download: ML20207S780 (6)

Text

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SAFETY EVALUATION BY THE REGION I 0FFICE REGARDING RELIEF FROM TECHNICAL SPECIFICATION 4.7.A.3 VERMONT YANKEE NUCLEAR POWER CORP 0r.ATION VERMONT YANKEE NUCLEAR POWER M ATION DOCKET NO. 50-271

1.0 INTRODUCTION

By letter dated March 3, 1987, Vermont Yankee Nuclear Power Corporation (VYNPC or the licensee) requested relief from the requirements of the Vermont Yankee Nuclear Power Station (Vermont Yankee) Technical Specification 4.7. A.3, in order to perform residual heat removal (RHR) pump wear ring replacement.

Subsequently, during discussions between the Vermont Yankee plant manager and the NRC senior resident inspector, as described in NRC Region I Inspection Report 50-271/87-06, additional conditions were specified to assure that ade-quate system integrity and leakage detection provisions are provided. Licen-see acceptance of these conditions (see Attachment 1) was confirmed on March 12, 1987 during a telephone conversation between Mr. J. Weigand, President and Chief Executive Officer, VYNPC and Mr. W. Kane, Director, Division of Reactor Projects, NRC Region I. The NRC Region I review and assessment of this matter is described below.

2.0 EVALUATION The licensee has proposed using alternate means instead of Type C leak rate testing to demonstrate the leak tightness of the following containment pene-trations: (1) X224A and B, which provides the torus suction for the four RHR pumps and contains one normally open isolation valve at the suction of each pump, V10-13A, B, C, and D; and (2) X12, which is the shutdown cooling letdown line from the A recirculation loop and contains two normally closed valves, V10-17 and V10-18. Valves V10-13A, B, C, and D are Walworth 20" dual wedge gate valves. Valve V10-17 is a Walworth 20" pressure seal wedge gate, and V10-18 is a Darling 20" double disc gate valve. Valves V10-15A, B, C, and D, downstream of the V10-17 and V10-18 valves, will be used as a pump main-tenance boundary, and are 20" Walworth dual wedge gate valves.

The alternate methods to verify acceptable leakage are discussed below along with the technical evaluation. The post accident containment pressure re-sponse, from FSAR Figure 14.6-5, is as follows: peak pressure of 44 psig within 1 second of the start of the accident; reduction to 27 psig at 10 seconds; reduction to 8 psig in 15 minutes; and, reduction to 5 psig in 1 day.

The post-LOCA source term discussed in FSAR Section 14.9 and in the bases for Technical Specification 4.7.A.3 is as follows: for a TID-14844 fission product release fraction and an assumed 1.5 wt % per day containment leak rate, the resultant site boundary doses are 1.65 rem whole body and 280 rem thyroid.

8703200265 870312 1 PDR ADOCK 0500

Safety Evaluation 2 Penetration X12 To assure an intact containment barrier for this penetration, the licensee states that the V10-15 valve will be administratively controlled closed with its motor supply breaker open when inspections are performed on the associated RHR pump. The licensee considers the closed valve equivalent to a blind flange. In addition to closing the valve, the licensee shall also lock it closed with a chain and padlock and subsequently attempt to open the valve using the associated CRP 9-3 control panel switch, to assure that the proper supply breaker is open. These controls will provide adequate assurance that a passive mechanical barrier is intact for containment isolation purposes during accident conditions.

Tail pipe leakage and pressure monitoring downstream of the V10-18 and V10-17 valves can provide verification that there is not excessive leakage from the primary system through this penetration. The licensee shall back up the ex-isting PS-118 pressure switch with an on-line local pressure indicator that is surveiled once every four hours to provide early detection of any leakage.

The licensee shall keep the V10-17 and V10-18 valves closed, which will pro-vide three valves as a barrier between the primary system and an opening in any RHR pump. The penetration can be verified to be leak tight by the above methods under present conditions with 1000 psig on the primary side. Under postulated post accident conditions, the piping upstream of V10-18 could lose its water seal from the reactor vessel, but the pressure on the valve seat under those conditions will be limited to 44 psig and water between V10-17 and V10-15 should maintain a water seal on the V10-15 valve seat.

The administrative controls on the V10-17, V10-18, and V10-15A, B, C, and D valves and the use of the pressure sensing instrumentation downstream of V10-17 provide an adequate test alternative to demonstrate the acceptability of leakage through penetration X12 for the time it will take to replace the RHR pump wear rings.

Penetration X224A and B To assure an intact containment barrier for this penetration, the V10-13A, B, C, and D valves (one for each pump) will be administrative 1y controlled closed with the motor supply breaker open when inspections are performed on the asso::tated RHR pump. The licensee shall also lock the respective V10-13A, B, C, or D valve with a chain and padlock and subsequently attempt to open the valve using the associated CRP 9-3 control panel switch, to assure that the proper supply breaker is open. These controls will provide adequate as-surance that the torus will not be inadvertertly drained through the open RHR pump during normal operations, and will furth3r assure that a passive mechani-cal barrier is intact for containment isolation purposes during accident con-ditions.

Safety Evaluation 3 The licensee stated that primary containment integrity is maintained for pene-tratio- X224A and B by the water seal on valves V10-13A, B, C, and D assured by the torus. The minimum torus water level allowed by the technical speci-fications (68,000 cu. ft.) provides at least 9 feet of submergence.for the torus suction line for anticipated normal, transient, and accident conditions.

The minimum torus level can be maintained from several systems (core spray, HPCI, RCIC) in the post-accident configuration, so that no single active failure can prevent maintenance of the water seal for at least 30 days. Use of the core spray system to replenish the torus assumes entry into the RHR corner room to open a manual suction valve from the condensate storage tank.

The combination of the above controls on the V10-13A, B, C, and D valves, the water seal on the these valve seats maintained by the torus, and control of any leakage out of the penetration (discussed below) provides an adequate test alternative to demonstrate the acceptability of leakage through penetrations X244A and B.

Leakage Potential The licensee estimated that the amount of acceptable leakage through the V10-13 and V10-15 barriers will be small in order to complete the necessary work inside the pump casing. The actual amount of leakage past the valves cannot be estimated until the pump is drained in preparation for the inspection, or further quantified until the pump boundary is open.

The licensee stated thh~, their evaluation concluded for design basis LOCA con-ditions that the coolant leakage past the V10-13 and V10-15 boundaries and potential airborne activity would be minimized. Also, the boundary leakage would be controlled. However, the licensee failed to quantify the leakage source term for staff review and comparison with the acceptable release source terms established in the accident analyses.

While the V10-13 water seal will provide a barrier against the post accident gaseous source term, operation of the operable RHR train in the loop under repair will put water with the post accident particulate source term at the V10-13 valve seat. Potential leakage past the valve seat could result in the release of dissolved gases (from torus water at 170 degrees F) and the subse-quent airborne activity from resuspension of particulate activity following evaporation of any water leakage not contained. The leakage from the pene-tration for the open RHR pump shall be collected and contained by the licensee.

As shown in FSAR Drawing 4.8-2, there is a 3/4 inch drain line from the bottom of the pump bowl that is hard piped to the floor drain sump located in each RHR corner room. The floor drain sump is in turn pumped to radwaste via the reactor building floor drain sump pump (FSAR Drawings 9.2-3 and 9.2-4) and is held for processing in the 25,000 gallon floor drain collector tank in the radwaste building. The post-accident source term available from this poten-tial leakage source shall be assumed to leak to the reactor building atmos-phere and subsequently be released to the environs following filtration from the SBGTS.

Safety Evaluation 4 The licensee informed the NRC orally on March 9, 1987, that in order to minim-ize the potential for containment boundary leakage and any resultant environ-mental release, the RHR pump disassembly would not be performed if there is any measurable leakage flow from the containment boundary. Further work on any pump not inspected as a result of potential containment boundary leakage will be deferred until the next refueling outage, currently scheduled for August 1987.

Some uncertainty exists relative to the additional rost-accident source term available for release to the environs from the essentially zero leakage from the established valve boundaries. The incremental increase (if any) is judged to be small and shall be minimized by the licensee by the provisions that there will be no measurable leakage flow. Acceptance of the source term un-certainty is justified by the low probability of an accident in the eight days (four pumps X two days / pump) of plant operation with the RHR pumps disassembled.

Based on the above, the licensee shall establish a hold point in the pump inspection procedure that will, after pump drain down and prior to RHR pump disassembly, verify that the leakage is essentially zero (i.e., no steady stream) to assure the barrier is intact. This measure will assure that the post-accident radioactive source term emanating from any leakage will result in an acceptably small incremental increase to the post accident source term released to the environs described in the bases for Technical Specification 4.7.A.3.

Contingency Plans The licensee estimates that it will take about five days to perform the in-spections on any one pump, and during that period, the pump boundary would be open for about two days. During this period, surveillance shall be per-formed of the open boundary using operator tours once every four hours. If any leakage is detected while the RHR inspection is in progress, actions shall be taken immediately to close the opening in the RHR pumps as soon as possible.

The licensee stated that the pump could be closed within about three hours by installing the old pump internals if necessary, along with the stuffing box and shaft seal package. The three hours includes calling in workers on a back shift since the licensee intends to run the job only on ten-hour day shifts.

The preferred course of action, conditions permitting, is to restore the RHR loop to its full operability. Therefore, should a need arise to implement this contingency plan, at the discretion of the shift supervisor, the RHR pump reassembly shall be performed on a round-the-clock basis, restoring the pump to an operable status with a new impeller. This action could take up to three days to complete.

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Safety Evaluation 5

3.0 CONCLUSION

The staff has reviewed the licensee's submittal dated March 3: 1987 and per-formed its own independent evaluations. The staff has determined that, sub- I ject to the conditions described in Attachment 1, the licensee can safely perform the RHR pump inspections.

Based on the above considerations, and the fact that the planned inspection and maintenance of residual heat removal (RHR) pumps should enhance overall system safety and reliability, the staff concludes that the licensee's meas-ures to provide alternative leakage detection and piping isolation of the RHR system are adequate to meet the intent of Technical Specification 4.7.A.3.

4.0 Relief from Technical Specification 4.7.A.3 Based on the authority of the Regional Administrator to exercise enforcement discretion regarding Technical Specifications, as delegated by the Director, Office of Nuclear Reactor Regulation and the Director, Office of Inspection and Enforcement, in a memorandum dated February 27, 1987, I have determined that the above circumstances, in which a license amendment would not be ap-propriate, are suitable for the use of such enforcement discretion. Further-more, the granting of specific relief from Technical Specification 4.7.A.3, for the limited purposes described above and subject to the conditions of Attachment 1, will not endanger life, property, or the common defense and security of the public.

Therefore, the requested relief from Technical Specification 4.7.A.3, as described in the VYNPC letter dated March 3, 1987, and subject to the condi-tions in Attachment 1, is hereby granted.

' Original ~ signed lif, Thru r. Murley, -

Thomas E. Murley Regional Administrator Dated: 12 MAR 1987

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ATTACHMENT 1 The licensee shall:

1. Administratively control the associated V-10-13A, B, C, and D and the V-10-15A, B, C,.and D valves as follows: locked closed with chain and lock, along with opening their MOV supply breakers, and subsequently test that the proper sup-ply breakers are.open by attempting to open each valve using the CRP 9-3 con-trol panel switch for each valve, one at a time.

2. Provide an on-line local pressure indicator in addition to the PS-118 pressure switch both of which shall be surveiled once every four hours to provide early indication of any leakage through the X12 penetration.

3. Establish a hold point in the pump inspection procedure that will verify, after pump drain down and prior to RHR pump disassembly, that there is no measurable leakage from penetrations X12, X224A, and X224B, i.e., no steady stream.

4. Maintain surveillance over the open boundary using operator tours once every four hours. If measurable leakage from the containment is detected, i.e.,

a steady stream, the shift supervisor shall order the pump to be reassembled expeditiously. Should a need arise to implement this action, the RHR pump reassembly shall be performed on a round-the-clock basis, with the intent to restore the pump to an operable status with a new impeller.

5. Develo; and implement a contingency plan (using the pump reassembly procedure) that wil! provide for rapid closure of the RHR piping boundary (e.g., in about three hcurs) in response to a decision by the operations shift supervisor that such action is necessary, if leakage conditions do not permit reassembly using the new impeller, i

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