ML20207E514

Safety Evaluation Supporting Amend 88 to License DPR-61

ML20207E514
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	12/24/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207E479	List: ML20207E475 ML20207E504 ML20207E514
References
NUDOCS 8701020195
Download: ML20207E514 (13)

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[ g NUCLEAR REGULATORY COMMISSION 5 l WASHINGTON. D. C. 20$55

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 88 TO FACILITY OPERATING LICENSE NO. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET NO. 50-213

1.0 INTRODUCTION

By letter dated December 17, 1986, as supplemented by letter dated December 19,1986 the Connecticut Yankee Atomic Power Company (CYAPCO)submittedarequestforchangestotheHaddamNeckPlant technical specifications.

The amend:nent would establish a plant configuration which provides assurance that adequate cooling will be maintained during sump recir-culation while satisfying single failure requirements. That configuration I

involved repositioning and locking flow control valve RH-FCV-796 in the Residual Heat Removal (RHR) system in the partially open position and initiating, under prescribed conditions, charging system flow to proper flow distribution and pump operability in the event the specific small line break at issue were to occur.

The staff's evaluation of the licensee's request for an emergency license amendment is provided below.

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2.0 BACKGROUND

During engineering implementation of the proposed residual heat removal /high pressure safety injection (RHR/HPSI) cross-tie for high pressure pump 87o1020195 DR 861224 ADOCK 05000213 PDR

recirculation (Reference 1), the licensee, Connecticut Yankee Atomic Power Company (CYAPCO), for the Haddam Neck Plant identified an intermediate break LOCA in the core deluge line for which adequate core cooling would not be ob-tained during sump recirculation. CYAPC0 notified the NRC of this condition on December 12, 1986 in accordance with 10 CFR 50.72. Additionally, CYAPCO and the staff held a conference call that day and met on December 15. During the telepnone call and the meeting, CYAPC0 presented the results of their engineering review and proposed a resolution for this issue.

On December 17, 1986, the licensee formally submitted (Reference 2) their

, proposed resolution; supplementary information was provided via Reference 3.

To resolve this issue, the licensee has proposed, as an interim measure, an amendment to the Haddam Neck Plant Technical Specification 3.6, " Core Cooling Systems," for the remainder of the current fuel cycle, Cycle 14. The proposed Technical Specification revision consists of blocking flow control valve RH-FCV-796 in a throttled position in order to assure adequate injection flow during sump recirculation for all postulated LOCAs. The cur-rent Technical Specification requires this valve to be blocked in a fully open position. Additionally, the licensee's submittal discusses procedural o modifications made to assure adequate flow during the sump recirculation period of a LOCA.

3.0 EVALUATION

3.1 Nature of Problem The Haddam Neck core deluge system consists of two flow paths each containing a check valve adjacent to and upstream of a motor operated valve (MOV) normally maintained in a closed position. Downstream of the MOVs, each line branches (Tee-connection) into two lines which -

inject directly into the reactor vessel. The piping network for the -

core deluge system consists of six inch diameter (OD) piping upstream of the Tee-connection, M0V and check valves, and four inch diameter (0D) piping downstream of the Tec-connection.

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During CYAPCO's engineering of the RHR/HPSI cross-tie (Reference 1),

CYAPC0 examined the effect of a break in the core deluge system piping downstream of the HOVs. Upon receipt of a safety injection signal, the MOVs in the core deluge system would open. Thus, assuming a break in the core deluge system piping, a significant fraction, if not all, of the flow through the core deluge system from the low pressure safety injection (LPSI) would be lost directly through the break during the injection phase of the accident. The core deluge piping configuration, and the restrictive nature of the nozzles I

through which the core deluge enters the reactor vessel, restricts the break size to less than 0.08 ft2 For breaks of this size, system pressure would remain above the shut-off head for the LPSI pumps during the injection phase. Thus, only the charging pumps and/or the HPSI pumps would be available during the injection phase of the LOCA for the core deluge line break. Previous Haddam Neck small break LOCA evaluations, performed to demonstrate conformance with the Interim Acceptance Criteria for Emergency Core Cooling Systems, have demonstrated that these systems would provide adequate core cooling during the injection phase for j breaks of this size.

Upon ente' ring the sump recirculation phase of the LOCA (deterinined by depletion of at least 100,000 gallons of water from the Refueling

! Water Storage Tank (RWST) supply), CYAPCO determined that the ECCS flows may be insufficient to assure adequate core cooling. During the recirculation phase of the accident, water is drawn from the

{ containment sump using the residual heat removal (RHR) pumps. If the re0ctor coolant system (RCS) pressure is above approximately l

150 psig (RHR pump shut-off head), the RHR pumps must be aligned,

[ per procedure, to feed either the HPSI or charging pump. For the deluge line break, water would be diverted from the high pressure 1

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(HPSI or charging) pump suction and would flow out the break and ,

result in high pressure and/or RHR pump cavitation. For RCS i pressures bel.ow 150 psig, the high pressure pumps are not used and the RHR pump will supply flow directly to the core deluge system with the majority of the flow being lost directly through the break. Thus, the current Haddam Neck Plant Emergency Core Cooling system arrangement does not assure adequate core cooling during the recirculation phase of a LOCA for a core deluge line break.

3.2 Proposed Solution The 1.icensee has proposed, as an interim response to the core deluge line break problem, a change to the present emergency core cooling system configuration by repositioning flow control valve RH-FCV-796 from its currently open position to a partially open, locked position.

Additionally, a procedural modification was proposed to initiate char-ging system flow when the RHR flow exceeds 600 gpm during the re-circulation phase of a LOCA. The licensee stated that these modifi-cations would assure adequate core cooling durir.g the recirculation a

phase for all LOCAs.

Presently, Technical Specification 3.6 requires that flow control valve RH-FCV-796 be left in a locked open position during operation to prevent spurious valve movement. Locking open of this valve was re-quired to comply with Branch Technical Position EICSB-18 provisions regarding application of the single failure criterion to manually controlled, electrically-operated valves. When this Technical Specification was originally submitted and approved, the licensee believed that RHR runout would not occur with the valve locked fully open. However, as discussed in Section 2.1, it has n'ow been detemined that RHR pump runout could occur for the core deluge line break.

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5-The licensee has proposed locking the valve in a partially open po- j sition to assure continued compliance with Branch Technical Position EICSB-18, while protecting against pump runout and also insuring adequate flow during the recirculation phase of a LOCA. The licensee has established a core deluge flow range of 1050-1600 gpm(atzero system back pressure) through the throttled valve. The maximum flow value (1600 gpm) provides protection for pump runout; the minimum flow (1050 gpm was chosen to assure adequate cooling during the recirculation phase.

To demonstrate the adequacy of the minimum flow requirement, the licensee provided a conservative assessment of the peak cladding temperature for the worst case small break during the recirculation phase. The analysis assumed the earliest time for entering the recirculation phase based upon operation 'of all the emergency core cooling pumps in order to maximize the core decay heat. In addition, initial power operation at 102% of full power and a decay heat based upon 120% of the ANS decay heat standard was assumed. To minimize injection flow, the licensee assumed RCS pressure

/ was constant at 133 psia which results in the minimum injection flow, in-cluding single failure considerations, for the worst case break. Based upon these assumptions, the licensee predicted a peak cladding temperature l of 1760*F.

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The staff finds that the licensee's assumptions yield a conservative as-i sessment of the peak cladding temperature. Accordingly, since the conser-vatively calculated peak cladding temperature is less than the 2300*F limit specified by the Interim Acceptance Criteria, the minimum flow of 1050 gpm assures adequate core cooling.

To substantiate that the valve is throttled properly, the licensee committed to perform a flow test prior to plant operation with the

! revised Technical Specification. During the flow test, the reactor will be placed in mode 4 with the RCS in its normal RHR lineup. To perform the test, the RHR heat exchanger bypass valve (RH-FCV-602)

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,4 will be fully closed thereby diverting flow through RH-FCV-796 RH-FCV-796 will then be throttled until flow through the core deluge line is within the acceptance criteria of the procedure. The valve will then be marked, a locking collar will be manufactured and installed, and the test will be repeated to assure that the flow remained acceptable.

The licensee has also modified its existing emergency procedures which dictate operator action upon switching to recirculation.

Should RHR flow exceed 600 gpm upon switching to recirculation, the operator will know that there is either a large break LOCA or a break in the core deluge line. The oper'ator will be directed to start one )

charging pump for this situation. In this manner, adequate flow will be available for all postulated LOCAs. CYAPC0 stated that it will indoctrinate the plant operators in the revised procedure prior to their first shift of power operation following implementation of the proposed amendment.

. The licensee further committed to revise existing procedures to re-mind operators of the accessibility of RH-FCV-796 (absent radiolo-gical implications) and the potential for manual manipulation to l adjust ficw delivery to the core should the procedurally directed responses appear not to provide expected results.

The staff has reviewed the infonnation provided by the licensee and finds that the proposed change to Technical Specification 3.6 is acceptable. Specifically, the staff finds that the flow rates for the throttled valve provides sufficient flow to assure compliance L with the, Interim Acceptance Criteria,and will also limit the RHR-flows to protect against pump runout. The staff examined the poten-tial flow configurations and concluded that adequate flow would be provided for all postulated LOCAs. The staff's review also concluded that the proposed flow test will provide further assurance that the valve will be throttled properly.

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The procedural modifications proposed by the licensee are also deemed acceptable. The specified actions will provide adequate flow, via the charging pump, for the core deluge line break; for a charging line break, adequate flow would be provided by either the

. HPSI pump or RHR pump through the core deluge line dependent upon the RCS pressure. For other postulated LOCAs, the system lineups will still assure adequate flow. Provisions for local, manual manipulation '

of RH-FCV-796 is appropriate for the unexpected case where procedurally

. directed actions appear not to provide expected results.

The licensee stated in Reference 1 that it is evaluating permanent measures to respond to the core deluge line break problem. Since the licensee has stated that the proposed amendment request is an interim response for use during the remainder of Cycle 14 operation, the staff requires the licensee to propose and implement its permanent solution to the core deluge line break problem prior to operation of Cycle 15.

/ 3.3 Stress Analysis The licensee has stated in his December 17, 1986 submittal, that a stress analysis was conducted on the core deluge 4" Schedule 160 p'iping attar'ied to the reactor vessel upper head. According to the licensee, this analysis shows that the maximum stress is j approximately 0.58 of the code allowable. Also, the licensee asserted that due to the extensive restraints and the inherent rigidity of the piping, a double ended guillotine break with complete offset is unlikely and that the break flow rate assumed

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4 in the thermal-hydraulic analysis assumes a double ended complete offset guillotine break.

Based on the information provided in the licensee's submittal and during the December 15, 1986 meeting with the licensee, the ,

staff agrees with this assessment. However, the staff notes that 4

thissectionofpiping,despiteitscategoriza'tionasASME i

Class 1 and therefore subject to Section XI requirements, is j not included in the licensee's ISI program and has never been inspected. During the December 15, 1986 meeting, the licensee stated his intention to inspect one fourth of the 32 welds in '

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this piping during the next refueling outage. We concur with this approach and strongly reconnend that the other welds be expeditiously inspected. We have no special concerns regarding ongoing degradation in this piping but believe that the nonnal inservice inspection requirements should be followed, especially in light of this system's importance in LOCA mitigation and the fact that it has never been inspected. Furthermore, we require that the stress analyses infonnally presented during the .

December 15, 1986 meeting, be formally submitted for staff review.

This analysis may be submitted within 90 days of issuance of this amendment.

3.4 Sunnary i

Based upon the foregoing, the staff finds the proposed licensing f

amendment to revise Technical Specification 3.6 " Core Cooling j Systems" to reposition and block Flow Control Valve RH-FCV-796 in a throttled position, coupled with the proposed procedural modifications, to be acceptable for the remainder of Cycle 14 operation. The staff requires the licensee to propose and implement its pennanent solution to the core deluge line break problem prior to operation of Cycle 15.

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l 4.0 EMERGENCY CIRCUMSTANCES '

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On December 12, 1986, the licensee discovered that for an inter-mediate break loss-of-coolant accident in the core deluge lines, adequate core cooling may not be obtained during sump recirculation.

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The performance of the Haddam Neck Plant emergency core cooling systems (ECCS) are governed by the Interim Acceptance Criteria of 10 CFR 50, Appendix K. CYAPC0 has determined that, to satisfy the IAC, a change to its operating license (i.e., Technical Specifications), is required before operation with the planned change would be pennitted. As previously communicated to the NRC, CYAPC0 intends to shut down the Haddam Neck Plant reposition the valve and perform a system flow test.

i Absent issuance of the requested amendment, the Haddam Neck Plant will not be able to return to operation. Pursuant to 10 CFR 50.91(a)(5),

an emergency situation exists in that failure to act in a timely manner will result in an unnecessary shutdown (in this instance failure to permit restart).

CYAPC0 also provided an explanation as to why the emergency situation occurred and why'it could not tz avoided. CYAFC0 stated that the Haddam Ncck Plant had been pres ,1y reviewed and found te satisfy i the IAC. CYAPC0 was unaware of and had no reason to expect the

! possibility this condition existed, but it recently completed 6 comprehensive Probabilistic Safety Study designed, among other' tasks, i i to identify conditions such as the core deluge break. In any event, upon promptly notifying the NRC of this condition, CYAPC0 diligently pursued an acceptable response to this condition. Thus, CYAPC0 has made its best efforts to pursue this matter and make a timely application for the requested amendment.

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The staff has reviewed the licensee's assessment and concludes that the licensee had no prior knowledge of the emergtacy circumstances and acted in a timely fashion to pursue resolution of this issue.

I t Based upon the information in the licensee's letters dated December 17 and December 19, 1986, the staff has determined that the above circumstances '

constitute an emergency situation since, if no action were taken, plant shutdown would (in this case, an inability to restart) be required.

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4.1 No Sionificant Hazards Consideration Determination In accordance with 10 CFR 50.S2, the Commission may make a final deter-mination that a license amendment involves no significant hazards consid-erations if operation of the facility in accordance with the amendment would not:

(1) involve a significant increase in the probability or conse-quences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident praviously evaluated; or (3) involve a significant reduction in a margin of safety.

The information in this section provides the staff's evaluation of this license amendment against these criteria:

. A. Involve a sionificant increase in the probability or consequences of an

accident previcusly evaluated.

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CYAPC0 has assessed whether the proposed change could affect ECCS cap-atiilities with respect to accidents other than that involving the core deluge system. CYAPC0 found that if valve RH-FCV-796 is throttled as proposed, and a charging pump is activated during sump recirculation, there will be no significant adverse effect on ECCS-related equipment (i.e., pumps, valves)orECCSresponseduringtheinjectionphase. In addition, because greater flow will be directed to other coolant delivery systems during the sump recirculation mode, and other ECCS response phases, i.e., high and low pressure injection, are not affected by this

action, there can be no adverse effect either as to the probability or consequences of previously evaluated accidents.

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1 B. Create the possibility of a new or different kind of accident from any accident previously evaluated.

ECCS response to other previously postulated accidents remains within '

previously assessed limits of flow distributions and flow paths. Further, all systems and coolant delivery mechanisms remain, respectively, within their i

applicable performance limits (i.e., pump cavitation will not occur) and flow delivery capabilities to provide adequate core cooling. Thus, system and component performance is not adversely affected by this change, thereby assuring that the design capabilities of those systems and components are not challenged in a manner not previously assessed so as to create the a

possibility of a new or different kind of accident. In addition, because the subject valve will be locked in a predetermined position, there will be no possibility that the valve itself may actuate inadvertently to cause a different type 'of accident than previously evaluated.

Operation of RH-FCV-796 in the partially open position should not create a new kind of accident because throttled operation of this valve is a normal plant operation. During every refueling RH-FCV-796 is throttled

to control the amount of decay hat removed from the reactor coolant system. Operation in this mode has not created any reported operational

. problems with the valve integrity.

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l The manufacturing and placement of the collar should not result in a new or different kind of accident. On December 20, 1986 CYAPCO manufactured the collar for RH-FCV-796 and, having installed the collar, verified that the flow rate through the RHR system was with the acceptance criteria. At power, the valve position can not be changed because power is removed from

, y the valve operator. There is no flow through,this system when the reactor l

1s at power.

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C. Involve a significant reduction in a marcin of safety.

The proposed change will not cause existing Technical Specification opera-tion limits or system performance criteria to be exceeded. Throttling valve RH-FCV-796 will redistribute certain coolant flows in the event of ECCS actuation and ECCS perfomance will not be degraded by this action.

In fact, this action will improve certain aspects of ECCS perfor-mance, primarily by preventing pump cavitation during the sump recirculation phase of ECCS response. In addition, the proposed change will assure that the single failure criterion will be satisfied in the event of a postulated break in the core deluge system. Thus, the inherent safety margin provided i by the defense-in-depth. intent of the single failure criterion will not be compromised.

4.2 State Consu1'tation l

Mr. K. McCarthy, Director, Radiation Control Unit, Department of Environmental Protection, State of Connecticut, was contacted concerning the proposed emergency license amendment on December 19 and December 22, l

1986. After a discussion of the subject amendment, Mr. McCarthy indicated

, that all his comments have been resolved.

5.0 ENVIRONMENTAL CONSIDERATION

, This amendment involves a change to a requirement with respect to the l

installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance require-ments. The staff has determined that the amendment involves no significant l

increase in the amounts, and no significant change in the types, of any l effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The l

Commission has detemined that this amendment involves no significant hazards consideration. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environ-

mental assessment need be prepared in connection with the issuance of this

amendment.

6.0 CONCLUSION

The staff has concluded, based cn the considerations discussed above, that:

(1) the amendment does not (a) significantly increase the probability or consequences of an accident previously evaluated, (b) increase the possi-bility of a new or different kind of accident from any previously evaluated or (c) significantly reduce a safety margin and, therefore, the amendment does not involve significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of the amendment will not be inimical to the common defense and the security or to the health and safety of the public.

7.0 REFEP.ENCE

1. J. F. Opeka (CYAPCO) letter to C. I. Grimes (NRC), dated September 30, 1986,Sub.iect: Small Break LOCA Permanent Resolution, Request for

. Extension of Single Failure Exemption.

2. EdwardJ.Mroczka(CYAPCO)lettertoC.I. Grimes (NRC), dated

. December 17, 1986,

Subject:

Proposed Revision to Technical Specifications - Flow Control Valve Repositioning.

f 3. Edward J. Mroczka (CYAPCO) letter to C. I. Grimes (NRC), dated l December 19, 1986,

Subject:

Proposed Revision to Technical Specification - Flow Control Valve Repositioning.

8.0 ACKNOWLEDGEMENT I

This Safety Evaluation has been prepared by R. Jones, PBRS, C. Cheng, PBEB and F. Akstulewicz, ISAPD.

I Dated: December 24, 1986 l

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