ML20206G400

Safety Evaluation Supporting Amend 3 to License NPF-57

ML20206G400
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Site:	Hope Creek
Issue date:	04/07/1987
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8704140542
Download: ML20206G400 (8)

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Me4 9'o UNITED STATES NUCLEAR REGULATORY COMMISSION

[Td )"w >[ gc WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 3 TO FACILITY OPERATING LICENSE NO. NPF-57 PUBLIC SERVICE ELECTRIC & GAS COMPANY  ;

ATLANTIC CITY ELECTRIC COMPANY HOPE CREEK GENERATING STATION DOCKET NO. 50-354

1.0 INTRODUCTION

By letter dated May 30, 1986, as supplemented by letters dated December 24,  !

1986, and February 6, 1987, Public Service Electric and Gas Company (PSE8G), requested changes to the Hope Creek Generating Station (HCGS)

Technical Specifications (TS) to permit reactor operation with one of the two recirculation loops out of service (single-loop operation, (SLO 1). j Presently, the TS require that the reactor be in at least hot shutdown if an idle recirculation loop cannot be returned to service within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Resolution of Generic Issue B-19 regarding themal-hydraulic stability has provided a basis to pemit operation in the single loop mode with appropriate restrictions relatino to stability concerns. General Electric Company (GE), in SIL No. 380, Revision 1, addressed these con-cerns by providing the boiling water reactor licensees generic ouidance for actions which suppress thermal-hydraulic instability induced neutron flux oscillations.

The proposed changes requested by the licensee consist of: (11 deletion of the Technical Specification requirement restricting SLO which involves limiting the allowable pump speed during SLO, increasing the Minimum Critical Power Ratio (MCPR) Safety Limit b priate Average Power Range MonitorFlow (APRM)

Biased yScram 0.01,Trip establishing setpoints, appro revising the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits and revising the Rod Block Monitor (RBM)/APRM Control Rod Block setpoints; (21 for single-loop operation, incorporating requirements in the Technical Specifications which should result in the detection and suppression of themal-hydraulic instability induced neutron flux oscilla-tions if they should occur; and (3) including an applicability section and appropriately revising SURVEILLANCE requirements and the BASES.

The proposed amendment pas noticed in the FEDERAL. REGISTER on January 28, 1987 (52 FR 2889). The May 30, 1986, letter had requested that the words "for initial core loading only" be added to the double-asterisk footnote on Table 3.3.6-2 on page 3/4 3-59 of the Technical Specifications.

Initial core loading for HCGS was completed on April 27, 1986.

In a letter dated February 6, 1987, the licensee requested the note be removed in its entirety, because the note, if revised as requested in 0704140542 070407 4 FDR ADOCK 0500

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the May 30, 1986, letter, would no longer be applicable in the post-initial core load condition. The request to delete the note clarifies the note's applicability; therefore, the February 6, 1987, revision to the amendment

, request does not affect the proposed no significant hazards consideration i discussed in the Federal Register notice.

! 2.0 EVALUATION l

! In its letter dated May 30, 1986, the licensee provided a General Electric (GE) report entitled, " Hope Creek Single-Loop Operation Analysis." This report evaluated the SLO safety issues in order to justify extended opera-tion with one recirculation loop out of service. The staff evaluation of l

the SLO safety issues and the proposed Technical Specification changes follow.

2.1 Minimum Critical Power Ratio (MCPR) Fuel Claddina Intecrity Safety i Limit i

For SLO, the MCPR fuel cladding integrity safety limit is increased j by 0.01 to account for increased uncertainties in the core total flow l

and Traversing In-Core Probe (TIP) readings. The limiting transients 4

were analyzed to verify that there is more than enough margin during SLO to compensate for this increase in safety limit. The proposed 1

change is acceptable. l 1

j 2.2 MCPR Operating Limit - Accidents (Other Than Loss of Coolant Accident (LOCA)) and Transients Affected by One Recirculation Loop out of j 5ervice

• One Pump Seizure Accident

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i A plant specific analysis was not performed for this event. The i licensee stated that GE completed SLO analyses for 23 domestic l BWRs and 4 overseas BWRs thus establishing a data base of SLO l information and analysis techniques which could be applied to I l

HCGS. The results of the SLO analyses for the other plants have ,

always demonstrated that SLO was a non-limiting event and boiling l j transition was not experienced during recirculation pump seizure. l l Also, there was significant margin to the safety limit MCPR, even j though a small fraction of the fuel is pemitted to exceed the I

safety limit MCPR for this event. Because the safety limit MCPR is not exceeded, the 10 CFR 100 "small fraction" limits are

satisfied for pump seizure during SLO. $1nce significant margin j to the safety limit MCPR has been previously demonstrated for I other BWRs, a plant specific analysis for HCGS is not necessary.

Abnomal Operating Transients l The SLO abnormal operating transients were analyzed assuming an i

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initial power of 75% of rated and 60% of rated core flow. The most limiting events were pressurization transients resulting from Feedwater Controller Failure-Maximum Demand (FWCF) and Generator Load Rejection with Bypass Failure (LRBPF). The cor-responding MCPRs were 1.17 and 1.16. Although the increased uncertainties in core total flow and TIP readings resulted in a 0.01 increase in the MCPR fuel cladding integrity safety limit i to 1.07 during SLO, the limiting transients analyzed in the GE report indicate that there is more than enough MCPR margin during SLO to compensate for this increase in safety limit. For SLO at

' off-rated conditions, the steady-state operating MCPR limit is i established by flow dependent Kf curves.

b Since the maximum core flow runout during SLO is only about 60%

of rated, the current flow dependent MCPR limits which are gener-ated based on the flow runout up to rated core flow are also I adequate to protect the flow runout events during SLO. Since the SLO transient analysis is bounded by the two-loop transient i'

analysis, power dependent MCPR curves used for two-loop operation are also applicable for SLO.

I Rod Withdrawal Error

( The rod withdrawal error at rated power is analyzed in the FSAR

! for the initial core load. This analysis was perfonned to demon-

- strate that, even if the operator had ignored all instrument indications and alanns during the course of the transient, the rod block system would stop rod withdrawal at a MCPR which is higher than the fuel cladding integrity safety limit. Correction j of the rod block equation for single-loop operation assures that the MCPR safety limit is not violated.

Additionally SLO results in backflow through 10 of the 20 jet ,

i pumps while flow is being supplied to the lower plenum from the l l

10 active jet pumps. Because of this backflow through the '

inactive jet pumps, the present rod block equation and Average

PowerRangeMonitor(APRM)settingsweremodified.Thelicensee I has also modified the two-pump rod block equation and APRM l j

settings that exist in the TS for SLO. l The staff finds that one-loop transients and accidents other than

~; LOCA, which is discussed in Section 2.4 of this evaluation, are bounded by the two-loop operation analysis, which has been found to be acceptable.

2.3 Loss of Coolant Accident Analysis The licensee evaluated the LOCA for SLO. That evaluation utilized the GE methodology outlined in NEDO-20566-2, Rev.1. Results show the maximum average planar linear heat generation rate

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4 (MAPLHGR) needed to be multiplied by a factor of 0.86. The methodology for the SLO MAPLHGR assumes a boiling transition time of 0.1 seconds. The evaluation methodology was approved by the staff in a letter dated March 5,1986 (H. Berkow to J. Ouirk, GE).

The licensee's use of this methodology and its evaluation are acceptable.

2.4 Stability Analysis With one recirculation loop not in service, the primary contri-

buting factors to stability performance are the power / flow ratio and the recirculation loop characteristics. At forced circulation with one recirculat, ion loop not in operation, the reactor core stability is influenced by the inactive recirculation loop.

Staff evaluations have considered whether increased noise .

in SLO was being caused by reduced stability margin as SLO core flow was increased. Results of analyses and tests indicate that the SLO stability characteristics are not significantly different from two-loop operation. At low core flows, SLO may be slightly less stable than two-loop operation but as core flow is increased and reverse flow is established, the stability performance is similar.. At higher core flows with substantial' reverse flow in the inactive recirculation loop, the effect of cross flow on the flow noise results in an increase in system noise (jet pump, core flow and neutron flux noise); however, core thermal-hydraulic stability margin remains high, similar to two-loop operation. GE has developed Service Information Letter-380, Revision 1 (February 10,1984) informing plant operators how to recoonize and suppress unanticipated oscillations when encountered during  ;

plant operation. The NRC staff has approved the GE generic stability analysis for application to SLO in a letter dated April 24, 1985 (C. Thomas, NRC to H. Pfefferlen, GE), provided that the recommendations of SIL-380 have been incorporated into the plant TS.

The staff compared these GE recommendations with the proposed HCGS TS for SLO. The proposed changes are in conformance with the SIL-380, Revision I recommendations and are, therefore, acceptable.

2.5 Jet Pumo Surveillance Significant increase in APRM noise and core plateop fluctua-tions have been observed in some plants during operation at high flow rates while in SLO. Even though the fluctuations are no longer associated with thermal-hydraulic stability (see Section l 2.4), there is a remaining concern that excessive vibration l leading to mechanical failure of a jet pump could result. Since this could lead to unacceptable consequences if the condition O

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-S-existed during a LOCA, we require that an acceptable jet pump surveillance requirement be in place during single-loop operation.

A similar concern was addressed in IE Bulletin 80-07 for the resolution of a generic problem relating to the integrity of the t hold down beams for BWR jet pumps. In response to this bulletin, i many licensees incorporated jet pump surveillance requirements into the Technical Specifications. Hope Creek has incorporated such requirements into its Technical Specifications and we find them acceptable for assurance of jet pump integrity while in SLO.

2.6 Technical Specification Chances 1

The proposed TS changes for SLO are as follows:

1. Incorporation of the safety limit MCPR for SLO in TS 2.1.2, Bases for Safety Limits, Section 2.1, Bases for APRM Set Points, Section 3/4.2.2, Bases for Control Rods Section 3/4.1.3, and Bases for Minimum Critical Power Ratio Section -

3/4.2.3. [

2. Revision of the APRM scram, APRM Rod Block and Rod Block Monitor setpoints and allowable values to include SLO in

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Table 2.2.1-1, TS 3.2.2 and Table 3.3.6-2.

3. Incorporation of uncertainties for core total flow and TIP l readings for SLO in the Bases Table B2.1.2-1.

l 4 Incorporation of a MAPLHGR multiplier for SLO in TS 3.2.1 and l Bases for Average Planar Liner Heat Generation Rate, Section 3/4.2.1.

5. Incorporation of 0.1 see time for loss of nucleate boiling during SLO in Bases Table B 3.2.1-1 of significant parameters to LOCA,

6. Revision of the jet pump surveillance requirements, TS 4.4.1.2 i to include SLO and baseline data.

7. Revision of the section describing the recirculation system recirculation loops, T.S 3/4.4.1 to include SLO.

8. Revision of Figure 3.4.1.1-1, which illustrates thermal power vs. core flow to be compatible with the TS.

9. Revision of section 3/4.4.1 Bases for the recirculation system to include factors necessary for SLO. Also revised in this section was the basis of temperature differences greater than 145'F between the reactor vessel bottom head coolant and coolant in the upper region of the reactor vessel as related to vessel stress.

F.

We have reviewed the changes and find them consistent with GE-SIL-280 and the results of the GE analysis and also with the SLO TS approved for other BWR 4/5s (e.g., Susquehanna 1

, and 2). Accordingly, the proposed changes are acceptable.

The licensee also proposed other TS changes as follows:

10. TS 3.4.1.3 and 4.4.1.3 to be restricted in applicability to two loop operation and also to reflect loop flows instead of recirculation pump speeds.

11. TS 3.2.2 footnote to be revised to extend above 90% themal power, the region in which APRM gain adjustment rather than APRM setpoint adjustment is allowed. This revision has also been accepted on other BWRs.

12. Table 3.3.6-2 to be revised to eliminate the footnote for the source range monitor downscale trip setpoint option of 0.7 cps. The footnote was developed for and used during the initial core loading only.

We have reviewed the changes and find them consistent with GE-SIL-380 and the results of the GE analysis and also with SLO TS changes approved for other BWR/4s (e.g., Susquehanna 1 and 2). ,

We have reviewed these changes and find them acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change in the installation and use of a facility component located within the restricted area as defined in 10'CFR Part 20 i and changes in surveillance requirements. The staff has detennined that this amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumula-tive occupational radiation exposure. The Connission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accord-ingly, this amendment meets the elioibility criteria for categorical exclusionsetforthin10CFR51.22(c)(9). Pursuantto10CFR51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

4.0 CCNCLUSION The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register (52 FR 2889) on January 28, 1987, and consulted with the state of New Jersey. No public connents were received, and the state of New Jersey did not have any connents.

We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Ccmmission's regula-tions and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Don Katze, Reactor Systems Branch, DBL David Wagner, BWR Project Directorate No. 3 DBL Dated: April 7,1987

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AMENDMENT NO. 3 TO FACILITY OPERATING LICENSE NO. NPF-57 HOPE CREEK GENERATING STATION DISTRIBUTION:

No. 50-3 J NRC PDR^

Local PDR PRC System NSIC BWD-3 r/f DWagner(2)

EHylton EAdensam Attorney, OELD CMiles RDiggs JPartlow EJordan BGrimes LHarmon TBarnhart(4)

EButcher

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