ML20214K955

Application for Amend to License NPF-57,revising Tech Spec Tables 2.2.1-1,3.3.2-1 & 3.3.2-2 for Main Steam Line Radiation hi-hi Full Power Radiation Background Levels & Associated Trip Setpoints.Justification Encl.Fee Paid

ML20214K955
Person / Time
Site:	Hope Creek
Issue date:	05/22/1987
From:	Corbin McNeil Public Service Enterprise Group
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20214K962	List: ML20214K955 ML20214K972
References
NLR-N87091, NUDOCS 8705290122
Download: ML20214K955 (10)

Text

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db -  % A. g Public Service Electric and Gas Cornpany Corbin A. McNeill, Jr. Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609339-4800 Senior Vice President -

Nudar May 22, 1987 NLR-N87091

. United States _ Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

REQUEST FOR AMENDMENT FACILITY OPERATING LICENSE NPF-57 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 In accordance with the requirements of 10 CFR 50.90, Public Service Electric and Gas Company (PSE&G) hereby transmits a request for amendment of Facility Operating License NPF-57 for Hope Creek Generating Station (HCGS). In accordance with the requirements of 10 CFR 170.21, a check in the amount of $150.00 '

is enclosed. In accordance_ with the requirements of 10 CFR 50.91(b)(1), a copy of this request has been sent to the State of New Jersey as indicated below.

This amendment request revises Technical Specification Table 2.2.1-1, 3.3.2-1 and 3.3.2-2 for the Main Steam Line Radiation -

High - High full power radiation background levels and.

associated trip setpoints (see Attachment 2). These changes permit PSE&G to test the feasibility of a Hydrogen Water Chemistry System as a mitigator of Intergranular Stress Corrosion Cracking (IGSCC) of stainless steel recirculation-piping at HCGS. Attachment I contains further' discussion and justification for these proposed revisions. This submittal includes one (1) signed original, including affidavit, and thirty-seven (37) copies pursuant to 10 CPR 50.4(b)(2)(ii).

Should you have any questions on the subject transmittal, please do not hesitate to contact us.

Sincerely, l

p Encicsure (check) j} 4( 0\

Affidavit r Attachments (2) [L 9 B705290122 B70522 PDR ADOCK 05000354 1 P PDR i L

- . . . .- - - . . .. . _ . = . .

l Document' Control' Desk 2 5-22-87 i

C' Mr. G. W. Rivenbark' USNRC Licensing Project Manager.

Mr. R. W. Borchardt USNRC Senior Resident Inspector 1 .

Mr. D. M.. Scott, Chief Bureau of Nuclear Engineering Department of Environmental Protection

-380 Scotch Road Trenton, NJ 08628 i'

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Ref: HCGS LCR 87-08

-STATE OF NEW JERSEY -)

) SS.

COUNTY OF SALEM )

~

Corbin'A..McNeill, Jr.,_being' duly sworn according to law deposes

and says:

I.am Senior Vice President of Public Service Electric and Gas Company, and as such, I find :the matters set forth in' our letter dated May 22, 1987 , concerning Hope Creek Generating Station License Change Request 87-08, are true to the best of my knowledge, information and belief.

A s N Subscribje and Sworn to before,me this 2d day of M 1987 d'

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c(Notary Puplic of New Jersey lARMNE Y. BEARD Notory Public of New Jersey My Commission Empires May 1,1991 My-Commission expires on

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l e Attachment 1 i PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS LCR 87-08 HOPE CREEK GENERATING STATION Page 1/6

• FACILITY OPERATI'NG LICENSE NPF-57

. DOCKET NO. 50-354 I

i I. Description of the Chance Revise Technical Specification ( TS) Table 2. 2.1 -1, Reactor Protection System Instrumentation Setpoints. Table 3.3.2-1 Isolation Actuation Instrumentation and Table 3.3.2-2, Isolation Instrumentation Setpoints to footnote a discussion regarding the hydrogen injection test and its affect on the Main Steam Line Radiation - High High trip function.

l These revisions will temporarily permit full power radiation background levels and trip'setpoints to be changed based on either calculations or measurements of actual radiation levels resulting from the hydrogen injection test discussed below.

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II. Reason for the Chance Public Service Electric and Gas Company ( PSE&G) intends to i'

perform a hydrogen injection test ( currently scheduled for the late July, 1987) on the primary coolant' system at Hope Creek Generating Station ( HCGS) . The purpose of'this-test '

! is to tetermine the feasibility of hydrogen water chemistry 4 contrci as a means of reducing Intergranular Stress

i Corrosion Cracking (IGSCC) of stainless steel piping.

l The test involves the addition of hydrogen to the primary

? coolant at increasing increments over a range of l approximately 0 to 70 standard. cubic feet per minute - ( scf m) .

As a result, the radiolysis of water is suppressed, thereby lowering the free oxygen content in the reactor coolant.

The reduction of free oxygen eliminates one of the necessary ,

causative agents of IGSCC ( see Final Safety Analysis Report '

( FS AR) Section 5.2.3),

f A by-product of oxygen suppression by hydrogen addition'is l an increase of radiation levels from the main steam lines l caused by Nitrogen-16 ( N-16) . The increased carry-over of l nitrogen is due to a conversion of N-16 from a soluble-form ,

to a gaseous form'in the reactor. The requested revisions  !

to the TS tables, identified above and included in ' this

! transmittal as Attachment 2, permit a temporary increase in the Main Steam Line Radiation - High '- 'High scram and

, isolation setpoints to allow operation with expected higher i l

radiation levels resulting from hydrogen injection. The l

! Main' Steam Line' Radiation - High - High'setpoint will remain ,

I i at 3 times the background level and the allowable value will remain at 3.C times the background. level; however, the l

) background radiation level used to determine these setpointa j i

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i LCR 87-08 Page 2/6 will be increased prior to the' hydrogen injection test based l - on a calculation of the anticipated background level. The ,

4 changes also permit the full load background radiation level

to be adjusted during the testito correct for uncertainties in the initial computation. At'the maximum planned hydrogen injection rate, an increase of approximately.one to five times the normal main steam line' background radiation levels is expected. The pretest main steam line radiation monitor setpoints will be restored promptly following the conclusion of the hydrogen injection. test and whenever power is decreased to less than 22% of rated thermal power ( this latter restriction is in accordance with guidance provi de d in References 1 and 2). The hydrogen injection -test .will be -

i discontinued whenever reactor power is reduced to less than 22%.of rated thermal power.

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) i i III. Hydrocen Test Summarv 4

PSE&G has decided to employ an experienced contractor to develop, prepare and execute the hydrogen injection test under the management of HCGS personnel. The contract'has not yet been awarded, although contractor selection is-

, expected during the month of May 1987; therefore, complete l test details have not yet been finalized. However, the i following information is available on the test itself:

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1) Rated thermal power at least 90 percent in order to develop full-power radiation background levels i and optimize hydrogen effectiveness,

2) Hydrogen will be supplied from a tank truck

outside the turbine building near the condensate l storage tank ( CST), to the turbine building i through flex hosing and/or tubing,

3) Hydrogen injection points into the feedwater i system will be on the suction side of the i secondary condensate pumps,

! 4) Radiation levels will be monitored at various

! points inside HCGS ( such as the administrative

offices, the turbine deck, the control room and at l selected points within the plant) as well as outside the plant ( such as the administration building, parking lots and at various points in l the site boundary and immediate general population areas) using monitors already in place and -

l additional monitors if required, i

5) Oxygen will be added to the offgas system prior to entering the catalytic recombiners to

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1- LCR 87-08 Page 3/6 scavenge.any free hydrogen and minimize: hydrogen build-up, These details have been developed'as general guidelines and 4

were included in request . for proposals - f rom contractor's.

Further refinement of these details, as well as additional ~

test information, will utilize the experience of'the contractor selected. PSE&G is also working with. selected r utilities which have previously performed a hydrogen water chemistry test,. including Carolina Power . and Light. Company's Brunswi'ck -Station and Georgia Power Company's Hatch Station, such that the HCGS test incorporates useful experiences and I

averts unnecessary' problems.

I V. Environmental 'moact Annraisal and Onsite Dose Imnact 4

From an analysis of previous tests, radiation levels can be expected to-increase approximately one- to five-fold.

Therefore, extensive radiatio preplanning willcidentify the proper areas to monitor radiation levels. During the test radiation surveys will be conducted at regular intervals to monitor the actual doses. The surveys will . be performed by qualified PSE&G Radiation Protection technicians using approved site procedures. The objectives of the survey 4

program will be:

1) To provide data for shielding design should additional shielding be necessary,

2) To determine radiation levels in and around the facility as well as at the site boundaries, I

3) To determine the potential effects on operation

and maintainence a c ti vi ti e s , and,

! l j 4) To determine the impact on the site ALARA ( as low as reasonably achievable) program.

k As stated in Paragraph II above, the increased radiation )

) levels are a result of an increased carry-over of N-16 in

j. the main steam, not from increased N-16 production. Due to 1

the short half-life of N-16 ( approximately' 7 seconds), PSE&G

.does not expect hydrogen injection to have a significant effect on the gaseous effluent release rates. Therefore, i .the increased radiation levels will have no discernible effect on the health and safety of the public. This'will be confirmed during the test through the~use of offsite

} monitors. Radiation protection measures will be implemented I to maintain doses to plant personnel ALARA. These measures include:

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1 LCR 87-08 Page 4/6 l

1) Scheduling the test'during a weekend and at nights to the extent feasible in order to minimize the number of on-site-peraonnel affected. l l

2) Establishing access control by Radiation :i I

Protection personnel in accordance-with existing site procedures, and,.

3) Training of operations personnel in test j- procedures prior to ths start of'the test to ensure-efficient performance of duties.

V. protective Measures Reaardina Hydrocen I

}i Compressed hydrogen will.be supplied to the plant site in gaseous form via tube trailer. The tube trailer will'be 3

! located in accordance with the requirements of NFPA. Code No.

50A, " Gaseous Hydrogen Systems at Consumer Sites," Section

5. 0, outside the turbine building near CST. A detailed  ;

review of the . hydrogen water chemistry test will be performed to insure proper bulk storage of hydrogen onsite during the test. This review will included an assessment of 4

10 CFR 50.59'to determined whether there ara any unreviewed i safety questions involved.

The hydrogen control and distribution system consists of supply lines, control valves, a safety relief valve, an j excess flow check valve, a safety valve and a mechanical

pipe break valve ( Ref. 2). The supply line will be routed i outdoors from'the supply trailer to the turbine building 1 ~

using high pressure flex hose and/or tubing. .The 1ines will i then be routed to the condensate booster pump rooms. The i hydrogen supply system will be leak tested and purged with l an appropriate gas prior to the introduction'of hydrogen.

l To further ensure that comuustible levels of hydrogen are not reached due to leakage, hydrogen area monitors will be utilized during the test ( Ref. 2). The hydrogen monitors will be located in various locations, such~as at the

, condensate booster pumps, near the control valves and/or at various locations along the supply line. The monitors will l' alarm when hydrogen concentration exceeds 2 percent end isolate the hydrogen supply when concentration reaches 4 l percent in order to prevent an explosive concentration from being reached.

1 PSE&G hac determined that the hydrogen supply and distribution system meets thw requirements of Section C. S. d( 2) of Branch Technical Position CMEB 9.5-1, NUREG-0800 (see FSAR Sections 9. 5.1.1.11 ' and 9. 5.1. 6) .

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LCR 87-08 Page 5/6 VI. Sionificant Hazards Analysis The proposed changes to the HCGS Technical Specifications:

a) Do not involve a significant increase in the ,

probability or consequences of an accident previously evo.uated. The only design basis accident which takes cr'.it for the main steam line high radiation scram and isolation setpoint is the Control Rod Drop Accident ( CRDA) as described in FSAR Section 15. 4. 9. Specifically, the Main Steam Isolation Valves ( MSIVs) are assumed to receive an LCR automatic closure signal at 0.5 seconds after detection of high radiation in the main steam lines and to be fully closed at 5 seconds from the receipt of the closure signal.

The main steam line radiation monitors are provided to detect a gross failure of the fuel cladding. When high radiation is detected, a trip is initiated to reduce the continued failure of fuel cladding. At the same time, the main steam isolation valves are closed to limit the release of fission products. The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures in the fuel cladding.

As indicated in the NEDO Report ( Ref erence 1), the consequences of the CRDA are most severe under Hot Standby conditions. In fact, the consequences of the CRDA are increasingly less severe above 10 percent power due to a faster Doppler response and a lower rodworth. Most importantly, above 20 percent power, the consequences of the CRDA are minimal. Since the Main Steam Line Radiation Monitor setpoint will only be adjusted for the purposes of the hydrogen injection test at power levels above 22 percent, there is no significant impact on the probability or consequences of the CRDA. Therefore, the change to the footnotes in the referenced TS tables have no affect on the probability or consequences of an accident pre vi o usly evaluated.

b) Does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not affect the design of any safety-related systems and as such do not affect the performance of any safety functions. The proposed changes do permit the performance of a hydrogen injection test; however, this test does not introduce a new kind of accident since the presence of hydrogen in the primary system has already been analyzed ( see PSAR Section 6.2.5, 10.4.2 and 11.3.2.1) and is already monitored and controlled ( see TS 3/4.3.7.10 and 3/4.11.2.6, re s pe c t i vel y) . Further, as discussed in Paragraph V above, additional protective measures are being applied which assure that the physical

LCR 87-08 Page 6/6 l

presence of test equipment does not create the potential for i a different kind of accident to occur.

Since the TS changes themselves do not affect existing  :

system function, nor do they create a situation which has not been previously analyzed and appropriately designed for,

the change to the TS footnotes shown in Attachment 2 do not create new or different accidents than previously evaluated.

c) Does not involve a significant reduction in a margin of

safety. The proposed temporary. increase in the Main Steam

Line. Radiation - High - High scram and isolation setpoints will be permissable only when reactor thermal power is above

- 22 percent. As discussed in Paragraph VI.a above, the only design basis accident which takes credit for this scram and

! isolation trip function is-the.CRDA. However,.above 20

percent power, the consequences of a CRDA are so minimal that they may be considered negligible ( Ref. 1), and hence, the change in the TS setpoints has no significant effect on i the margins of safety for this accident scenerio.

The proposed change is-necessary to conduct a hydrogen injection test which will increase the carry-over of N-16.

This in turn, will cause the background radiation levels in the main steam system to be i ncreased. As discussed in l Paragraph IV above, several precautionary and preplanning l measures are being taken to maintain plant personnel j exposures ALARA. In addition, radiation levels will be monitored to assure measured radiation levels are within acceptable site ALARA. Due to the relatively short half-life of N-16 ( approximately 7 seconds), gaseous

! effluent release rates will not be significantly affected, i

Therefore, it can be concluded that the proposed change will not present a risk to the public health and safety nor significantly reduce a margin'of safety for plant personnel.

Based upon the discussions in the above three subparagraphs, l PSE&G concludes that the proposed change does not involve a

! significant safety hazard.

VII. References l

1. NEDO-10527, Supplement 1, " General Electric Rod Drop Accident Analysis for Large Boiling Water Reactors" dated July 1972. .

l 2. EPRI Report NP-4500-SR-LD, " Guidelines for Permanent BWR-Hydrogen Water Chemistry Installations" dated March 1986, i

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Attachment 2 LCR 87-08 This attachment contains copies of the Hope Creek Generating Station Technical Specifications, from Facility Operating License NPF-57, which require revision to support the proposed changes discussed in Attachment 1. Revisions are highlighted within balloon enclosures on the existing Technical Specification pages or are typed on separate insert pages. Tho following pages are affected and enclosed within this attachment:

Table 2.2.1-1 Page 2-4, and Page 2-5 Table 3.3.2-1 Page 3/4 3-12, Page 3/4 3-16, and A separate Insert Page for Page 3/4 3-16 Table 3.3.2-2 Page 3/4 3-22, and Page 3/4 3-25 l

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