ML20087E559

Proposed Tech Specs Re Reactor Vessel Hydrostatic Testing

ML20087E559
Person / Time
Site:	FitzPatrick
Issue date:	01/09/1992
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20087E548	List: JPN-92-002, Forwards Application for Amend to License DPR-59,consisting of JPTS-91-014,allowing Performance of Inservice Hydrostatic Pressure & Leak Testing of Reactor Vessel at Temps Exceeding 212 F,In Ref to Generic Ltr 88-11 ML20087E556 ML20087E559
References
JPTS-91-014, JPTS-91-14, NUDOCS 9201210291
Download: ML20087E559 (20)

Text

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4 ATTACHMENT , to JPN 92-002 PROPOSED TECHNICAL SPECIFICATION CHANGES REACTOR VESSEL HYDROSTATIC TESTING (JPTS-91014)

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 9201210291 920107- , DPR-59 '

l{DR ADOCK 05000333 PDR

a.

JAFNPP' ,

3.5 (cont'd) 4.5 (com'd)

a. Rom and after the date that the HPCI System is- a. When it is determined that the HPCI subsystem is made or found. to be inoperable for any reason,. inoperable the RCIC, the LPCI suosystem, both core continued reactor operation is permissible only spray . subsystems, and the ADS subsystem during the succeeding 7 days unless such system is . actuation logic shall be verTKd to- be operable sooner made operable, provided that during such 7 immediately. The RCIC system and ADS subsystem days all active components of the Automatic logic shall be verified to be operable daily thereafter.

Depressuriza3cn System, the Core Spray System, LPCI System, and Reactor Core isolation Cooling System are operable.

b. - If the requirements of 3.5.C.1 ~cannot be met, the reactor shall be placed in the cold condition and pressure less than 150 psig within 24 hrs.

2. Low power physics testing and reactor operator training snail be permitted with reactor coolant temperature

<212*F with an inoperable component (s) as specified in 3.5.C.1 above.

3. The HPCI system is not required to be operable during hydrostatic pressure and leakage testing ydth reactor coolant temperatures between 212*F_ and 300'F and irradiated fuel in the reactor vessel provioed all control rods are inserted.

Arnendment No. , 1 7. 1 ', 1 ,

'l '

118

JAFNPP ..

~

3.5 (cont'd) : 4.5 (cont'6; during such tirne, the HNI System is operable.

2. . If the requirements of 3.5.D.1 cannot be rnet, the reactor 2. Alogic systern functionaltest.

shall be placed in the cold condition and pressure less

. than 100 psig, within 24 hr. a. When it is determined that one vafve of the ADS is

' inoperable, the ADS subsystem actuation logic for-the operable ADS valves and the HPCI subsystem

'shall be verified to be operabie immedia:ely and at -

least weekly thereafter.

b. When it is determined that more than one relief / safety valve of the ADS is ir@ able, the HPCI System shall be venfied to be operable immediately.

3. l.ow power physics testing and reactor operator training shall be permitted with inoperable components as specified in 3.5.1.a and 3.5.1.b above, provided that reactor coolant temperature is <212*F and.the reactor vessel is vented or reactor vessel head is removed.

4. The ADS is not required to be operable during hydrostatic pressure and leakage testing with reactor coolant temperatures between 212 F axi 300'F and irradiated fuel in the reactor vessel provided all control rods are inserted.

Amendment No. , 1 8, 12U

JAFNPP --

3.5 (Cont'd) 4.5 (Cont'd)

E Reactor Core isolation Cooling (RC1C) Sysb.n E . Reactor Core isolation Cooling (RCIC) System

1. The RCIC System shall be operabia whenever there is 1. RCIC System testing shall be pedcair.ed as follows irradiated fuel in- the reacter vessel and the reactor provided a reactor steam supply is available. If steam is pressure is greater than 150 psig and reactor coolant not available at the time the surveill&nce test is scheduled temperature is greater than 2127 except from the time 10 be performed, the test shall be performed within ters that the RCIC System is made or found to be inoperable days of continuous operation from the time steam for ' any reason, continued reactor power operation is becomes available.

permissible during the succeeding ? days unless the system is made operable earlier provided that during these item Frequency 7 days the HPC1 Systemis operable.

a. Simu!ated Automatic , Once/ operating

2. If the requirements ol 3.5E cannoi be met, the reactor Actuation (and Restart ) cycle shall be placed in the cold condition and pressure less Test than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. Pump Operabnity Once/ month

3. Low power physics testing and reactor operator training sha!I be permitted with . inoperable components as ated h/ month specified in 3.5.E2 above, provided that reactor coolant g;,y temperatureis <2127.

d. Flow Rate Ore:e/3 rnonths

4. The RCIC system is not required to be opeiable during hydrostatic pressure and leakage testing with reactor

e. Tatable M Tested h WMity h os q tim tM re is coolant temperatures between 2127 and 3007 and in the cold condition irradiated fuel in the reactor vessel provided all control g 4

rods are inserted.

ests he not been performed during the preceding _

31 days.

f. Logic System Once/ operating Functional Test cycle Automatic restart on a low water level signal which is Amendment No. f,1 ,1 ,

sutrW to al.;gh water le/eitrip.

t 121

4 JAFNPP

. 3.6 (cont'd)

G (cont'd)

a. <207 wht.' to the left of curve C.

b. < 1007 when on or to the right of curve C.

Specifications 3.5.C,3.5.D,3.5.E and 3.6.E which would become effective because of an increase in reactor coolant temperature above 2127 or pressures above 100 anri 150 psig are not required while conducting the RCS hydrostatic pressure and leakage tests between 2127 and 3007 provided all control rods are fu!!y inserted.

3. Ncn-Nuclear Heatup and Coofdown

3. Non-Nuclear Heatup and Cooldown During heatup by Non-Nuclear means, coolcown fo!!owing

- During heatup by non-nuclear mearu (mechanical), nucles? shutdown and low power physics tests, the reactor cooldown following nuclear shutdown and low power coolant system pressure and temperature shall be physics tests the Reactor Coolant System pressure and recorded every 30 minutes until two consocutive temperature shall be on or to the right of the curve B temperature readings are within 57 of each other.

shown in Figure 3.61 Part 1,2, or 3 and the maximum temperature change during any one hour shall be <1007.

4. Core Critical Operation .

4. Core Critical Operation 1 During all modes of operation with a critical core (except l During all modes of operation with a critical core (except for low power physics tests) the reactor Coolant System for low power physics tests) the reactor Coolant System pressure and temperature shall be recorded within 30 pressurc, and temperature shall be at or to the right of the curve C shown in Figure 3.6-1 Part 1, 2, or 3 and the minutes prior to withdrawal of control rods to bring the reactor critical and every 30 minutes during heatup until maximum temperature change during any one hour shall two consecutive temperature readings are within 57 of be <1007.

each other.

1 l

Amendment No. ,

,1 3, Ih,

/ 137

jai:NPP 4.6 (cont *d) 3.6 (cont'd)

S. Not Used l

5. With any of the limits of 3.6A1 through 3.6A4 above exceeded, either

a. restore the temperature and/or pressure to within the limits within 30 minutes, perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system, and determine - that the reactor coolant system remains acceptable for continued operations; or -

b. be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6. Id:e Recirculation Loop Startup

6. Idle Recirculation Loop Startup Wdhin 30 minutes prior to startup of an idle loop:

When Reactor Coolant System temperature is > 1407 an idle recirculation loop shall not be started unless:

a. The differential temperature between the reactor

a. The temperature differential between the reactor coolant system and the reactor vessel bottom head coolant system and the reactor vessel bottom head drainline shall be recorded,and drain line is < 1457, and When both loops are idle, the temperature difference b. When both loops are ;dle, the differerfjal

b. temperature between the reactor coolant system between the reactor coolant system and the idle and the idle loop to be started shall be recorded, or loop to be startedis1507, or When only one loop is idle, the temperature c. When only one loop is idie, the temperature

c. differential between the idie loop and the operating difference between the idle loop and the cperating loop sha!! be recorded.

loopis < 507.

i Amendment No. 3, 138

JAFNPP ,

h 3.6 (cont'd) 4.6 (cont'd)

7. Reactor Vessel Rux Monitoring The reactor vessel Rux Monitoring Surveitiance Program complies with the intent of the May,1983 rev'.sion to 10 CFR 50, Appendices G and H. The next flux monitoring surveillance capsule shall be removed after 15 effective full power years (EFPYsi and the test procedures and reporting requirements shall meet the requirements of ASTM E 185-82.

B. Deleted B. Deleted C. Coolant Chemistry C. Coolant Chemistry

1. The reactor coolant system radioactivity concentration in 1. a. A sample of reactor coolant sha'l be taken at least water shall not exceed the equilibrium value of 3.1 pCi/gm every 96 hr and analyzed for gross gamma acthty.

of doss equivalent 1-131. This limit may be exceeded, following a power transient, for a n 1ximum of 48 hr.

h Isot@ mW of a We &m W h'l During this sodino activity . transient the iodine be made at least once/ month.

concentrations shall not exceed the equilibrium limits by c. A sample of reactor coolant shal be taken prior to more than a factor of 10 whenever the main steamline startup and at 4 hr intervals during startup and isolation valves are open. The reactor sha9 not be analyzed for gross gamma actmty.

operated more than 5 percent of-its annual power d. During plant steady state operation and following an operation under 'his exception to the equilibrium limits. If oMgas a% Mease @ the Sim Jet M the iodine concentration exceeds the equilibrium limit by , Ejectors) of 10,000 pCi/sec within a 48 hr. penod or more than a iactor of 10, the reactor shall be p! aced in a a power level change of >20 percent of fu!! rated cold condit, ion within 24 hr. -

AreWMWsMNN and analyzed for gross gamma actmty. At icast ihree samples will be taken at 4 hr intervals. These sampling requirements tr.:y be orrutted wtienever the . equilibrium I-131 concentration in the reactor

Jantisless than 0.007 pCi/mi.

Amendment No.

139 A.. . _ _ -._ _ _ _ _ .

j

Jt,FNPP 3.6 (cont'd) 4.6 (conic) ,

2. At least one sa'ety/rc4ief vahm sha!i be disassembled and 2.

inspected once/operatng cycle.

a. From and atter the date that the safnty va've furs-t of one safety /relits va've is made or fcus iroperaNe, .Winued operabon is penh .ty dunng the succeeding 30 days ur8ess such va've is made operable soorer.

b. From and after the time that the safetv valve function on two sa'cty/retief va!ves is nade or found to be inoperable, continued reactor operaton is pewassible only during the succeeding 7 days unless such valves are soorer made operable.

If Specificabon 3E.E.1 and 3.6.E2 are not met, the reactor 3. The integ dy of .S -crogen system and c,ig;vreds

3. which provide tra1ual and ADS actua*Jon of the shall be p! aced in a cold condition witnin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

sa'ety/ relief valves sha*! be &2Tui4 4ed at Icast once every 3 months.

Low power physics testing and reactor operator training 4 An annual report of sa'ety/rc5cf vave faSures and 4.

sha!! be pernitted with inoperable components as cha!!enges wCI be sent to the NRC in wdance vnth specified in ! tem 82 above, provided that reactor coolant Sec en6SA2.b.

temperature is <212F and 17.c reactor vessai is vented or the reactor vesselheF:s arc.cd.

5. The Safety and Safety / Relief Valves are not required to be operable during hyd.tstatic pressure and lea'<age testing with reactor coolant temperatures betwee.12127 and 3007 and irradiated fuel in the reactor vessel provided a!!

controlrods areinserted.

Amendment No. ,

'30,1 ,

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i JAFNPP -

4 3.6 and 4.6 BASES (cont'd)

B. Deleted anrauGg at wup We wweidon laefs such M sampling for isch 4ac analysis can be initiated. The desgn details of such a system must te submitted for evaluation and C. Coolant Chavashy accepted by the Cwniassa.,i t:rior to its implemerf,abon and inu, pciation in these Tet.i . acal Specifications.

' A radioactivity ccrr,whation limit of 20 pCi/mi total iodine can i

l be reached if the gaseous effluerf.s are near the limit as set Since the cu ccetration of radioactmty in the reactor coolant is j forth in Radiological Effluent Technical Specification Seebon not continuously measured, coolant sampiing would be

32.a if there is a fadure or a prolonged shtt.down of the ineffectrve as a means to rapdy detect gross fuel eirua cleanup demineralizer. fai!ures. However, some capatility to detect gross fuel e!cment i faDures is inherent in the radiation monitors in the offgas

! In the event of a steam line rupture Ntside the dqrwe!!, with this system and on the marn steam Enes_

coolant activity level, the resultant radiciogical dose at the site boundary would be 33 rem to the thyroid, under adverse Matenals in the Reactor Coolant System are prearily 304 l meteorologica! wnd?Jur6 assuming no more than 3.1 pCi/gm stainless steel and Zircaloy fuel cladding. The reactor water l cf dose equivalent 1-131. The reactor water sample wi>l be used chemrstry limits are @# died to prevent damage to these

to assure that the limit of Specification 3.6.C is not exceeded. materials. Umits are placed on chloride ccncwcanon and

The total radioactive iodine activity wottd not be expected to conductivity. The most hirgo@d Iblit is that placed on

change rapid!y over a period of 96 hr. In addition, the trend of chfonde cca.cerdration to prevent stress cwcsion craciong of the stack offgas re%ase ra*e, which is continuously monitored, the stainless steel The attached graph, Fig. 4.6-1, i!!ustrates

is a good indicator of the trend of the iodine activity in the the results of tests on stressed 304 stain!ess steel spc6.u s.

recctor coolant. Also dunng reactor startups and large power Failures occurred at concentrabons above the curve; no

changes which could affect iodine levels, sampics of reactor failures occurred at cce w eidu 6 below the curve. Accu dag

! coolant sha!! be analyzed to insure iodine cum,whations are to the data, alfowable chionde cuncoLations could be set

] below allowable levels. Analysrs is required whenever the 1-131 several orders of magrutude above the wahrreled limit, at the l concentratW1 is within a factor of 100 of its a!!owable oxygen cuice d otion (02-03 ppm) eig:riunced dunng power I equiiibrium value. The necessity for continued sampling operatie-i. Zircafoy does not exhibit similar stress wudon following power and otigas transients will be reviewed within 2 failures.

years ofinitial plant s*artup.

However, there are vanotes condFJons under wtsch the Tne surveillance requirements 4.6.C.1 may be satisfied by a dW:fved oxygen content of the reactor coo: ant water could be i continuous monitoring system capable of determining the total higher than 02-03 ppm, such as refueling, reactor startup, and iodine concentration in the coolant on a rea! time basis, and hot standby. During these periods with steamrng rates less Amendment No.

) 149 i

JAFNPP -

3.6 and 4.6 BASES (confd) than 100,000 lb/hr, a more restrictive limit of 0.1 ppm has been startup penods, wtuch are in the category of less than 100.000 established to assure the chloride-oxygen cuisbinations of Fig. Ib/hr, conductmty may exceed 2 rnho/cm because of the incal 4.6-1 are not exceeded. At steaming rates of at least 100,000 evolution of gases and the initial evoktdon of gases and the initial lb/hr, boiling occurs causing deacration of the reactor water, addition of dissolved metals. Dunng this penod of time, when thus maintaining oxygen conca dration at low levels. the conductmty exceeds 2 mho/cm (other than short-term spikes), samples wi!! be taken to assure the chlande When condactmty is in its proper normal range, pH and chionde w n,w.o ation is less than 0.1 ppm.

and other impurities af5cting conductivity must also be within their normal ranges. When and if conductivity becurves The conductmty of the reactor coolant is cerfJnuous!y abovurel, then chionde measurements are made to determine mordtored. The sampLs of the coolant whrch are taken every 96 whether or not they are also out of their normal opwdig values. hr wi!! serve as a reference for calibration of these monitors and This is not necessarily the case. Conductivity could be high due is cuddened adequate to assure accurate readings of the 10 the presence of a neutral saft; e.g., Na 2 SO 4 , which would nct moriitors. If conductmty is within its norma! range, chiandes and have an effect on pH or chloride. In such a case, high other impuritics w!Il also be within their normal ranges. The conductivity alone is not a cause for shutdown. In some types of reactor coolant samples wi!! also be used to determine the water-cooleo reactors, conductivities are, in fact, high due to ch.ondes. Therefore, the seTpug frequency is considered purposeful addition of additives. In the case of BWR*s, however, adequate to detect long-term changes in the ch:onde ion where no additives are used and where neutral pH is mainta:ned, content. Isotopic analyses of the reactor coolant required by conductmty provides a very good rneasure of the quality of the Specification 4.6.C.1 may be pertvoi ed by a gan Te scan.

reactor water. Significant char @ therein provide the operator with a warrung mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with D. Coolant Leakage respect to variabies affecting the boundaries of the reactor coolant, are exceeded. Methods avaliable to the operator for A!!owabie leakage rates of coolant from the Reactor Coolant correcting the condition include operation of the Reactor System have been based on the predicted and expwin ercarry Cleanup System, reducing the input r$f impunties and placing the observed behavior of cracks in pW and on the ability to make reactor in the coid shutdown condition. The major benefit of un Reactor Coolant System leakage in the event of loss of off-cold shutdown is to reduce the tempe ature dependent site a-c power. The normally expected t'ackground leakage due corrosion rates and provide time for the Reac;or Water Cleanup to equipment design and the detecbon capability for detcwTaiig System to reestablish the punty of the reactor coolant. During system Araundinent No.

150

i

. + , j Attachment 11 to JPN 92-002 l

. I SAFETY EVALUATION FOR l PROPOSED TECHNICAL SPECIFICATION CHANGES l REACTOR VESSEL HYDROSTATIC TESTING (JPTS-91 014)

1. DESCRIPTION OF THE PROPOSED CHANGES The proposed technical and editorial changes to the James A. FitzPatrick Technical Specifications are as follows:

A. HYDROSTATIC AND LEAKAGE TESTING CHANGES  !

Pago 118, Specification 3.5.C.3 insort this now specification with the following words:

'The HPCI systom is not requirod to ')e operable during hydrostatic pressure and

• leakago testing with reactor coolant temperatures betwoon 2127 and 3007 and irradiated fuel in the reactor vossol provided all control rods are insorted.'

Pago 120, Specification 3.5.D.4 insert this new specification with the following words:  ;

'The ADS is not requirod to be operable during hydrostatic pressure and leakago -  !

testing with reactor coolant temperatures between 2127 and 3007 and Irradiated fuel in the reactor vossol provided all control rods are insortod."

Page 121, Specification 3.5.E.4 Insert this now specification with the following words:

• The RCIC system is not required to be oporable during hydrostatic pressure and leakage testing with reactor coolant temper 6tures between 2127 and 3007 and

-Irradiated fcol in the reactor vessel provided all control rods are inseriod." ,

Page 137, Specifichtlon 3.6.A.2 Insert the following sentence at the end of tho existing . ~miting Condition for Operation:

' 'Specifi Atlons 3.5.0,3.5.D,3.5.E and 3.6.E which would becomo offectivo l'ocause of i.! an increase in reactor coolant temperature above 2127 or pressures above 100 and 150 psig are not required while conducting the RCS hydrostatic pressure and loakago testt itween 2127 and 3007 provided all control rods a*o fully Insortod.'

i W

e .

Attachmont ll to JPN 92 002 SAFETY EVALUATION Pago 2 of 9 a

Pago 143, Specification 3.0.E.5 Insort this now spocification with the following words:

"The Safety and Safety /Rollof Valvos are not requirod to bo operablo during hydrostatic pressure and leakago testing with reactor coolant temperaturos betwoon 2127 and 3007 and irtadiated fuel in the roactor vcssol provided all contiol rods ato insorted.'

Pago 148, BASES 3.6 and 4.0 A Insort the following paragraphs at tho end of Bates A on pago 148:

"Specifhation 3.6.A.2 identiflos four LCOs that becomo offectivo with inctcased reactor coolant temperature or prossuro but aro not in offect during tho hydrostatic and loakago tests. This is nocessary because, as reactor fluenco increases, the minimum tost temperature and pressuro rises into rangos normally associated with startup or hot shutdown. RCS pressuto and temporature aro usod throughout the Technical Specifications as a basis for establishing plant modo and system operability requiromonts. Somo LCOs and rostrictions cannot bo satisfied during tho test at olevated temperatures. For oxamplo, Specifications 3.5.C.1 and 3.5.E.1 roqulto that HPCI and RCIC bo operablo when reactor pressuro excoods 150 psig and 2127.

HPCI and RCIC cannot be mado operablo during tho test becauso piping normally fillod with steam is fillod with water during tho test.

Hydrostatic and loakago tests shall be terminated botoro the reactor coolant temporaturo excocds 3007. This temporaturo limit is based on providing a 507 band for oporating floxibility betwcon tho 3007 limit and the highest estimated rninimum testing temperaturo at 32 EFPY (approximately 2507).

The protection provided by LCOs applicablo during cold shutdown plus the requiremont that all control iods bo fully inserted aro adoquato to ensuro protection of public health and safety. The hydrostatic test is performed onco evor) 10 years whilo the loakago test is performod after each refueling when condition : aro similar to cold shutdown (1.0., attor the rcactor has boon shutdown and decay heat and the onor_

stored in the coro is very low). The consoquences of accidents (small and largo tv <

LOCAs, MSLB, etc.) are bounded by analysos that ussumo full power oporation.

Spocification 3.5.A requires the low pressuro ECCS systems to bo operable.

Spocifications 3.7.A,3.7.B and 3.7.C requito the containment, SGTS and secondary containment to bo operable. Specifications 3.2.A,3.2.0 and Appendix B, Specification 3.8 requito Instrumentation that initiato contalnmont, low pressuro ECCS, SBGT and secondary containment be operablo. Emergoney power ls required by Specification 3.9.B.

.. t.

• Attachment ll to JPN 92 002 SAFETY EVALUATION Page 3 of 9 B. EDITORIAL CHANGES Pago 137, Specification 3.G.A.2 Add a *29' to the list of amendment numbers at the bottom of this pago. ,

l Page 138, Specifications 3.6.A.5 and 1.6.A.5 l i

Movo Specification 3.6.A.5 from page 137 to the top of page 138.

Insort Specification 4.0.A.5 with the words, 'Not Usod,' at the top of the second column. l Page 139, Specifications 3.6.0,4.6.A.7, and 4.6.B Move Specifications 3.0.B,4.6.A.7, and 4.G.B from page 138 to tho top of pago 139.

Pago 149, BASES 3.6 and 4.6 B and C ,

Move Bases Sections 'B' and *C' from pago 148 to the top of pago 149.

r P3e 150, BASES 3.0 and 4.6 B and C Move the last paragraph and the portion of the 'hlrd paragraph beginning with '... than 100,000 lb/hr, a moro restrictive ..." from pago 149 to the top of pago 150, 11.' PURPOSE OF THE PROPOSED CHANGES t A. HYDROSTATIC MD LEAKAGE TESTING CHANGES ~

The proposed changes revise the Tochnical Spocifications to pormit hydrostatic pressure and loakago testing of the Reactor Codant System (RCS) as requirod by Soction XI of the American Socloty of Mechanlcal Enginoors (ASME) Sollor and Prossuro Vossol (B&PV) Codo (Reference 3) at RCS temperatures exceeding 212"F. The hydrostatic test is performed overy 10 years. The leakage test is performed frequently during plant life.

- Specification 1.0 defines the hot shutdown mode as a condition when the reactor modo switch is  !

In the shutdown position and the RCS temperature is above 212 F, The Technical Specifications -

require a number of systems, including omorgency core cooling (ECCS), to be operable when the RCS temperature excoods 212*F. The required hydrostatic pressuro and Inservico loak testing cannot be conducted without making some of those systems inoporable. The croposed changes will allow testing to proccod with inoperablo systems.

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Attachment 11 to JPN 92-002 SAFETY EVALUATION .

Page 4 of 9 B. EDITORIAL CHANGES i

The application also makes two editorial changes. These include the addition of the number 29 to the list of amendments effecting pago 137 to cortoct a typographical error from Amendmont 113 (Reforence 8) and the addition of the words "Not Used* on pago 138 to clarify that thoto is no associated surveillance requitomont.

Ill. BAFETY IMPLICATION OF THE PROPOSED CHANGES i

A. HYDROSTATIC AND LEAKAGE TESTING CHANGES Hydrostatic testing and system leakage testing of the Reactor Coolant System (RCS) is requirod l by Section XI of the ASME B&PV code. NRC Genoric Letter 8811 (Reference 5) is usod to i calculate the reactor pressure vessel pressure and temperaturo (P.T) limits toquired for this tost.

- The P.T curves defining thoso limits are porlodically recalculated to considor the results of

. analysos of irradiated survoillanco specimons to account for accumulated reactor fluence.

Tho curtont curves (Figure 3.G 1) require thtt these tests be conducted at RCS temperaturos vproaching 1907. Because decay heat and mechanical heat used to heat the roactor coolant do not allow exact control, the operators require margin to maintain the tost temperatyre betwoon the minimum temperature limit and the maximum temperature limit of 2127. That margin is small at this timo, in addition, the Technical Specl5 cation curves will be revised to require temperatures 4 that exceed 2127 as the accumulated fluence increases. An extrapolation from the minimum test i temperature at 16 effective full power years (EFPY) Indicates that minimum testing temperaturo will peak at about 2507 at 32 EFPY The required test pressure is up to 1105 psig. Those values <

define the conditions for hydrostatic pressure and leak tostlng attor additional temperaturo margin

' is allowed to account for the control of heating, ,

Above 2127, the Technical Soccifications requito a number of systems to bo oporable. Some systems cannot be made operable during testing. The current Technical Specifications wero

- written in anticipation that the reactor would be going into operation when the temperature was raised to 2127 and requires the necessary complement of systems to be available. The proposed change would a: low the RCS to be tested at temperatures above 2127 with a roduced complement of sabty systems. The test duration, test frequency and limited system energy during testing do not require the same complement of systems as plant startup. This has boon qualitatively evaluated by looking at tho test conditions, technical specification requiromonts with the proposed change and the potential consequencos of an accidont during tho test. This will not rosult in a substantial reduction in safoty margin from the curront Technical Specifications, Tet conditions  :

The hydrostatic pressure test occurs onco every 10 years Leakago test typically occurs following a refueling outago and therefore has a frequency of about once por eighteen months.

- Recirculating pumps will bo in operation and a water solid reactor coolant system will be maintainod to control the noesst ry pressure and temperature. Roactor water makeup, pressure, and lovel will be controllod through the Control Rod Drive and Reactor Water Cleanup Systoms,

1 o ' .

Attachmont il to JPN 02-002 SAFETY EVALUATION Pago 5 of 9 During tho test all control rods will bo insorted to assuto shutdown reactivity. The transition from below 2127 to a higher temperaturo represents an increaso in the enthalpy of tho ronctor coolant >

whilo the coro romains subcritical since the temperature increaso is due to docay heat and muchanical heating. The docay heat in tho ooro is at a mlnlmum sinco tosting is performod attor rofueling or maintenanco activity with the rcactor in a cold condition.

The prossure and temperaturo toquired for this test will romain with$n design limits. Tho toqulrod pressuto is about 1105 psig (1.1 timos the operating pressure) for the hydrogt9fic lost and about 1005 psig (oporating pressuto) for the loakago test. At forty offectivo full power years, the tomporaturo at thoso pressures was estimated to bo about 2507. This is Increased by 507 to account for testing margins.

Technical Specification Effects Tho proposed chango will reviso tho technical spocifications to pormit this testing. The technical specification roquirements that will not bo in offoct during hydrostatic and loakage testing and their potential safoty significanco are discussed below:

. Specification 3.5.C: Requires that the High Prossuto Ooolant Injoction (HPCI) Systom bo operable when itradiated fuolis in the vossel, the reac'or pressuro is groator than 150 psig and the reactor coolant temporature is greator than 2127. HPCl in not oporablo during the test duo to the water solid condition of the plant. Sinco the core will bo maintalnod suberitical and tho operators can terminato the tost if thoro is excessivo leakago, which will bo dotocted during tho test, the safoty function of this system will not bo required during this tost, e Spocification 3.5.D: Roquiros that the Automatic Dopressurization (ADS) System be oporable when irradiated fuol is in the vossol, reactor pressuro is greator than 100 psig and prior to startup from the cold condition. The ADS has not boon ovaluated for operability in the water solid condition and may not be operable. Furtholmoro, safoty/rollof valvos (SRV) may havo to be gagged if test pressures excood the SRV sotpoints. Since the core will be malntalnod subcritical and the operators can terminato the test if thoro is excessivo leakago, which will be dotected during tho tost, the safety function of this systom will not be roqultod during this lost.

. Specification 3.5.E: Roquires that the Reactor Core Isolation (R0lC) System be operable under the samo conditions as the HPCI system (Specification 3.5.C). RCIC is not operablo during the test duo to the water solid condition of the plant. Since the coro will be maintalnod subcritical .

and the operators can terminato the test if thoro is excessive leakago, which will be detocted during the lost, the safety function of this system will not be requirod during this tost.

. Specification 3.6.E: Roqulros the SRV's to bo operablo when the roactor coolant system exccods atmospheric pressure and the temperaturo is groator than 2127. Thoso valvos nood not bo operablo as por Specification 3.5.D.

Potor tlal Consequences '

The consequences of plant design basis events while performing hydrostatic pressure and inservice loak testing above 2127 woro qualitatively assossad. Malntaining primary containment assures that design basis events at power are more limiting. Under the hydrostatic pressure and leak test conditions, the worst caso accident is a loss of coolant accident. A largo broak LOCA is

)

Attachment il to JPN 92-002 SAFETY EVALUATION Pago 0 of 9 postulated to occur even though curtont industry leak beforo break analysos (Rotoronco 0) havo indicated that the pipo broak will be procodod by a system Icak. The potential consequences of the LOCA were qualitatively assessed to de'3rmino the offects on the primary system and tho offects on the secondary containmont environmont.

The offects of a small or largo break LOCA are considorod to bo bounded by the existing plant analysos. No coro damago will result sinco the control rods are maintainod fully insorted to maintain subcriticality margins. Thoro will be adoquato roactoi coolant to romove the limited docoy heat. With a small broak, the RCS will depressurize while tho oporator terminatos the test and initiates RHR cooilng and or low pressure ECCS, as noodod. Docay heat from the reactor coro will tend to inctcaso RCS prossuro and temporaturo but tho offoct will bo limited because of tho decay timo. With a largo broak, *.ho reactor will deptossurizo immodlately and all the low prossure ECCS systems with their initiating instrumontation will be availablo. The nood to terminato the test duo to leakago can bo identified during leakago Inspections.

The consoquoncos of a break in primary contalnment are limlied. Tho onthalpy addod to bring the water in the RCS from 212"F and atmosphoric pressure to 300"F and 1105 psig is significant.

However, the primary containmont is designod for the LOCA during powor oporation when tho onthalpy 1. much groator, Tho systems to control containmont atmosphoro aro availablo in both casos, if a broak wero to occur in HPCI or " 91C steam lines outsido ol centainment, the resulting localized offocts ato expected to bo h a sovoro than for high enorgy lino brOaks because thoro is much loss energy. Although the low steam pressuro latation trips loads ato lifted, oporator action is available to terminato flow. Even if tho oporator did not term! nato flow immediately, thero is a 4

timited amount of water availabio in tho steam linct.

Primary containmont integrity will be maintained during the test even though thoro is a requiromont for personnel access. Personnel safety is not contidorod to bo jeopardized industry leak beforo broak analysos (Reforenco 6) indicatos that there will be e small loak well befor9 a crack could propagato into a largo break. Inspectors will ontor the containment and atcas of secondary containment whoro pressurized piping is located in ordor to inspect for leakago. For tho hydrostatk, toct, a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> waiting period is provided to allow for detection of potentialloakago concealed by insulation. The offects of significant steam or water loahago would be readily observnblo. Additionally, tho duration of the 10st8 are not sufficient for tho !cak to prcpagato into a larger broah Radiological Consequences in the LOCA ovaluated in the FSAR, primary containment is designed to contain radioactivo roloasos. During testing, primary containment is maintainod. Other systems designed to rostrict radiological releaso (e.g., secondary containment, SGTS) will be availablo. A LOCA during a hydrostatic or leakago test will result in a lim!!od sourco term sinco thoro is no fallod fuel. Tho LOCA during oporation is thorofore bounding.

For the steam lino broaks outsido containmont, any release of radioactivity would be to tho secondary containmont. The potential radiological consoquences are bounded by tno existing plant analysis of a fuel assembly drop during rofuoling. During refueling, the primary containment is not maintained and a fuel rod drop accident, discussed in FSAR Section 14.G.1.4, is postulated

r

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Attachment il to JPN 92 002 {

SAFETY EVALUATION i Page 7 of 9 l l

to result in the failure of 442 fuel rods for calculating doso consequencos, The resultant offsito -

, dosos aro loss than 2% of the 10 CFR 100 limits. For the steam line break, thoro is no fallod fuel l and the samo complomont of safoty systems used to limit doso consequences will be availablo. .

These changes to the Technical Spocifications do not attor the conclusions of the plant's accident i analyses as documented in the FSAR or the NRC staff's SER, i I

i B. EDITORIAL CHANGES  !

Editorial ;aanges include the addition of a number 23 to the list of amendmonts and tho addition  :

of the words 'Not Usod' to clarify that thoto is no associated survolllanco roqulrt aont. Thoso {

changos havo no safoty significance. ';

IV.- EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION  ;

Operation of the FitzPatrick plant in accordance with the proposed Amendment would not lnyc!vo  ;

- a significant hazards consideration as definod in 10 CFR 50.92, since it would not.

1. Involve a significant increase in the probability or consequences of an accidont previously evaluatod, The proposed amendment rovlsions involve no hardware changos, no changes to tho operation of any systems or components, no changos to structuros, and alters procedures only to the extent that the 212"F limit can be exceeded with certain -

systems inoperable. Those systems are required for core cooling at high pressure.

Any event requiring core cooling will rapidly depressurire the system. The increased i temperature adds enthalpy to the reactor coolant during the test but the consequences of previously evaluated accidents envelope any potential events. The probability of an accident during tosting is expected to increase by a minimal amount  ;

i but this probability is still below that for operations. The test temperatures and pressures are still within system design limits. The test is required to demonstrate the pressuro rotalning capabilities of the RCS pressuro boundary.

- 2, creato the possibility of a now or different kind of accident from thoso previously evaluated.

The proposed amendmont revisions involvo no hardwaro changes, no changos to (ho i operation of any systems or compononts, no changes to structures, and a!! ors procedures only to the extont that the 2127 limit can be exceeded with certain systems laoperablo. The testing procedure will not change the tost process but will

• allow increased temperature during testing te moet NRC guidance and allow margin to the minimum temperature limits.

._ ,~ ~ ,_,.._--,.--.m .-=-...& . . - . . _ , . - -.,.m. -- , - . - . - . . . _ . , _ . - . - , .- - . - - -

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. Attachmont 11 to JPN.92@2 SAFETY EVALUATION Page 6 of 9

3. Involvo a sl0nificant reduction in tho matgin of safety.

Tho proposed amondment ravisions involvo no hardware changos, no changos to tho operation of any systems, no changes to structures, and alter proceduros only to tho extent that the 212"F limit R.11;e excoodod with cortain systems inoporablo. Primary containment and most othot (Wtoms toqulrod for plant transionts and accidents are availablo. Tho coro cooling fection can be maintainod with no chango to the margin of safoty. Tho additional onthalpy to the reactor coolant will reduce by a small amount tho margin that oxisted during prior hydrostatic tests but remains within tho envelopo of previously evaluated plant conditions.

V. IMPLEMENTATION OF THE PROPOSED CHANGES Implomontation of the proposed changes will not adversely affect the ALARA or Fire Protection Program at the FitzPatrick plant, nor will the changos affoct tho environment. Those changes are limited to increased primary system temperatures dudng a prooxisting tost. Tho testing process will not chango and theroforo can havo no impact.

VI. CONCLUSION Thoso changos, as proposod, do not constituto an unrovlowed safety question as definod in 10 CFR 50.59. That is, they:

a. will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis oport;

b. will not increaso the possibility for an accident or malfunction of a typo different from any ovaluated previously in tho Safety Ana!ysis Report;

c. will not reduce the margin of safety as defined in the basis for any technical specification; and

d. Involves no significant hazards considoration, as definod in 10 CFR 50.92.

Vll. REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Sections 4.7,6.4.1, and 14.6.1.4.

2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplomonts.

3. ASME Boller and Pressuro Vossel Code,Section XI,1980 Edition through Wintor 1981, Article IWB 5000.

I l

. . - - _ _ . _ _ - - - - - . - _ - . - - - - - - . - - - - - - . _ _ - - --- - - _ . - - - - _ _ . _ _ _ - - , _ _ _ - , - _ . _ . _ _ m

o. o , l Attachment il to JPN 92-002 SAFETY EVALUATION Pago D of 9

4. Inservico Testing Program for James A. FitzPatrick Nuclear Power Plant, Second Inservico Interval, Revision 4, dated May 1,1991.

5. NRC Gonoric Lottor 8811, 'NRC Position on Radiation Embrittlomont of Reactor Vossol Materials and its impact on Plant Oporations*, dated July 12,1988.

O. NUREG 1001, 'Roport of the U.S. Nuclear Rogulatory Commission Piping Review Committoo,' Rovision 0, dated August 1984.

7. NRC lotter, D. LaBargo to J.C. Drons, dated Februar'y 13,1991, (JAF 91471) transmits Amendmont 168.

8. NRC lettor, H. Abolson to J.C. Brons, dated Octobor 22,1987, (JAF 87 252) transmits Amondmont 113.

9. James A. FitzPatrick Nuclear Power Plant Operations Survollianco Test Proceduro, ST 39H, *Roactor Vossol Operational Prosturo Test (ISI),' Rovlslon 15, dated Juno 8, 1990.

10. GE Report DRF 137-0010, "Imptomontation of Rogutatory Guido 1.99 Revision 2 for the James A. Fit 2 Patrick Nuclear Power Plant", June 1989.

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