ML20086T711

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Proposed Tech Specs Incorporating Updated Pressure Vs Temperature Operating Limit Curves Contained in TS Figure 3.4.6.1-1
ML20086T711
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 07/27/1995
From:
Public Service Enterprise Group
To:
Shared Package
ML20086T701 List:
References
NUDOCS 9508030131
Download: ML20086T711 (14)
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LR-N95102 LCR 95-07 TECHNICAL SPECIFICATION PAGES WITH PEN AND INK CHANGES The following Technical Specifications are affected by this requested amendment:

Facility Operating License No. NPF-57 Technical Soecification Pace 3/4.4.6 3/4 4-21, 22 F 3.4.6.1-1 3/4 4-23 B 3/4.4.6 B 3/4 4-5, 6,

7 9508030131 950727 PDR ADOCK 05000354 P

PDR

f REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION rester *r' Anncuto 3.4.6.1 The reactor coolant system temperat re and pressure s a be limited in accordance with the limit lines shown on.igure 2.1.5.1-1 (1) curve: ^ end n' fer hydrectati: er ! :E t : ting; (?) cerver 9 :nd 9' #er heater by ana-auc1===

n:, :::!deur fellowing : nucle:r :hutdeu :nd 1:u p urr ""YSICS TESTS; :nd

'3) curves C and C' for ep;retica; with ; critic 1 : r: Other ther !:r perer-EniSICS i TS, with:

a.

A maximum heatup of 100 F in any one hour period, b.

A maximum cooldown of 100 F in any one hour period, c.

A maximum temperature change of less than or equal to 20*F in any one hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves, and d.

The reactor vessel flange and head flange metal temperature shall be maintained greater than or equal to 79*F when reactor vessel head bolting studs are under tension.

APPLICABILITY:

At all times.

ACTION:

With any of the above limits exceeded, 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; determine that the reactor coolant system remains acceptable for continued operations or 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 />.

SURVEILLANCE REQUIREMENTS 4.4.6.1.1 *During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines ofdFigure 3.4.0.1-1 curve; A nd ^', 8 :nd S', er C ;nd C' as applicable, at leas" once er 30 minutes.

TNSERT " L' AITMHE0 HOPE CREEK 3/4 4-21

LR-N95102 LCR 95-07 INSERT 1 TO PAGE 3/4 4-21 Figure 3.4.6.1-1 (hydrostatic or leak testing), and Figure 3.4.6.1-2 (heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS), and Figure 3.4.6.1-3 (operations with a critical core other than low power PHYSICS TESTS),

INSERT 2 TO PAGE 3/4 4-21 Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 i

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be if determined to be to the right of the criticality limit line of Figure 3.4.6.1-Af 0 and C' within 15' minutes prior to the withdrawal of control rods to uurvc=

bring the reactor to criticality and at least once per 30 minutes d system heatup.

,,,,, g a g 4.4.6.1.3 The reactor vessel material surveillance s cimens s a e emoved and examined, to determine changes in reactor press e vessel material properties, as required by 10 CFR 50, Appendix H.

he results of these examinations shall be used to update the curves of F;3m e 3.4.5.1-1 based nn the gr :ter of the f0110 wing criteria:

e.

The :ctu:1 :bift ia reference t0 perature f0r plate ::terial frca 50:t 5K3238-1 and weld -etal 510-01205 as deter-4aed by Charpy 4= pact t;st, er b.

The predicted shift in reference temperatures for plate material

-free h::t 5K3025-1 :: deter ined by Regulatery Guide 1.99, "R:diation 0:::ge to Re:: tor Vessel Materials."

4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 70 F:

a.

In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:

1.

5 100*F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

j 2.

5 80 F, at least once per 30 minutes.

b.

Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

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HOPE CREEK 3/4 4-22 Amendment No. 46

4 LR-N95102 LCR 95-07 INSERT 1 TO PAGE 3/4 4-22 Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 i

1 1

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A* A-C

- SYSTEM MYDA0TE LtMIT WITH FUE L led V L

S-N0%NUC AA MEATessG

[f C-E AR ICORE CA4T8CALI R

T SASED ON G. E. CO.

t300 rygesoy gurm LecEseSteeG T0*lCAL LIMITiteGP REPORT NEDO.31778 A A*. S'. C' - COR E SE L E AFTER AN D 30*F TEMP

/

f SHIFT 00d AN INITIAL P

RTuoy OF WF f

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VES ARE NOT LeedITieeG let0 (SMCmWN FOR DNFOmbd Af t g

y,

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880UlTY f

800TE.

g 800 p

f CURVESA.

sed C ARE PRE DICT TO APPLY AS THE Li TS FOR 40 YE ARS

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O2 PYi OF OPER ATION

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I W 10 CPR to i

APP 9N0tX G

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0 200 300 000 l

hetN8 Mund RE ACT VESSE L MET AL TEMP ATURE l'Fl MINI REACTOR PRESSURE SSEL METAL TEMP ATURE VS. RE OR VESSEL P SURE 1

F.3vus 3 HOPE CREEK 3/4 4-23

s a

1600 l

I A - SYs?EM m Re?Eg? t+w;*.

I CURWA ALL PRECPERATIONAL SYSTEM LEAKAGE i

AND HYDROSTATIC PRESSURE TESTS. AND 1400 ALL SYSTEM LEAKAGE AND HYDROSTATIC I

PRESSURE TESTS PEaroRxED cURINa THE SERVICE LIFE OF THE PRESSURE BOUNCARY l

IN COMPLIANCE WITH THE ASME CCCE, I

f SECTION XI.

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I 1200 I

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1000 i

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400 312

_ PSIG 200 BOLTUP 79'F CURVES ARE VALID FOR 32 EFPY OF OPERATION O

0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

FIGURE 3.4.6.1-1 HOPE CREEK 3/4 4-23

1600 I

8 I

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B - NON-NUc' EAR MEA *"P/

COCLDOWN L'MI* -

ALL HEAT UP AND COOLDOWNS THAT ARE I

PERFORMED WHEN THE REACTCR IS NOT 1400 CRITICAL AT THE NCRMAL HEAT-UP AND

~

C00LDOWN RATE OF 100*P/HR.

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1200 9

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REQ'MTS

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53.5'F SHIFT D

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m m

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- - BOTTOM HEAD

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LIMITS 400 l

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1/

200

- BOLTUP '

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79'F CURVES ARE VALID FOR 32 EFPY OF OPERATION o

o.o 100.o 200.0 300.0 400.0 500.o 600.o MINIMUM REACTOR VESSEL METAL TEMPERATURE l'F)

FIGURE 3-.4.6.1-2 HOFE CR2EK 3/4 4-23a

1600 i

e-mrurentenmC m ALL HEAT-UP AND COOLDOWNS THAT ARE PERrORMED WHEN THE PE. ACTOR IS CRITICAL CURVE C AT THE. NORMAL HEATUP AND COOLDOWN RATE Or too r/HR 1400

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CURVES ARE VALID FOR 32 EFPY OF OPERATION O

0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

FIGURE 3.4.6.1-3 HOPE CREEK 3/4 4-23b i

RENCTORCOOLANTSYSTEM BASES 3/4.4.6 PRESSURE / TEMPERATURE LIMITS 3

All components in the reactor coolant system ar esigned to withstand the effects of cyclic loa's due to system temperatu e and pressure changes.

d These cyclic loads are introduced by normal load t ansients, reactor trips, and startup and shutdown operations.

The various ategories load cycles used for design purposes are provided in Section (.9) of t AR.

During startup and shutdown, the rates of temperature and pressure anges are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy th

'm cyclic operation.

rumr T Armeneo The operating limit curves of,, is., C. M.11 -

derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section III, Appen-dix G.

The curves are based on the RT and stress intensity factor information NDT for the reactor vessel components.

F ure toughness limits and the basis for compliance are more fully discussed i AR Chapter 5, Paragraph 5.3.1.5, " Frac-ture Toughness."

CentL 0 The reactor vessel materia s have been tested to determine their initial ugg RT The results of these tests are shown in Table B 3/4.4.6-1.

Reactor NDT.

"2" operation and resultant fast neutron, E greater than 1 MeV. irradiation will i

,ggy cause an increase in the RTNDT' lT.refore, an justed ref ence tempe ture, based up he fluen

, phosphor s content copper co ent of the aterial 4

in ques n, can b predicted sing Bases igure B 3/4

.6-1 and t recommend tions Regulat y Guide 1 9, Revisio

" Effects f Residual lements o Pre 'cted Radi ion Damag to Reactor essel Mater' 1s."

The ressure/ t era t e limit c ve, Figur 3.4.6.1-1, urves A', B' and C', i udes an ac med ift in R f r th end of lif fluence.

DT The actual shift in RT f the vessel material will be established period-NDT T ically during operation by removing and evaluating, irradiated flux wires 3

installed near the inside wall of the reactor vessel in the core area.

Since N

the neutron spectra at the flux wires and vessel inside radius are essentially identical, the irradiated flux wires can be used with confidence in predicting reactor vessel material transition temperature shift. jThe ope ting lim curves of Figg e 3.4 f.1-1 sh be adjysted, as quired /on the b is of th flux wire /ataap(recomm dations pr Regula ry Guidp/1.99, Re sion 1.

HOPE CREEK B 3/4 4-5

LR-N95102 LCR 95-07 INSERT 1 TO PAGE B 3/4 4-5 Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 INSERT 2 TO PAGE B 3/4 4-5 Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6.-1 and the recommendations of Regulatory Guide 1.99, Rev.

2, " Radiation Embrittlement of Reactor Vessel Material".

The pressure / temperature limit curves, Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3, includes an assumed shift in RT for the end of gg life fluence.

INSERT 3 TO PAGE B 3/4 4-5 The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 shall be adjusted, as required, on the basis of the flux wire data and the recommendations of Regulatory Guide 1.99, Rev.

2.

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) restar"f Mxpro The pressure-temperature limit lines shown in rig r a 3.4.5.1-1, curve:

C, and C', and A end A', for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are pro-vided in UFSAR Section 5.3 and Appendix SA.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE.

The surveillance requirements are based on the operating history of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1977 Edition and Addenda through Summer 1978.

The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda c: required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(1).

3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera-ture indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operstion.

HOPE CREEK B 3/4 4-6 Am0ndaent No. 46

=. _.

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l LR-N95102 LCR 95-07 t

INSERT-1 TO PAGE B 3/4 4-6 i

Figures 3.4.6.1-1 and 3.4.6.1-3, curves for inservice leak and hydrostatic testing and reactor criticality

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s BASES TABLE B 3/4.4.6-1 6

g REACTOR VESSEL TOUGHNESS Pal;M

=" Cit" f...C0'-!"".'."_' T " f. EOL njg HEAT / SLAB HIGHEST MAX m

BELTLINE WELD SEAM I.D.

OR N.'

UPPER SHELF RT a RT RT COMe0NENT OR MAT

  • L TYPE HEAT / LOT CU(%)

MDT(*F)

MDT(*F)

(FT-LBS)

MDT(*F)

~

Plate 3 GR B CL.1 5K3025-1

.15

+19

(, 7 Weld

- c,. seams for

C"/

.06

-30 62.8 12 o J2.6 shells 4&5 xd ;;;i-th

41. Tab o.n n'd ictn;. ".5 O novo /

i n tr-o zzer NOTE:

  • These values are given on1 or the be fit ~of calculating the end-of-life (EOL) RT NDT' HEAT / SLAB HIGHEST REFERENCE MON-BELTLINE MT'L TYPE OR OR P ERATURE C0090NENT WELD SEAM I.0.

HEAT / LOT NDT (*F)

Shell Ring Connected to SA 533, GR.B. C1.1 All Heats

+19 im R

Vessel flange Botton Head SA 533, GR.B. C1.1 All Hests

+30 1

t Botton He o

SA 533, GR.B, C1.1 All Heats

+30 LPCI Nozzi s0)

SA 508, C1.2 All Heats

-20 w

Top Head T us SA 533, GR.B. C1.1 All Heats

+19 Top Head Fla SA 508, C1.2 All Heats

+10 Vessel Flange SA 508, C1.2 All Heats

+10 Feedwater Nozzle SA 508, C1.2 All Heats

-20 Weld Metal All RPV Welds All Heats O

Closure Studs SA 540, GR.B. 24 All Heats Meet 45 ft-Ibs & 25 mils lateral expansion at +10'F 1/4T of the vessel thickness of %el results in these nozzles experiencing a predicted EOL ffuence at (Q

he design of the Hope Creek vess x 1087 n/cm.

Therefore, these nozzles are predicted to have an 2

EOL RT of F.

n MOT 2.s4 4 Z'l I

I

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