ML18052A844
| ML18052A844 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 01/21/1987 |
| From: | Wambach T Office of Nuclear Reactor Regulation |
| To: | Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8702030476 | |
| Download: ML18052A844 (6) | |
Text
January 21, 1987 Docket No. 50-255 Mr. Kenneth W. Berrv Director, Nuclear Licensing Consumers Power Company 1945 West Parnall Road Jackson, Michigan 49201
Dear Mr. Berry:
By letter dated October 7, 1986, we transmitted a number of Technical Specification pages reflecting corrections and/or retyped pages (for legibility). Inadvertently, in retyping and correcting some of the pages some typographical errors were made, which you called to our attention in a subsequent phone call.
Enclosed are corrected pages ii, 3-3, 3-18, and 6-6. Please accept our apologies for any inconvenience these errors may have caused you.
Enclosures:
As stated cc w/enclosures:
See next page Sincerely,
.Original signed by Thomas V. Wambach, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B DISTRIB~H:S' C tJocket BG rimes TBarnhart-4 WJones EButcher ACRS-10 CMiles RDiggs PBD#8
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NRC PDR L PDR PBD-8 Rdo FMiraglia PKreutzer TWambach OGC-Bethesda LHarmon EJordan l(f PBD#8 AThadani 1!-z-r87 Gray File NThompson JPartlow Region V-4
Mr. Kenneth W. Berrv
..- ~nsumers Power Company cc:
M. I. Miller, Esquire*
Isham, Lincoln & Beale 51st Floor..
Three First National Plaza Chicago, Illinois 60602 Mr. Thomas A. McNish, Secretary Consumers Power Company 212 West Mich1aan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Jerry Sarno Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Palisades Plant ATTN:
Mr. Joseph F. Firlit Plant General Manager 27780 Blue Star Memorial Hwy.
Covert, Michigan 49043 R~sident Inspector c/o U.S. NRC Palisades Plant
- 27782 Blue Star Memorial Hwy.
Covert, Michigan 49043 Palisades Plant Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health P.O. Box 30035 Lansing, Michigan 48909
TECHNICAL* SPEClFICATIONS TABLE OF CONTENTS (Co~ti~~ed) 3.0 LIMITING CONDITIONS FOR OPERATION (Continued)
PAGE
.. ~.J 1 Power Di stri buti on Instrumentation.*......................... 3-65
- 3. l l. 1 I ncore Detectors.................................... 3-65
- 3. 11.2 Excore Power Distribution Monitoring System......... 3-66a 3.12 Moderator Temperature Coefficient of Reactivity............. 3-67 3.13 Containment Building and Fuel Storage Building Cranes....... 3-69 3.14 Control Room Ventilation.~... ~.............................. 3-70 3.15 Reactor Primary Shield Cooling System................*...... 3-70
- 3. 16 Engineered Safety Features System lniti ation..
Instrumentation Settings....... *........................... 3-71 3.17 Instrumentation and Control Systems..................*...... 3-76 3.18 Secondary Water Monitoring Requirements..... ~............... 3-82 3.19 Iodine Removal System....................................... 3-84 3.20 Shock Suppressors (Snubbers).........................*...... 3-88 3.21 Movement of Shielded Shipping Cask In Fuel Handling Areas... 3-92
- 3. 22 Fire Protection System...................................... 3-96 3.22.1 Fire Detection.Instrumentation.. :................... 3-96 3.22.2 Fire Suppression Water System....................... 3-98
- 3. 22. 3 Fi re Sprinkler System................................. 3-100 3.22.4 Fire Hose Stations........... ~... ;.................... 3-101 3.22.5 Penetration Fire Barriers... -...... ;................. 3-102 3.23 Power Distribution Limits................................... 3-103 3.23.1 Linear Heat Rate (LHR)........... r ****** r ** ~ ******** 3-103 3.23.2 Radial Peaking factors................ :............. 3-111 3.23.3 Quadrant Power Tilt - Tq**************************** 3-112 3.24 Radiological Effluent Releases.............................. 3-114 3.24.1 Radioactive Liquid Effluent Monitoring
- lnstrurilentati on...... *...................... *......... 3-114 3.24.2 Radioactive Gaseous Effluent Monitoring Iristrumentation................................... 3-117 3.24.3 Liquid Effluents Concentration...-.................... 3-120 3.24.4 Liquid Effluent Dose...... ~************************* 3-121 3.24.5 Gaseous Effluents Dose.............................. 3-122
- 3. 24. 6 Gaseous Waste Treatment System...................... 3-125 3.24.7 Solid Radioactive Waste............................. 3-126 3.24.8 Total Dose.......................................... 3-127 Bases for 3.24 - Radiological Effluent Releases............. 3-128 ii Amendment No.
37,~3,~~'~J,85
3.1 PRIMARY COOLANT SYSTEM (Cont'd)
-:~Ji allowed during nonnal operation, so that substantial safety margfo exists between this pressure differential and the pressure differential required for tube rupture.
Secondary side hydrostatic and leak testing requirements are consistent with ASME BPV Section XI (1971).
The differential maintains stresses in the*
steam generator tube walls within code allowable stresses.
The minimum temperature of 100°F for pressurizing the steam generator secondary side is set by the NOTT of the manway cover of + 40°F.
The transient analyses were6performed assuming a vessel flow at hot zero*
power (532°F) of 126.9 x 10 lb/h minu$ 6% to account for flow measure-ment uncertainty and core flow bypass.l3J A steady state DNB analysis was also perfonned (assuming ll5% overpower, 50 psi for pressure uncertainty, 3% for flow measurement uncertainty, and 3% for core flow bypass) in a parametric fashion to detennine thecore inlet temperature as a function of pressvrj and flow for which the minimum DNBR at 115% overpower is equal to l.3o. l4 The result. of this steady state DNB analysis was the following equation for limiting *reactor inlet temperature:
Tinlet ~ 541.0 + 0.03938 (P-2060) + 0.00004843 (P-2060) 2 + 1.0342 {W-120.2)
A temperature measurement uncertainty of 3°F was subtracted from this limit in arriving.at the LCO given in Section 3.1.1.g. The nominal full power inlet temperature is 2°F less than the value given in Section 3.1.1.g.to allow for drift within the temperature control band.
Thus, a total uncer-tainty of 5°f is applied* to the limiting reactor inlet temperature equation.
The limits of validity of this equation are:
1850 < Pressure ~ 2250 Psia 110.0 x 106 < Vessel Flow < 130 x 106 Lb/h The restrictions on*starting a Reactor Coolant Pump with one or more PCS cold legs < 250°F are provided to prevent PCS pressure transients, caused by energy additions from the s,econdary system, which would exceed the limits of Appendix G to 10 CFR Part 50.
The PCS will be protected against over-pressure transients and will not exceed the limits of Appendix G by either
{l) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting o~ the.RCPs to when the secondary water temperature of each ste9m generator is less than 70°F above each of the PCS cold leg temperatures.l5)
References (1)
FSAR, Sections 6.1.2.2 and 14.3.2.
(2)
FSAR, Section 4.3.7.
(3)
XN-NF-77-18.
(4)
XN-NF-77-22.
(5)
"Palisades Plant Overpressurization Analys.is,H*June, 1977, and "Pali sades Pl ant Primary Cool ant System Overpressurization Subsystem Description," October, 1977.
3-3 Amendment No.
~J, 51
. -~*-... ~
3.1 PRIMARY COOLANT SYSTEM (Cont'd)
Basis The limitation~ on ttJe.spec;ific;.activity of..the*primarrco9Ja~t' ens*ure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite electrical power.
P~rmitting power operation to continue for limited time periods with the primary coolant's specific activity> 1.0 µCi/gram dose equivalent I-131, but below 40 µCi/gram, acco111T1odates possible iodine spiking phenomenon which may occur following changes in thermal power.
Operation with specific activity levels exceeding 1.0 µCi/gram dose equivalent I-131 but below 40 µCi/gram must be restricted to no more than *10 percent of the unit's
- yearly operating time since the higher activity levels increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 15 following a postulated steam generator tube rupture.
Reducing Tavg to< 500°F prevents the release of activity should a steam generator tube rupture since the saturation pressure of *the primary coolant is below the lift pressure of the atmospheric steam relief valves.
The survei.llance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained ori iodine spiking will be used to assess the parameters associated with spiking phenomena.
A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
The basis for the 40 µCi/gram limit is to provide margin for increased tube leakage as a result of a postulated steam line failure and to ensure that the consequences of this event would be within the 10 CFR Part 100 exposure guidelines.
3-18 Amendment No. 20 (next page is 3-20)
RESPONSIBILITIES (Continued)
- c. Review of all proposed changes to Appendix "A" Technical Specifications.
- d.
Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
- e.
Investigation of all violations of the Technical Specifications.
(A
, report shall be prepared covering evaluation and recommendations to prevent recurrence and forwarded to the Vice President - Nuclear Operations and to the Executive Engineer - NAPO.)
- f.
Review of plant operations to detect potential nucl~ar safety hazards.
- g.
Performance of special reviews and investigations and reports thereof as requested by the Plant General Manager or Chairman of NSB.
- h.
Review of the Site Emergency Plan and tmplementing procedures.
- i. Review of all reportabl~ events as defined in Section 1.4.
PRC review may be performed through a routing of the item.subject to the requirements of Specification 6.5.1.7.
6.5.l.7 AUTHORITY The PRC shall :
- a.
Recommend in writing to the Plant General Manager approval or dis-approval of items considered under Specifications 6.5.l.6.a through
- d. above.
- b.
Render determinations in writing with regard to whether or not each item considered under Specifications 6.5.1 l6.a through e. above constitutes an unreviewed safety question.
- c.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President -
Nuclear Operations and to the Vice Chairman of NSB of any disagreement between the PRC and the Plant General Manager; however, the Plant General Manager shall have responsibility for the resolution of such disagreements pursuant to Specification 6.1.1 above.
The PRC Chairman may recommend to the Plant General Manager approval of those items identified in Specifications 6.5.1.6.a. through d; above based on a routing review provided the followfog conditions are met:
(1) at least five PRC members, including the Cha.irman and no more than 2 alternates, shall review the item, concur with determination as to whether or not the item constitutes an unreviewed safety question, and provide written comments on the item: (2) all comments shall be resolved to the satisfaction of the reviewers providing the comments; and (3) if the PRC Chairman determines
- that the comments are significant, the item (including comments and resolu-tions) shall be recirculated to all reviewers for additional comments.
The item shall be reviewed at a PRC meeting in the event that:
(1) comments are not resolved; or (2) the Plant General Manager overrides the recommen-6-6 Amendment No.
i~,7~,~~,86