ML19224D611
| ML19224D611 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/30/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19224D609 | List: |
| References | |
| NUDOCS 7907130140 | |
| Download: ML19224D611 (7) | |
Text
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UNITED STATES I'
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NUCLEAR REGULATORY COMMISSION 5
WASHINGTON. O C. 20555
=, %Lh/ I JERSEY CENTRAL POWER & LIGHT COMPANY C0CKET ';0. 50-219 OYSTER CREEK NUCLEAR GENEPATING STATION, UNIT NO. 1 A"ENCMENT TO D00VISICNAL OPERATING LICENSE Arendment No. 36 License No. CPR-16 1.
The Nuclear Regulatory Comission (the Comission) has fcund that:
A.
The 3colication for amendment by Jersey Central Power & Licht Comcany (the licensee) dated May 19, 1979, corolies with the standards and recuirements of the Atomic Energy Act of 1954, as arended (the Act), and the Comission's rules and reculations set fortn in 10 CFR Chacter I; 3.
The #acility will ooerate in conformity with the acclication, the orovisions of the Act, and the rules and reculations of the Comission; C.
There is reasonable assurance (i) that the activities autnorized by this 3rendment can be C;nducted withcut endangerinq the health and safety of the cuolic, and (ii) that sucn activites will be conducted in cornliance with the Comission's recul stions; J.
The issuance of this amendrent will not be ininical to the corron defense and security or to the health and safety of the oublic; and E.
~he issuance o# :nis amencnen: is in accordance with 10 CFR Part 51 of tne Commission's re<;ula -' and all aco'icable requirements have been satisfiec.
369 031 7907130/YA'
. 2.
Accordingly, the license is amended by changes to the Technical Scecifications as indicated in the attachment te this license amendment and paragraph 3.3 of Provisional Ocerating License tio.
OPR-16 is hereby amended to read as follows:
B.
Technical Scecifications The Technical Scecifications contained in Accendix A, as revised nrough Anendment No. 36, are hereby incorocrated in tne license. The licensee shall ocerate the facility in accordance with the Technical Specifications.
3.
This license amencrent is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COM?iISSION
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Dennis L. Ziemann, Chief Ocerating Reactors Branch #2 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: May 30, 1979 369 032
ATTACHMENT TO LICENSE AMEN 0 MENT N0. 36 PROVISIONAL OPERATING LICENSE NO. DPR-16 00CKET N0. E0-219 Revise Accendix A Technical Soecifications by removino the caces identified below and insertina the enclosed 'caces.
The revised caces are identified by the caotioned anendment number and contain vertical lines indicating tne areas of chance.
REMOVE INSERT 2.1 -2 2.1-2 2.1-2a 2.1-4 2.1-4 2.1 -4a 369 033
. 1.,
D.
During all modes of reactor operation with '.rradiated l
fuel in the reactor vessel, the water level sna11 not be less than 4'-S" above the top of the no: mal active fuel :ene.
E.
The existence of a minimu= critical power ratio (MCPR) less than 1.32 for 7 x 7 fuel and 1.34 for S x 3 fuel shall constitute violation of the fuel cladding integrity safety limit.
F During all modes of operation except when the reactor j
head is off and the recctor is flooded to a level above the main steam no:cles, a: ! east two (2) recirculation loop suc: ion valves and t'.;eir associated discharge valves will be in the full open position.
Bases:
The fuel cladding represents one of the primary physical barriers which separate radioactive material from the environs. The integrity of this clacding barrier is related to its relative freeden from pe-forations or cracking. Although sc=e corrosion or use-related cracking may occur during the life of the cladding, fissica product migration ' rom this source is incrementally cumulative, continuously neiaurable and tolerable.
Fuel cladding perforations. however, could result from thermal effects if reactor operation is significantly above design conditions and the associated protection system se:c.oint.
While fission product migration frca cladding perforation is just as measurable as that from use-related cracking, the thermally-caused cladding perforations si;nal a threshold, beyond which still greater thermal ccnditions may cause ;ross rather than incremental cladding deterioration.
Therefore, the fuel cladding safety lini: is defined in terms of the reacto; operating conditions which may resul in cladding perforation.
A critical heat flux occurrence results in a decrease in hea:
transferred from the clad and, therefore, high clad temperatures and the possibility of clad failure. However, the existence of a critical heat flux occurrence is not a directly observable parameter in an operating reactor-Furthermore, the critical heat flux correlation data which relates observable par 1 eters to the critical heat flux magnitude is statistical in nature.
369 034 Arendrent ':o. 76,36
2.1-Ca
.se margin to boiling transition is calculated free plant operating parameters such as core pressure, core flow, feedwater temperature, core power, and core power distribution. The margin for each fuel assemb1;. is characteri:ed by the critical power ratio (CPR) which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle pcwer.
The minimum value of this ratio for any bundle in the 7re is the minimum critical power ratio (3CFR) (10),
The safety limit curves shcwn in Figure 2.1.1.cpresent ccnditions which assure with be::er than 95 percent ccnfidence a 95 percent prc' abili:y cf avoiding a critical heat flux occurrence. The c
critical pcwer value was determined using the design basis critical power corre.ation given in Reference 1.
The operating range wit!
':CFR >1.32 fer 7 x ' fuel and 1.34 for S x 5 fuel is below and to the right of these curves.
369 035 Arendrent ':0. 76, 26
2.1-4 The range in pressure used for Specification 2.1. A in the calcu-lation of the fuel cladding integrity safety limit is fro: 600 to 1250 psia. Specification 2.1.B provides a requirement on power level when operating below 600 psia or 10' flow.
In general, Specification 2.1.S will only be applicable during startup or shutdown of the plant. A review of all the applicable low pressure and low f'.cw data (6,7) has shown the lowest data point for transition boili.g to have a heat flux of 144,000 BTU /hr-ft2 To insure arpacability to the Bh2 fuel rod geometry, and provide a carpin, a factor of one-half was used, giving a critical heat flux of 72,000 3TU/hr-ft2 This is equivalent to a core average power of 354 MW: (18.3'. of rated).
This value is applicable to ambient pressure and no flow conditions.
For any rzeater pressure or flow conditions, ther e is increased margin.
During transient operation, the heat flux (thermal power-to-water) would lag behind the neutron flux due to the inherent heat transfer time constant of the fuel of S-9 seconds. Also, the limiting safety syste~ e am settines are at values which will not allow the reactor to be operated abova ^ e safety limit during normal operation or during other plant operating ;ituations which have been analyced in detail (2,3,4,5,9,10).
If the scra= occurs such tha the neutron flux dwell time above the limiting safety system setting is less than 1.75 seconds, the safety limit will not be exceeded for normal trubine er generator trips, which are the most severe no mal operating tran-sients expected.
Following a turbine or generator trip, if it is determined that the bypass system malfunctioned, analysis of plant data will be used to ascertain if the safety limit has been exceeded, according to Specification 2.1.A.
The dwell time of 1.75 seconds in Specification 2.1.C provide,s increas-' margin for less severe power transients.
Should a power transient occur, the event recorder would show the time interial the neutron flux is over its scram setting. Knen the event recorder is out of service, a safety limit violation eili be assumed if the neutron flux exceeds the serm se::ing and control rod scran does not cecur.
The event recorder shall be returned to an operable condition as scon as practical.
M reactor water level should drop below the top of the 10:ive
.uel, the ability to cool the core is reduced.
This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.
With a water level above the top cf the active fuel, adecuate cooling is maintained and the decay heat can easily be acco=modatad.
S.e lowest peint at which the water level can presenti, be moni:crei is c '-S" above the tcp of the active fuel.
Altho g h the lower:
reactor water level limit waich ensures adecuate core coolin; is the tcp of the a::i'te fuel, the safety limit has been established 2:
l'-3" to provide a roint which can be tenitored.
369 036 Arencrent '!c. N. 26
1-4a Specification F assures that an adecuate flow path exists from the annular space, between the pressure vessel wall and the core shroud, to the core region. This provides for good ce=unication between these areas, thus assuring that reactor water level instru:nent readings are truly indicative of the water level in the core region.
Arendrent 'c.
If,
'E 369 037